NL-19-023, Response to Request for Additional Information (RAI) - Proposed License Amendment Request Regarding Storage of Fresh and Spent Nuclear Fuel in the Spent Fuel Pool

Response to Request for Additional Information (RAI) - Proposed License Amendment Request Regarding Storage of Fresh and Spent Nuclear Fuel in the Spent Fuel Pool

ML19157A309
Person / Time
Site:	Indian Point
Issue date:	06/06/2019
From:	Gaston R Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
EPID L-2017-LLA-0408, NL-19-023
Download: ML19157A309 (114)

Text

~ Entergx Entergy Nuclear Operations, Inc.

1340 Echelon Parkway Jackson, MS 39213 NL-19-023 June 6, 2019 Tel 601-368-5138 Ron Gaston Director, Nuclear Licensing 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

References:

Response to Request for Additional Information (RAI) - Proposed License Amendment Request Regarding Storage of Fresh and Spent Nuclear Fuel in the Spent Fuel Pool Indian Point Nuclear Generating Unit No. 2 NRC Docket No. 50-247 Renewed Facility Operating License No. DPR-26

1) Entergy Nuclear Operations, Inc. (Entergy) letter to U.S. Nuclear Regulatory Commission (NRC), "Indian Point Nuclear Generating Unit No. 2 - Proposed License Amendment Regarding Spent Fuel Storage,"

dated December 11, 2017 (Letter No. NL-17-144) (ADAMS Accession No. ML17354A014)

2) NRC Email to Entergy, "Indian Point Unit 2 - Request for Additional Information - Proposed License Amendment Request Regarding Storage of Fresh and Spent Nuclear Fuel in the Spent Fuel Pool (EPID:

L-2017-LLA-0408)," dated February 14, 2019 (Letter No. RA-19-016)

(ADAMS Accession No. ML19045A625)

In accordance with Title 1 O of the Code of Federal Regulations (1 O CFR) 50.90, Entergy Nuclear Operations, Inc., (Entergy) submitted a License Amendment Request (LAA) for Indian Point Nuclear Generating Unit No. 2 (IP2) to the U.S. Nuclear Regulatory Commission (NRC)

(Reference 1) to revise Technical Specification (TS) Limiting Condition for Operation (LCO) 3.7.13, "Spent Fuel Storage," and TS Design Features Section 4.3, "Fuel Storage." These proposed changes resolve a non-conservative I P2 TS associated with TS LCO 3. 7.13, and negate the need for the associated compensatory measures while taking no credit for the installed Boraflex panels.

NL-19-023 Page 2 of 3 The purpose of this letter is to provide the final Entergy responses to an NRC request for additional information (RAI) (Reference 2) regarding the IP2 LAR to revise the spent fuel storage TSs. This letter also includes the final TS LCO 3. 7.13 and TS Section 4.3 changes. A draft of the Entergy responses to the RAI and proposed TS changes had previously been provided to the NRC for review in preparation for a planned audit of Curtiss-Wright Nuclear Division, Northeast Technology Corporation (NETCO) (i.e., the preparer of the analytical report used as the basis for the TS changes). Considering the information provided by Entergy, the NRC later cancelled the NETCO audit. Note that this letter includes a minor clarification that was not in the previously provided draft RAI responses and TS changes. The added text is indicated by marginal revision bars in the response to RAI 1.g and the markup of Note (c) to TS Table 3.7.13-1.

The Enclosure to this letter provides the text of each of the NRC RAI questions provided in the Reference 2 email followed by the corresponding Entergy responses. Attachments 1 and 2 to the Enclosure provide the IP2 TS and TS Bases markups, respectively. Attachment 3 to the Enclosure provides the clean retyped TS pages with marginal revision bars indicating the changes. The TS pages provided in this letter replace the TS pages previously provided in Reference 1 in their entirety.

Entergy has determined that the supplemental information provided in this letter does not alter the conclusion reached in the original LAR (Reference 1) that the proposed change presents no significant hazards consideration under the standards set forth in 1 O CFR 50.92(c). The supplemental information also does not alter the original LAR's bases for concluding that, in accordance with 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with issuance of the amendment.

