ML17200C927

07/26/2017 Entergy Presentation Slides for Public Meeting Regarding Spent Fuel Pool Management Project, New Criticality Analysis

ML17200C927
Person / Time
Site:	Indian Point
Issue date:	07/06/2017
From:	Robert Walpole Entergy Nuclear Operations
To:	Plant Licensing Branch 1
Guzman R
References
CAC MF9931
Download: ML17200C927 (80)

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1 Indian Point Energy Center Spent Fuel Pool Management Project July 26, 2017 AGENDAObjective -B. Walpole (IPEC)Overview -G. Delfini (IPEC)License Amendment Requests -B. WalpoleIP2 SFP Criticality Safety Analysis -N ETCO 2 OBJECTIVE Provide long term plan for Spent Fuel Pool (SFP) management at IPEC Unit 2 (IP2) and Unit 3 (IP3) Provide time frames associated with SFP Management actions Provide details on IP2 Criticality Safety Analysis (CSA) approach 3

4 OVERVIEW Spent Fuel Pool Criticality Control at IP2: Perform new SFP CSA Qualify IP3 Spent Fuel population for storage at IP2 Spent Fuel Pool Inventory Control Manage IP2 SFP inventory via cask loading Manage IP3 SFP inventory via fuel transfer and cask loading 5 LICENSE AMENDMENT REQUESTSInter-Unit Fuel Transfer -Submitted to NRC 12/14/16 Currently in Technical Review ProcessIP2 SFP CSA -Submittal 4 th Qtr 2017Implement in 2019 -Addresses all concerns in the current OperabilityEvaluation and Eliminates the Non-conservative Technical

SpecificationInter-Unit Fuel Transfer -Final submittal -2019Bounds the remainder of fuel for transfer IP2 SFP Criticality Safety Analysis Overview of New IP2 SFP Criticality Safety Analysis Using No Absorber PanelsDale Lancaster NETCO/NuclearConsultants.com 6

Moving ForwardIn the August 2013 submittal (NET-300067-01) Flexibilityand Margin was emphasized as IP2 was intending to add new Metamic absorber panelsIt is expected that IP2 will have only one more refueling.

And the implementation of the criticality analysis is after

the last fresh fuel is in the core. Therefore, flexibility is no longer a priority.The new approach, which is based on the methodology in the previously submitted NET-300067-01 report, emphasizes Realistic assessment of reactivity of thefinitenumber of assemblies to be stored in the pool 7

Governing Documents and Software Codes 10CFR50.68 DSS-ISG-2010-01 and NEI-12-16 Analysis uses SCALE 6.1.2, Keno V.a., TRITON (t5-depl), 238 group ENDF/B-

VII.0 cross sections 8

RealisticActualirradiation history for all of the assemblies burned at IPEC have been reviewed1 of 5 reactivity categories is assigned for each assembly that has ended its irradiation The assessment is based mainly on the burnup and enrichment.

However, when the assembly is near a reactivity category boundary, the actual burnable absorbers, ax ial burnup distribution and soluble boron ppm is also used. Details of each assembly assessment will be in the criticality

analysis report 9

Realistic (cont.)The traditional loading curve is only used for fuel that may still be in the core in the last few cyclesBut even for these assemblies the loading curve is mademore realistic by adding a dependence on the averagepeaking factor to the loading curve. (Note -This is further discussed in the next presentation.)

10 Realistic (cont.)Analysis is dependent on more full pool calculations since the realistic effect of the pool

edge will be fully utilized The analysis will credit control rods that are in

the pool by specifying locations which must have

a control rod or be empty 11 IP2 Spent Fuel Pool Layout 12 5 Reactivity CategoriesCategory requirements for Fuel after Batch X for IP2 and after Batch U for IP3 (older fuel have assigned categories) 1.Fresh 5 wt% fuel with 64 IFBA (Integrated Fuel Burnable Absorber)rods stored as a 2 out of 4 checker board in Region 1 2.21 GWd/MTU (up to 5 wt% U-235) (All once burned fuel will meet this.) stored in a 3 out of 4 pattern in Region 1 3.28.5 GWd/MTU stored on the periphery of Region 1.

This is expected to cover about half of the once burned fuel and all twice burned fuel.

