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)

Text

Indian Point Energy Center Spent Fuel Pool Management Project July 26, 2017 1

AGENDA Objective - B. Walpole (IPEC)

Overview - G. Delfini (IPEC)

License Amendment Requests - B. Walpole IP2 SFP Criticality Safety Analysis - NETCO 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

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 4

LICENSE AMENDMENT REQUESTS Inter-Unit Fuel Transfer - Submitted to NRC 12/14/16 Currently in Technical Review Process IP2 SFP CSA - Submittal 4th Qtr 2017 Implement in 2019 - Addresses all concerns in the current Operability Evaluation and Eliminates the Non-conservative Technical Specification Inter-Unit Fuel Transfer - Final submittal - 2019 Bounds the remainder of fuel for transfer 5

IP2 SFP Criticality Safety Analysis Overview of New IP2 SFP Criticality Safety Analysis Using No Absorber Panels Dale Lancaster NETCO/NuclearConsultants.com 6

Moving Forward In the August 2013 submittal (NET-300067-01) Flexibility and Margin was emphasized as IP2 was intending to add new Metamic absorber panels It 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 the finite number 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

Realistic Actual irradiation history for all of the assemblies burned at IPEC have been reviewed 1 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, axial 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 cycles But even for these assemblies the loading curve is made more realistic by adding a dependence on the average peaking 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 Categories Category 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 B1.2 = (a1 + a2*E + a3*E2) x exp[-(a4 + a5*E + a6*E2) x C] + a7 + a8*E + a9*E2 B0.8 = (b1 + b2*E + b3*E2) x exp[-(b4 + b5*E + b6*E2) x C] + b7 + b8*E + b9*E2 MRB = B0.8 + (PF - 0.8) x [ B1.2 - B0.8) ] / 0.4 Where:

MRB is the Minimum Required Burnup E is the enrichment C is the cooling time PF Assembly Burnup / (Sum of Cycle Burnups) a1-b9 are fitting coefficients 14

15 Improvements in the Analysis The 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 CSA Depletion Analysis - Realistic Approach Charles T. Rombough CTR Technical Services, Inc.

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Realistic Approach Emphasis is on using actual depletion conditions for discharged fuel while still being conservative The fuel inventory is split into ten batch 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. G thru L (Pyrex, no blanket, Zircaloy guide 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, Zircaloy guide tubes) 20

Moderator Temperature The moderator temperature increases as water is heated traveling upward through the core The moderator temperature is highest at the exit and so the moderator temperature is highest for the top node The core average exit temperature can be calculated from the core average temperature and the inlet temperature T(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: Tin = 538.6 oF, Tout = 607.6 oF Enthalpy at 538.6 oF , 2250 psia = 533.1 btu/lb Enthalpy at 607.6 = 624.0 btu/lb For a peaking factor of 1.40, the enthalpy rise would be 1.40 x (624.0 btu/lb - 533.1 btu/lb) = 127.3 So the peak outlet enthalpy = 533.1btu/lb + 127.3 btu/lb

= 660.4 btu/lb Temperature at 660.4 btu/lb enthalpy = 631.0 oF Density of water at 631.0 oF, 2250 psia = 0.6426 g/cc 22

Moderator Temperature Exit temperature is a function of peaking factor 610 605 600 Linear Fit 595 Points 590 585 580 575 Moderator Temperature (oK) 570 0.5 0.6 0.7 0.8 0.9 1 1.1 1.2 1.3 1.4 1.5 Peaking Factor

Fuel Temperature The fuel temperature will also increase in proportion to the peaking factor 24

Relative Power CTR Technical Services, Inc. 25

Fuel Temperature 800 780 760 Linear Fit 740 Points 720 700 680 660 Fuel Temperature (oK) 640 620 600 0.5 0.6 0.7 0.8 0.9 1 1.1 1.2 1.3 1.4 1.5 Peaking Factor Fuel Temperature at 25 GWd/T in the top node versus radial peaking factor 26

Node Specific Temperatures The 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 node The third node has a lower moderator temperature than the second node but a higher fuel temperature than the second node These 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

Temperatures To simplify the modeling, the most limiting axial burnup profile is selected that maximizes the combined reactivity effect of the moderator and fuel temperatures When all nodes are considered, the reactivity is highest when using the 46+ profile for fuel and moderator temperatures For the 4th and lower nodes, calculations show that it is conservative to use the 3rd node 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 3rd node using the 46+ burnup profile The 3rd node temperatures are then used for the 2nd, 3rd, 4th, 5th, 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 Credit The 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 core Using a large peaking factor for all assemblies is acceptable but very conservative An assembly that has a large peaking factor will also have a larger burnup (by definition) and so would meet the loading criteria An 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 assemblies Depletion 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 assemblies Depletion 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 Absorbers Burnable absorbers harden the spectrum and increase the reactivity of burned fuel Early cycles used Pyrex which displaces more water than WABA Later cycles used WABA in combination with a varying number of IFBA rods The 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.