In accordance with 10 CFR 50.91 (b), a copy of this request and the associated Enclosure and Attachments are being submitted to the designated New York State official.

If you have any questions or require additional information, please contact Mr. Robert Walpole, Manager, Regulatory Assurance, at 914-254-6710.

There are no new regulatory commitments made in this letter.

I declare under penalty of perjury that the foregoing is true and correct. Executed on June 6, 2019.

Sincerely,

~~;6 Ron Gaston RWG/cdm

NL-19-023 Page 3 of 3

Enclosure:

Response to Request for Additional Information, Entergy Nuclear Operations, Inc., Indian Point Nuclear Generating Unit No. 2, Proposed License Amendment Regarding Spent Fuel Storage, Docket No. 50-247 Attachments to

Enclosure:

1. Technical Specification Page Markups

2. Technical Specification Bases Page Markups (Information Only)

3. Retyped Technical Specification Pages cc:

NRC Senior Project Manager, NRC NRR DORL Regional Administrator, NRC Region I NRC Senior Resident Inspector, Indian Point Energy Center President and CEO, NYSERDA New York State (NYS) Public Service Commission

Enclosure NL-19-023 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NO. 2 PROPOSED LICENSE AMENDMENT REGARDING SPENT FUEL STORAGE DOCKET NO. 50-247

Enclosure NL-19-023 Page 1 of 54 By letter dated December 11, 2017, Entergy Nuclear Operations, Inc. (Entergy), the licensee for Indian Point Unit 2 (IP2), submitted a license amendment request (LAR) proposing to revise the Technical Specification for IP2 spent fuel storage. On the basis of the provided information, the NRC has determined that additional information is needed to assess whether the licensee has acceptably demonstrated that the spent fuel pool will remain subcritical under all conditions and to complete its review. These requests for additional information (RAls) and Entergy's responses are as follows:

The licensee's analysis cites J.C. Wagner, "Impact of Soluble Boron Modeling for PWR Burnup Credit Criticality Safety Analyses," Trans. Am. Nucl. Soc., 89, pp. 120 (2003) for its use of a constant soluble boron concentration rather than time-dependent soluble boron letdown curve for its depletion calculations. However, the licensee is not using the constant soluble boron concentration in a manner consistent with the reference. The reference is based on the fuel completing three full cycles of burnup and not an individual fuel assembly's lifetime burnup averaged soluble boron. An individual fuel assembly under the licensee's control may not actually complete three full cycles of burnup, which could lead to a non-conservatism. With respect to using the individual fuel assembly's lifetime burnup averaged soluble boron provide the following information to address the potential non-conservatism:

This RA/ raises two issues, 1) the validity of using an average soluble boron concentration rather than a soluble boron letdown curve, and 2) the determination of an individual assembly's averaged soluble boron. Parts a through g only address the determination of the correct average ppm for an assembly. Following the response to part g is additional analysis showing that using a constant average soluble boron concentration (ppm) rather than a time dependent letdown curve is appropriate.

Before discussing the details, it is important to have perspective on the sensitivity of the soluble boron concentration used in the depletion analysis. NUREGICR-6665 (repeated in NUREG/CR-6800) provides a sensitivity of 3-3.5 pcm/ppm. Methods employed in the current analysis result in a sensitivity of 1-2 pcm/ppm. NET-28091-003-01, Rev. O reports calculated reactivity values to O.OOOUk which is equivalent to 10 pcm. Using 2 pcm/ppm, it would require a difference in the average soluble boron concentration during depletion of 5 ppm to exceed the round-off and 50 ppm to achieve the minimally significant change of 0.001 ink. Nominal estimates for the boron concentration during depletion are used for NET-28091-003-01, and justification for this is given in the following responses.

a. Describe how the "... multi-cycle burnup averaged soluble boron concentration... " is calculated.