4.Full loading curve including cooling time needing 49.5 GWd/MTU for 5 wt% fuel stored in a 3 out of 4 pattern in Region 2 5.Category 4 plus 11 GWd/MTU stored on the periphery of Region 2 13 Category 4 (Region 2, 3 of 4)

Loading Criteria B 1.2= (a1 + a2*E + a3*E

2) x exp[-(a4 + a5*E + a6*E

2) x C] + a7 + a8*E + a9*E 2 B 0.8= (b1 + b2*E + b3*E

2) x exp[-(b4 + b5*E + b6*E

2) x C] + b7 + b8*E + b9*E 2MRB = B 0.8+ (PF -0.8) x [ B 1.2-B 0.8) ] / 0.4 Where: MRB is the Minimum Required Burnup Eis the enrichment Cis the cooling time PFAssembly Burnup / (Sum of Cycle Burnups) a1-b9are fitting coefficients 14 15 Improvements in the AnalysisThe following improvements have been made since the previous submission and will be discussed in the next presentations:

-Depletion analysis with the peaking factor credit-Axial blanket credit-Validation updates (control rods and temperature dependence)

-Misload analysis assumptions

-Grid growth and creep biases

-Eccentric placement bias (no longer all poisoned cells)

-Interface calculations (more complex full pool) 16 IP2 SFP CSADepletion Analysis -Realistic ApproachCharles T. RomboughCTR Technical Services, Inc.

17 Realistic Approach Emphasis is on using actual depletion conditions for discharged fuel while still being conservativeThe fuel inventory is split into tenbatch groupings that have similar depletion

characteristics 18 Realistic Approach 19 Batch Groupings 1.A, B, C, D (Pyrex, no blanket, SS guide tubes) 2.E, F (Pyrex, no blanket, SS guide tubes) 3.Gthru L (Pyrex, no blanket, Zircaloyguide tubes) 4.M, N, P (WABA/IFBA, no blanket) 5.Q, R, S (WABA/IFBA, 6 inch 2.6 w/o blanket) 6.T, U, V (WABA/IFBA, 8 inch 3.2 w/o blanket) 7.W (WABA/IFBA, 8 inch 3.4 w/o blanket) 8.X (WABA/IFBA, 8 inch 3.6 w/o blanket) 9.2A+future fuel (WABA/IFBA, 8 inch 4.0 w/o blanket) 10.IP3 (Pyrex, no blanket, Zircaloyguide tubes) 20 Moderator TemperatureThe moderator temperature increases as water is heatedtraveling upward through the coreThe moderator temperature is highest at the exit and so the moderator temperature is highest for the top nodeThe core average exit temperature can be calculated from the core average temperature and the inlet

temperatureT(exit) = T(in) + 2 x [T(avg) -T(in) ]The exit temperature for any assembly is higher or lower depending on the radial peaking factor 21 Moderator Temperature (cont.)Example: T in= 538.6 o F, T out= 607.6 o FEnthalpy at 538.6 oF, 2250 psia = 533.1 btu/lbEnthalpy at 607.6 = 624.0 btu/lbFor a peaking factor of 1.40, the enthalpy rise would be 1.40 x (624.0 btu/lb -533.1 btu/lb) = 127.3So the peak outlet enthalpy = 533.1btu/lb + 127.3 btu/lb

= 660.4 btu/lbTemperature at 660.4 btu/lbenthalpy = 631.0 o F Density of water at 631.0 oF, 2250 psia = 0.6426 g/cc 22 Moderator Temperature Exit temperature is a function of peaking factor570575580 585 590 595 600 605 6100.50.60.70.80.911.11.21.31.41.5Moderator Temperature (o K)Peaking FactorLinear Fit Points Fuel Temperature The fuel temperature will also increase in proportion to the peaking factor 24 CTR Technical Services, Inc.25Relative Power Fuel TemperatureFuel Temperature at 25 GWd/T in the top node versus radial peaking factor 26600620640660680700720 7407607808000.50.60.70.80.911.11.21.31.41.5Fuel Temperature (o K)Peaking FactorLinear Fit Points Node Specific TemperaturesThe top node has the highest moderator temperature but a lower fuel temperature because the relative axial power is small (typically about 0.5 of the average)The second node has a lower moderator temperature than the top node but a higher fuel temperature than the

top nodeThe third node has a lower moderator temperature than the second node but a higher fuel temperature than the

second nodeThese temperatures depend on the axial burnup

distribution 27 Moderator Temperature(Radial Peaking Factor of 1.4)