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Burnable Absorbers The burnable absorbers used during depletion depends on the batch grouping since the absorber characteristics for each batch grouping are known BA Max BA IP-2 Batches Type Max BA Loading Burnup A,B,C,D Pyrex 20 rodlets 18.5 E thru F Pyrex 12 rodlets 12.2 G thru L Pyrex 20 rodlets 16.7 IFBA 116 (1.0x)

M, N, P 28.1 WABA 20 rodlets IFBA 148 (1.5x)

Q, R, S 26.7 WABA 20 rodlets IFBA 148 (1.5x)

T, U, V 33.8 WABA 20 rodlets IFBA 148 (1.25x)

W, X 32.6 WABA 20 rodlets 2A and future IFBA 148 (1.5x) 33.2 fuel and IP-3 WABA 20 rodlets 35

Control Rods Assemblies that were depleted with control rod insertion are more reactive The first 17 cycles at IP2 were operated with D-bank in the bite position For these assemblies, the top node is depleted with a control rod for the entire depletion For 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.

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Other Inserts Other 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 Boron The 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 Xe initial 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 crediting the lower enriched axial blankets, these shapes cannot be used Axial burnup distributions for each blanketed assembly are available from reactor records A 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 renormalizing By 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 well Therefore, 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 node The top 9 nodes are modeled as is while the 9th node from the top is used for all lower nodes (the next 13 nodes down always have a burnup that is higher than the 9th from the top node)

Calculations show that this approach is conservative compared to modeling all nodes for both top peaked and center peaked distributions 44

Special Treatment for Pm-149 Pm-149 is produced during depletion and decays into Sm-149 with a half-life of several days The 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 cycle To account for this effect, the Pm-149 content can be reduced to conservatively account for coastdowns Also, 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 both effects (coast down and low power during the last cycle of depletion), the Pm-149 content for all assemblies is reduced by 75%

47

Depletion Model Depletion models use nominal dimensions Dimension changes that would harden the spectrum during depletion would have the opposite effect in the spent fuel pool For 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 reactivity Although grid growth impacts depletion, it is ignored since it is conservative to ignore it 48

Individual Assembly Depletion Although assemblies are analyzed in groups, assemblies near a reactivity category are analyzed with full actual operating data to find the proper reactivity category.

Assy ID Fuel Type Enrichment FTD Burnable Absorber PPM PF A10 HIPAR 2.21 0.943 None 500 0.92 F44 HIPAR 3.35 0.933 None 495 1.05 L48 LOPAR 3.69 0.944 16 WABA 520 0.69 W52 OFA 4.96 0.946 20 WABA/100 IFBA 880 0.84 X18 OFA 4.95 0.950 20 WABA/116 IFBA 880 0.87 49

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 known Some assemblies are analyzed individually if near a reactivity category Peaking 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 Analysis Dale 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 oC is 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

Validation Changes (cont.)

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Validation Changes (cont.)

LCT-08 analysis has been improved This set was done at B&W and the same facility as the B&W-1810 needed for Ag-In-Cd LCT-08 benchmark sample runs in the criticality handbook used a 2D model Revised to a 3D model (included fuel rods above the water)

Increased average k by about 0.07% in k.

All pertinent changes from RAIs on the previous IP2 submittal (ML15261A528) have been incorporated 57

Multi-Misload Analysis IP2 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 locations are 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 all water holes but the rest of the fuel at the Tech Spec requirements. (Calculated k95/95=0.939 with 2000 ppm using unborated uncertainties) 58

Clad Creep Clad 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 59

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.

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Creep Model Basis

4. Y. Irisa, et al, Segmented Fuel Rod Irradiation Program On Advanced Materials For High Burnup, 61 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

1. G.P. SABOL, G. SCHOENBERGER, and M.G. BALFOUR, IMPROVED PWR FUEL CLADDING, Materials 62 for Advanced Water Cooled Reactors, Proceedings of a Technical Committee Meeting, Plzen, Czechoslovakia, May 14-17, 1991, IAEA-TECDOC-665, IAEA, VIENNA, 1992.

3. Garzarolli, F., Manzel, R., Reschke, S., and Tenckhoff, E., "Review of Corrosion and Dimensional Behavior of Zircaloy 63 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 64

5. 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 Assume returns to initial density by 30 GWd/MTU

7. 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 65 City, Utah, April 10-13, 2000, American Nuclear Society, La Grange Park, Illinois.

Oxide Layer Growth

6. David Mitchell, Anand Garde, and Dennis Davis, Optimized ZIRLOTM Fuel Performance in Westinghouse PWRs, 66 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.

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Grid Growth Basis David Mitchell, Anand Garde, and Dennis Davis, Optimized ZIRLOTM Fuel Performance in Westinghouse 69 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.)

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

Grid Growth Basis (cont.)

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

Grid Growth Summary Grid growth is temperature dependent. Therefore, the higher grids experience more growth.

The Inconel top grid grows much less A uniform grid growth of two adjacent assemblies by 0.69% (Westinghouse 15x15) would make the grids touch Model assumes 0.78% growth at 60.5 GWd/MTU (highest credited burnup).

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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

IP2 Spent Fuel Pool Layout 74

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/MTU Similar approach taken for Region 1.

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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.

K95/95 =0.9697+0.0195 = 0.9892 less than 0.99 77

Arrangement Flexibility 78

Summary New pool criticality analysis uses data on the individual assembly depletion to provide Realistic analysis The analysis also uses full pool analysis to gain benefits from the leakage at the edge of the pool No credit is taken for absorber panels or cell inserts Although 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.)

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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