Response to RAI 1.a The average soluble boron concentration for a particular assembly is determined by:

rEOL BU ppm(BU)dBU ppmave = Jo EOL BU (Equation 1)

Enclosure NL-19-023 Page 2 of 54 where:

BU is the assembly burnup and EOL BU is the end of life burnup of the assembly.

Equation 1 is numerically solved by using the assembly burnup and core soluble boron at discrete burnup steps (about 1000 or 2000 MWd/T). The numerical solution uses the trapezoidal method where the average soluble boron is used for each burnup step.

Table 1.1 shows the calculation for one of the limiting assemblies of Batch L (the depletion analysis uses 660 ppm). The data used in the analysis is the predicted soluble boron versus burnup values from the fuel vendor core design reports. Measured data can also be used, but the calculated average soluble boron has a low sensitivity to small variations in soluble boron and therefore using either measured or predicted data does not produce an appreciable difference. For the example presented below, the chosen assembly was in the core only two cycles. Both cycles were shut down early (208 ppm and 255 ppm). This example shows how the calculated average soluble boron incorporates an early shutdown.

Table 1. 1: Demonstration of the Calculation of the Burnup Weighted Average Soluble Boron Concentration (ppmave)

Assembly Core Average ppm L1BU over the time ppm*L1BU Burnup (MWclff) ppm step 0

1245 178 947 178 1096 195088 1189 881 1011 914 924054 2395 824 1206 852.5 1028115 4853 702 2458 763 1875454 7369 573 2516 637.5 1603950 9931 442 2562 507.5 1300215 12521 302 2590 372 963480 14401 208 1880 255 479400 14401 1421 0

814.5 0

14582 1114 181 1267.5 229417.5 15519 1050 937 1082 1013834 16618 1000 1099 1025 1126475 18799 891 2181 945.5 2062135.5 20958 777 2159 834 1800606 20958 750 0

763.5 0

21653 715 695 732.5 509087.5 23096 629 1443 672 969696 25218 494 2122 561.5 1191503 27323 350 2105 422 888310 28844 255 1521 302.5 460102.5 Sum 18620923.0 ppmave (sum/total 646 BU)

Enclosure NL-19-023 Page 3 of 54

b. Were any of the IP2 or IP3 cycles stopped appreciably short of reaching O ppm of soluble boron in the reactor?

Response to RAI 1.b The majority of the IP2 and IP3 cycles ended with HFP critical soluble boron concentrations less than 20 ppm. However, a few cycles ended with the HFP critical soluble boron concentration greater than 20 ppm. Table 1.2 lists the short cycles and the end of cycle HFP critical soluble boron concentration (ppm).

Table 1.2: End of Cycle Soluble Boron (ppm) for Cycles which Shutdown Prior to Hot Full Power Critical Boron of less than 20 ppm EOCHFP Unit Cycle Critical Boron (ppm) 2 2

76 2

4 156 2

9 208 2

10 255 2

13 99 2

14 300 2

15 32 2

20 137 3

5 43 3

8 81

c. Where were the cycle average soluble boron concentrations for the cycles listed in NET-28091-0003-01 Table 5.9?

Response to RAI 1.c Table 5.9 of NET-28091-0003-01 provides the soluble boron assumed in the depletion analysis. Tables 3.5 and 3.6 provide the actual average soluble boron concentrations for each cycle. The burnup weighted average soluble boron for each cycle is calculated using Equation 1 but using the core burnup rather than the assembly burnup. The actual end of cycle burnup is used for the last burnup step, however all of the intermediate burnup steps were based on predicted burnup data from the fuel vendor supplied nuclear design reports.