Fuel Temperature (Radial Peaking Factor of 1.4) 28 TemperaturesTo simplify the modeling, the most limiting axial burnup profile is selected that maximizes the combined reactivity effect of the moderator and fuel temperaturesWhen all nodes are considered, the reactivity is highest when using the 46+ profile for fuel and moderator

temperatures For the 4 thand lower nodes, calculations show that it is conservative to use the 3 rdnode temperatures because the moderator temperature effect dominates the fuel

temperature effect 29 Temperatures (cont.) For the standard model, the moderator and fuel temperature are calculated for the top node and

the 3 rd node using the 46+ burnup profile The 3 rd node temperatures are then used for the 2 nd , 3 rd ,4 th , 5 th, and lower nodes Calculations have been performed that show

this is conservative by 0.0005 to 0.0007 in k

compared to a 5 node model 30 Peaking Factor CreditThe Peaking Factor is determined as the assembly burnup divided by the sum of the cycle average burnups for the cycles that assembly is in the coreUsing a large peaking factor for all assemblies is acceptable but very conservativeAn assembly that has a large peaking factor will also have a larger burnup (by definition) and so would meet the loading criteriaAn assembly that has a small peaking factor would also have a smaller burnup and may not meet a loading criteria based on a large

peaking factor 31 Peaking Factor Credit (cont.)Since a smaller peaking factor implies lower fuel and moderator temperatures, this fact can be used to credit the lower peaking factor assembliesDepletion is done at a PF of 1.20 (using high temperatures) and again at a PF of 0.80 (using lower temperatures)The burnup requirement for any given peaking factor can be interpolated between these two values because the relationship is linear in this range and has been shown to be conservative when

extrapolated beyond this range 32 Peaking Factor Credit (cont.)Since a smaller peaking factor implies lower fuel and moderator temperatures, this fact can be used to credit the lower peaking factor assembliesDepletion is done at a PF of 1.20 (using high temperatures) and again at a PF of 0.80 (using lower temperatures)The burnup requirement for any given peaking factor can be interpolated between these two values because the relationship is linear in this range and is conservative when extrapolated beyond

this range 33 Burnable AbsorbersBurnable absorbers harden the spectrum and increase the reactivity of burned fuelEarly cycles used Pyrex which displaces more water than WABALater cycles used WABA in combination with a varying number of

IFBA rodsThe worst case absorber loading for recent fuel designs is a 20 rodlet WABA with 148 IFBA rods (1.5X)For the discharged batches, the WABAs are pulled at the highest burnup with WABAs. For the current and future fuel design, the WABAs are never removed.

34 Burnable Absorbers The burnable absorbers used during depletion depends on the batch grouping since the absorber characteristics for each batch grouping are known 35IP-2 Batches BATypeMax BA Loading Max BA Burnup A,B,C,D Pyrex20 rodlets18.5E thru FPyrex12 rodlets12.2 G thru LPyrex20 rodlets16.7M, N, P IFBA WABA 116 (1.0x) 20 rodlets 28.1Q, R, S IFBA WABA 148 (1.5x) 20 rodlets 26.7T, U, V IFBA WABA 148 (1.5x) 20 rodlets 33.8W, X IFBA WABA 148 (1.25x) 20 rodlets 32.62A and future fuel and IP-3 IFBA WABA 148 (1.5x) 20 rodlets 33.2 Control RodsAssemblies that were depleted with control rod insertion are more reactiveThe first 17 cycles at IP2 were operated with D-bank in the "bite"

positionFor these assemblies, the top node is depleted with a control rod for the entire depletionFor any assembly under D-bank in any cycle, the nodes below the top node are depleted with a control rod for 2 GWd/MTU to cover operation with some control rod insertion. The loading criteria for future fuel assumes all assemblies are under D-Bank.

36 Other InsertsOther inserts are handled as before.

Specifically:-Assemblies that contained a hafnium flux suppression insert are subject to a 2 GWd/MTU burnup penalty -Neutron source inserts displace some water but are covered by burnable absorber

assumptions 37 In-core Detector Water Displacement In-core detector systems displace some water but not significant to criticality To conservatively model the in-core detectors, the depletion is performed with a void in the

instrument tube 38 Soluble BoronThe soluble boron used is the average over all cycles the assembly was burned (NUREG/CR-6665 showed it is the same as or conservative to use a burnup averaged ppm compared to a letdown curve)The soluble boron level used for each batch grouping was determined by taking 0.55 times the highest equilibrium Xeinitial boron concentration at full power for cycles without IFBA and 0.70 times the peak boron for

cycles with IFBA 39 Axial Blankets Assemblies with low enriched axial blankets are less reactive than non blanketed fuel assemblies The amount of credit depends on the axial

blanket design-Blanket enrichment (0.7, 2.6, 3.2, 3.4, 3.6, 4.0 w/o)-Blanket length (6 inch, 8 inch)

-Solid or annular 40 Axial Blankets (cont.)For non-blanketed assemblies, the axial burnup distribution is obtained from the DOE data base but when creditingthe lower enriched axial blankets, these shapes cannot be usedAxial burnup distributions for each blanketed assembly

are available from reactor recordsA conservative burnup distribution for a particular blanket design is obtained by using the lowest relative burnup in

each node 41 Axial Blankets (cont.)The most reactive burnup shape for a group of assemblies is the smallest relative power in each node without renormalizingBy doing this, the most reactive shape is selected whether the reactivity is being driven by the center or the

top (end effect)Uniform analysis was required in the past because limiting shapes were based on a top peaked distribution and so uniform analysis would cover center peaked

distributions 42 Axial Blankets (cont.)Selecting the smallest relative power in each node without renormalizing covers center peaked distributions

as wellTherefore, there is no need to analyze a uniform burnup distribution because the most limiting center peaked distribution is covered 43 Axial Blankets (cont.)For the blanketed assemblies 6 inch nodes are used. For a blanket that is more than 6 inches the second node

is split into two nodes. For the IP2 analysis, the top 9 nodes are used to find the smallest relative power in each nodeThe top 9 nodes are modeled as is while the 9 th node from the top is used for all lower nodes (the next 13 nodes down always have a burnup that is higher than

the 9 th from the top node)Calculations show that this approach is conservative compared to modeling all nodes for both top peaked andcenter peaked distributions 44 Special Treatment for Pm-149Pm-149 is produced during depletion and decays into Sm-149 with a half-life of several

daysThe higher the specific power, the more Pm-149

is produced After the assembly is removed from the core, this Pm-149 decays into Sm-149 (an absorber) 45 Special Treatment for Pm-149 (cont.)It is non-conservative to assume a nominal specific power during depletion if there is a coastdown at end of cycleTo account for this effect, the Pm-149 content can be

reducedto conservatively account for coastdownsAlso, if a fuel assembly is in a low power region in the last cycle, the Pm-149 would have to be reduced even

more 46 Special Treatment for Pm-149 (cont.)To account for botheffects (coast down and low power during the last cycle of depletion), the Pm-149 content for all

assemblies is reduced by 75%

47 Depletion ModelDepletion models use nominal dimensionsDimension changes that would harden the spectrum during depletion would have the opposite effect in the spent fuel poolFor example, increasing the clad OD hardens the spectrum for the depletion resulting in more reactive atom densities but the reduction of moderation in the pool is a negative reactivity that more than cancels the slightly increased depletion reactivityAlthough grid growth impacts depletion, it is ignored since it is conservative to ignore it 48 Individual Assembly DepletionAlthough assemblies are analyzed in groups, assemblies near a reactivity category are analyzed with full actual operating data to find the proper reactivity category.

49AssyIDFuel TypeEnrichmentFTDBurnable AbsorberPPMPF A10HIPAR2.210.943None5000.92 F44HIPAR3.350.933None4951.05 L48LOPAR3.690.94416 WABA5200.69 W52OFA4.960.94620 WABA/100 IFBA8800.84 X18OFA4.950.95020 WABA/116 IFBA8800.87 Summary Depletion parameters include

• Burnup averaged soluble boron

• Moderator temperature as a function of radial peaking factor and node

• Fuel temperature as a function of burnup and radial peaking factor and node

• Maximum loading of burnable absorbers for

each batch grouping

• Coast down and low power effects 50 Summary (cont.)The new approach emphasizes actual depletion conditions instead of bounding assumptions that would penalize most assemblies for which the depletion conditions are knownSome assemblies are analyzed individually if near a reactivity categoryPeaking factor credit is taken since the peaking factor for each assembly is known Assemblies with axial blankets are treated separately from full length fuel 51 IP2 SFP CSA Changes in Validation, Misload, Dimension Changes, Eccentricity and Interface AnalysisDale Lancaster NETCO/NuclearConsultants.com 52 Validation Changes The new CSA utilizes Ag-In-Cd control rods for reactivity control Previous validation set had no Ag or In critical

experiments Added 51 critical experiments with Ag-In-Cd

control rods (LCT-088, 92, (Brazil), B&W-1810, WCAP-3269) Mean of Ag-In-Cd critical experiments is 0.9987, the mean of all the critical experiments is 0.9981

+/-0.0015 53 Validation Changes (cont.) New calculation uses water holes so elevated temperatures are limiting. Therefore, added temperature dependent criticality experiments (LCT-46) A temperature bias of 8.6E-6 times the

difference in temperature between the desired

temperature and 20 oCis applied 54 Validation Changes (cont.) Added LCT-96 with 6.9 wt% U-235 fuel in order to:-Fill in trend on spectrum (EALF)-Prevent slight extrapolation on enrichment needed for going from 4.74 to 5 wt% U-235-Add a new set of independent critical experiments (new lab -Sandia) Follows very close to the previous k as a function of EALF 55 56 Validation Changes (cont.)