Due to the difficulty of evaluating Equation 1 for each assembly, a simplifying assumption is made. It is assumed that the relative power of the assemblies remain the same throughout the cycle. If the assumption is valid, then the soluble boron concentration (ppmave) for each assembly simplifies to the following equation:

Enclosure NL-19-023 Page 4 of 54 ppmaveapproximation = Sum(CycleBurnup;

• Cyc/eAveragePPM)!Sum(CycleBurnup;)

Eq 2.

where: CycleBurnupi is the cycle burnup of the assembly in cycle i; CycleA veragePPM is the solution of equation 1 with core average parameters; and ppmave approximation is an approximation of the assembly average ppm Although Equation 2 is a good approximation, the change in the relative power distribution of an assembly can result in some assemblies achieving more burnup during the beginning of cycle (high boron concentration) than at the end of cycle (low boron concentration).

NET-28091-0003-01, Rev. O actually uses Equation 2 rather than the more exact Equation 1(that takes into account the assembly relative power variation). Table 1.3 below shows the maximum Equation 2 calculated soluble boron concentration (ppmave) and the soluble boron value used in the depletion calculations for each fuel batch. The cycles marked in yellow are close enough to the soluble boron used in the depletion calculations(< 20 ppm) that considering the change in relative power (Equation 1) could result in a slightly higher assembly specific burnup weighted average boron concentration than Equation 2.

Batches Wand X are relatively recent discharges and so electronic data for the predicted power distribution is available. Equation 1 is used to calculate the assembly specific burnup weighted average boron concentration (ppmave) for all of the Batch Wand X assemblies.

These values are shown below on Table 1.4. The maximum value in Table 1.4 is 12 ppm above the value assumed in the depletion analysis (880 ppm). In order to accommodate this slightly higher soluble boron content than the soluble boron assumed in the depletion analysis, the date for the cooling time is updated. The categorization assumes a cooling time reference date of 6/30/2018. The updated analysis assumes a cooling time reference date of 1/30/2019. The reactivity decrease resulting from the longer cooling time is more than a factor of 5 greater than the reactivity effect of the slightly lower soluble boron concentration assumed in the depletion analysis. Therefore, no categorization change is necessary nor is there any reduction in the margin.

Enclosure NL-19-023 Page 5 of 54 Table 1.3: Maximum Assembly ppmave Using Equation 2 Batch Maxppmave Depletion Using Equation 2 Assumption A

546 570 B

559 570 C

559 570 0

551 570 E

545 660 F

533 660 G

<513 660 H

<513 660 J

535 660 K

611 660 L

654 660 M

785 820 N

717 820 p

789 820 Q

837 850 R

844 850 s

<814 850 T

<814 880 u

<848 880 V

835 880 w

865 880 X

880 880 Rest of

<891 950 Batches Note: Highlighted Reqwred lndlVldual Assembly Review For older fuel assemblies, the effect of changing the cooling time reference date is much less, so for batches B, C, 0, L, Q, and R the margin for each assembly to the burnup requirements is examined. A 20 ppm change in the depletion assumption changes the burnup requirement by about 50 MWd/T (using the data given on Table 8.27 of the CSA).

The minimum excess burnup (actual burnup minus the required burnup for the category selected) is 516, 45, 165, 615, 107, and 660 for Batches B, C, 0, L, Q, and R, respectively.

Therefore, the excess burnup is sufficient to offset the reactivity effect due to any small increase in the soluble boron content used in the depletion analysis for all but Batch C.

However, only one Batch C assembly (C29) has a small burnup margin. The rest of Batch C fuel assemblies have over 1000 MWd/T excess burn up to the burn up requirement. The Equation 2 calculated soluble boron concentration for assembly C29 is 542 ppm which is 28 ppm less than the soluble boron used in the depletion analysis. In summary, crediting the available excess burnup offsets any slightly lower soluble boron used in the depletion analysis for all assemblies. No change in categorization is necessary and there is no impact on the margin claimed (1% LJk).