57 Validation Changes (cont.) LCT-08 analysis has been improvedThis set was done at B&W and the same facility as the B&W-1810 needed for Ag-In-CdLCT-08 benchmark sample runs in the criticality handbook used a 2D modelRevised to a 3D model (included fuel rods above the

water)Increased average k by about 0.07% in k.

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All pertinent changes from RAI's on the previous IP2 submittal (ML15261A528) have been incorporated Multi-Misload Analysis 58IP2 Region 1 analysis assumes all locations are filled with 5 wt% fuel and no burnup. Calculated k at 2000 ppm is 0.824.IP2 Region 2 analysis assumes all locations except water hole or control rod locationsare loaded with 5 wt% fuel with 24 GWd/MTU burnup. Calculated k with 2000 ppm is 0.910 (less than 0.94 after

bias and uncertainty).After the completion of the next cycle (expected to be the last cycle) nearly all the fuel in the pool is expected to be less reactive than 5 wt % and 24 GWd/MTU. (Currently only 4 once-burned assemblies had less than 24 GWd/MTU.)IP2 Region 2 analysis is also done with 5 wt% fresh assemblies w/64 IFBA placed in allwater holes but the rest of the fuel at the Tech Spec requirements. (Calculated k 95/95=0.939 with 2000 ppm using unborated uncertainties)

Clad Creep 59Clad creep is real, resulting in a small positive bias at some burnups.Given the small size of the bias, it would take a big error in the creep model to produce a significant effect on k.

Table 7.1.1 presents our calculations of the creep worth using the creep model identified in the following slides Creep Model The clad is modeled to creep down linearly with burnup for the first 40 GWd/MTU. At that point

the clad has creeped in 100 microns but has

also added a 25 micron oxide layer (net creep of

75 microns). At 40 GWd/MTU the clad is on the pellet and

pellet then grows such that at 56 GWd/T the clad OD (including oxide layer) is the same as the original OD. No credit for further creep out is credited.

60 Creep Model Basis 614. Y. Irisa, et al, "Segmented Fuel Rod Irradiation Program On Advanced Materials For High Burnup," An International Topical Meeting on Light Water Reactor Fuel Performance, Park City, Utah, April 10-13, 2000, American Nuclear Society, La Grange Park, Illinois.

Creep Model Basis 621. G.P. SABOL, G. SCHOENBERGER, and M.G. BALFOUR, "IMPROVED PWR FUEL CLADDING," Materials for Advanced Water Cooled Reactors, Proceedings of a Technical Committee Meeting, Plzen, Czechoslovakia, May 14-17, 1991, IAEA-TECDOC-665, IAEA, VIENNA, 1992.

633. Garzarolli, F., Manzel, R., Reschke, S., and Tenckhoff, E., "Review of Corrosion and Dimensional Behavior of Zircaloy under Water Reactor Conditions," Zirconium in the Nuclear Industry (Fourth Conference), ASTM STP 681, American Society for Testing and Materials, 1979, pp.91-106.

Creep Basis 645. C. B. Lee, et al, "Post-irradiation Examination of High Burnup UO2 Fuel," Proceedings of the 2004 International Meeting on LWR Fuel Performance, Orlando, Florida, September 19-22, 2004, American Nuclear Society, La Grange Park, Illinois.

Pellet Growth 65Assume returns to initial density by 30 GWd/MTU7. R. Manzel and C. T. Walker, "High Burnup Fuel Microstructure And Its Effect On Fuel Rod Performance," An International Topical Meeting on Light Water Reactor Fuel Performance, Park City, Utah, April 10-13, 2000, American Nuclear Society, La Grange Park, Illinois.

Oxide Layer Growth 666. David Mitchell, Anand Garde, and Dennis Davis, "Optimized ZIRLOTM Fuel Performance in Westinghouse PWRs," Proceedings of the 2010 LWR Fuel Performance Meeting/Top Fuel/WRFPM, September 26-29, 2010

• Orlando, Florida, USA, American Nuclear Society, La Grange Park, Illinois.