Enclosure NL-19-023 Page 6 of 54 Table 1.4: Assembly Specific Weighted Average Soluble Boron Concentration for Batches Wand X Using Equation 1 ID PPfflave ID PPfflave ID PPfflave ID IJDfflave ID IJDfflave W01 878 W41 873 W81 861 X28 880 X68 874 W02 854 W42 873 W82 872 X29 881 X69 889 W03 853 W43 873 W83 861 X30 880 X70 878 W04 878 W44 880 W84 856 X31 881 X71 880 was 879 W45 851 W85 879 X32 883 X72 873 W06 853 W46 865 W86 877 X33 881 X73 878 W07 853 W47 854 W87 861 X34 881 X74 889 W08 877 W48 867 W88 861 X35 882 X75 880 W09 849 W49 867 W89 871 X36 881 X76 880 W10 848 wso 854 W90 861 X37 890 X77 878 W11 847 W51 867 W91 878 X38 890 X78 889 W12 848 W52 854 W92 861 X39 881 X79 873 W13 849 W53 854 W93 872 X40 881 X80 888 W14 848 W54 854 X01 885 X41 888 X81 888 W15 848 wss 853 X02 885 X42 881 X82 889 W16 849 W56 867 X03 888 X43 880 X83 880 W17 849 W57 866 X04 872 X44 874 X84 880 W18 849 W58 866 xos 890 X45 875 X85 873 W19 847 W59 854 X06 881 X46 880 X86 880 W20 849 W60 853 X07 881 X47 881 X87 888 W21 872 W61 867 X08 881 X48 883 X88 888 W22 882 W62 856 X09 876 X49 880 X89 873 W23 873 W63 878 X10 881 xso 889 X90 888 W24 880 W64 872 X11 874 X51 878 X91 877 W25 873 W65 878 X12 875 X52 878 X92 874 W26 880 W66 862 X13 881 X53 880 X93 889 W27 873 W67 861 X14 874 X54 889 X94 880 W28 853 W68 856 X15 892 xss 888 X95 880 W29 879 W69 861 X16 876 X56 878 X96 877 W30 853 W70 861 X17 890 X57 873 W31 851 W71 872 X18 875 X58 881 W32 853 W72 855 X19 888 X59 889 W33 853 W73 861 X20 892 X60 878 W34 880 W74 878 X21 881 X61 878 W35 874 W75 871 X22 883 X62 878 W36 882 W76 872 X23 882 X63 888 W37 877 W77 872 X24 882 X64 880 W38 851 W78 878 X25 882 X65 873 W39 874 W79 879 X26 883 X66 888 W40 851 W80 861 X27 881 X67 878

Enclosure NL-19-023 Page 7 of 54 Finally, for IP3 Batches A through U, the depletion analysis was performed using 560 ppm.

Batches T and U exceed this by about 20 ppm. All Batch T and U assemblies exceed the burnup requirement by more than 4500 MWd/T except T53 (930 MWd/T margin) and U12, U21, U31, and U41 (1200 MWd/T margin). The minimum margin is 930 MWd/T which is much larger than the 50 MWD/T required to offset a 20 ppm reduction in boron concentration.

d. How is the EOC burnup for each assembly determined? What does it represent? How does the EOC burnup compare to the cycle burnup.

Response to RAI 1.d The EOG burnup for each assembly is determined by the utility by time integration to the end of cycle of the reactor power and relative power distribution to calculate the burnup (in GWd), divided by the mass of uranium (in metric tons (T) of initial uranium) in the assembly.

The simplified (Equation 2) calculation of the soluble boron concentration (ppmave) uses the utility measured EOG burnups (not the predicted EOG burnups found in the design reports).

The cycle burnup is the average change in assembly burnups from the beginning to the end of the cycle.

e. How is the EOC burnup used to weight the cycle average PPM?

Response to RAI 1.e As stated in the response to RAJ 1.c, the cycle average soluble boron is determined using Equation 1 above, using the core burnup rather than the assembly burnup. These cycle average soluble boron contents are then used in Equation 2 to derive an approximate assembly average soluble boron. Note in this case, the EOG burnup is the core burnup not the assembly burnup. The assembly cycle average soluble boron is not used in this criticality safety analysis. This method only uses the assembly average soluble boron over the life of the assembly. Historical techniques have compared the core average soluble boron for every cycle to the depletion assumptions. That approach is overly conservative since it does not credit the impact of low soluble boron cycles on the assembly reactivity.