Creep Summary Using public domain information an approximate creep model is made Due to the small size of the creep bias

more detailed modeling is not required 67 Grid Growth Grid growth is another real effect which causes a small bias.

Table 7.2.1 shows our calculated bias using our grid growth model.

68 Grid Growth Basis 69David Mitchell, Anand Garde, and Dennis Davis, "Optimized ZIRLO TMFuel Performance in Westinghouse PWRs," Proceedings of the 2010 LWR Fuel Performance Meeting/Top Fuel/WRFPM, September 26-29, 2010

• Orlando, Florida, USA, American Nuclear Society, La Grange Park, Illinois.

Grid Growth Basis (cont.)

70DENNIS GOTTUSO, JEAN-NOEL CANAT, PIERRE MOLLARD, "A FAMILY OF UPGRADED FUEL ASSEMBLIES FOR PWR," Top Fuel 2006, 2006 International Meeting on LWR Fuel Performance, October 22-26, 2006, Salamanca, Spain, European Nuclear Society.

Grid Growth Basis (cont.)

71King, S. J., Kesterson, R. L., Yueh, K. H., Comstock, R. J., Herwig, W. M., and Ferguson, S. D., "Impact of Hydrogen on Dimensi onal Stability of ZIRLO Fuel Assemblies," Zirconium in the Nuclear Industry: Thirteenth International Symposium, ASTM STP 1423, G. D. Moan and P. Rudling, Eds., ASTM International, West Conshohoc ken, PA, 2002, pp. 471-489.

Grid Growth SummaryGrid growth is temperature dependent. Therefore, the higher grids experience more growth.The Inconel top grid grows much lessA uniform grid growth of two adjacent assemblies by 0.69% (Westinghouse 15x15) would make the grids

touchModel assumes 0.78% growth at 60.5 GWd/MTU (highest credited burnup).

72 Eccentricity In a pool crediting absorber plates eccentricity reactivity is negative (as was the case for the

previous design)Water holes decrease eccentricity effect so the effect is small Region 2 Category 4 has no eccentricity

increase in k Region 1 Category 2 has an eccentricity

increase of 0.2% in k Peripheral Categories are modeled eccentric

toward the 3 of 4 zones 73 74 IP2 Spent Fuel Pool Layout Interface Analysis Region 2 Category 5 burnup limits established to force the Category 4 fuel to

be more limiting 75 Interface Analysis Calculated the change in Bias and Uncertainty between Category 4 and

Category 5 fuel and converted to a delta

burnup requirement. (1.3 GWd/MTU in

Region 2) Burnup requirement determined by

matching the full pool k to 3 of 4 infinite k

plus 1.3 GWd/MTUSimilar approach taken for Region 1.

76 Acceptance Criteria Regions done separately since the calculated k will always be determined by

the most reactive region. Region 1 Full pool model with 5 wt% fuel and burnups of 0 (64 IFBA), 21, and 27.7 (28.5-0.8) for Cat

1, 2, and 3, 16 Cat 2 assemblies asymmetric, Cat 3 and 5 pushed in toward Cat 2. K 95/95=0.9697+0.0195 = 0.9892 less than 0.99 77 Arrangement Flexibility 78 SummaryNew pool criticality analysis uses data on the individual assembly depletion to provide Realistic analysisThe analysis also uses full pool analysis to gain benefits from the leakage at the edge of the poolNo credit is taken for absorber panels or cell insertsAlthough realistic analysis is performed, the intent is to have a 1% margin in k to the requirements of 10CFR50.68. (k < 0.99 or 0.94 for accident conditions.)

79 Summary (cont.)Although there are improvements due to the Realistic approach taken, much of the analysis was based on models and depletion analysis similar to that in the submittal that resulted in: "the NRC staff finds that the CSA methodology is acceptable,"

[1] In particular the depletion uncertainty of 5% of the delta k of depletion is still used along with the 1.5% of the Fission Products and Minor Actinide bias

Reference:

1. INDIAN POINT NUCLEAR GENERATING UNIT NO. 2 -STAFF REVIEW OF NETCO REPORT NET-300067-01, "CRITICALITY SAFETY ANALYSIS OF THE INDIAN POINT UNIT 2 SPENT FUEL POOL WITH CREDIT FOR INSERTED NEUTRON ABSORBER PANELS" (CAC NO. MF5282), Nov 23, 2015. ML15292A161 80