When there is insufficient margin between the depletion assumed soluble boron content and the Equation 2 approximated soluble boron, the more correct Equation 1 is used. With the Equation 1 approach, the assembly EOG burnup is used to precisely end the portion of the integration for each cycle.

f.

How is that used to determine whether or not a fuel assembly exceeds the soluble boron concentration used in the depletion analysis?

Enclosure NL-19-023 Page 8 of 54 Response to RAI 1.f See responses to 1.a through 1.e.

g. NET-28091-0003-01 Section 8.15, Burnup Penalty for High Soluble Boron Conditions, attempts to address the potential for an early cycle shutdown. However it is not clear what was modeled in the analysis. Provide the description of what was modeled.

Response to RAI 1.g Table 8.27 of NET-28091-0003-01 shows the calculations performed to support the burnup penalties (0.2, 0.3, 0.6, and 0.9 GWd/T for Categories 2, 3, 4, and 5, respectively). Since the effect of higher soluble boron increases with burnup, only the highest burnup requirement for each category requires calculation. The burnups prior to adding the penalty given on Table 8.27 are the maximum burnup requirements for each category (Z fuel at 5.0 w/o, no cooling, 1.2 peaking factor). The single high burnup penalty for each category applies to all enrichments, peaking factors, and cooling times. Table 8.27 verifies that the burnup penalties are adequate since the 1200 ppm case analyzed with the penalty is always less reactive than the reference case (no penalty) analyzed at 950 ppm.

Based on this RA/ response. Table 3.7.13-1 Note (c) of the Proposed Technical Specification Changes has been clarified for the use of the burnup penalties based on the soluble boron conditions.

Comments Regarding Applicability of Wagner Paper As the NRG raised concerns in the preamble to RAJ 1 regarding the applicability of the J.C.

Wagner paper, the following comments are provided in addition to the subparts of RA/ 1 to further justify the applicability and appropriateness of using the method in NET-28091-0003-

01. The J. C. Wagner, "Impact of Soluble Boron Modeling for PWR Burnup Credit Criticality Safety Analyses," Trans. Am. Nucl. Soc., 89, pp. 120 (2003) paper studies the effect of using an average soluble boron in the depletion calculations compared to using the boron letdown curve. The Wagner paper concludes:

"Because the constant boron modeling has significantly more boron present late in burnup, it results in slightly more reactive isotopic compositions for discharged SNF.

Therefore, it is a conservative modeling approximation to employ in burnup credit evaluations. "

Analysis confirms that this conclusion applies to the methods used in NET-28091-0003-01.

Wagner's paper uses a soluble boron letdown curve given in Figure 1 of his paper. Figure

1. 1 below shows what the average soluble boron content is for an assembly depleted using Wagner's Figure 1 letdown curve. For this confirmation analysis, depletion is performed using the Wagner boron letdown curve and depletion cases were run for each burnup (in GWd/T) using the average soluble boron shown on Figure 1.1 (45 depletion cases plus the one Wagner boron letdown case). Using the atom densities from these depletion cases, a 2x2 model of Region 2 with a 3 out of 4 /oading was used to determine the k values (45

Enclosure NL-19-023 Page 9 of 54 calculations of k with the average soluble boron content and 45 calculations of k using the letdown curve). The results are shown on Figure 1.2 below.

The following provides more details on the models. Wagner used 4. O wt% fuel, so for this analysis 4.0 wt% is also used. The depletion assumptions used were the same as used for Batch Grouping M, N, P (except for the soluble boron content and removal of the WABA).

For the Batch Grouping M, N, and P, the burnable absorbers modeled for depletion were 20 WABA rodlets removed at 28.1 GWd/T and 116 IFBA rods. The WABA was not removed for this study. The peaking factor of 1.4 was used in order to set the temperatures. Batch Grouping M, N, and P was selected since it best matched the 4 wt% enrichment selected by Wagner.

The Wagner paper used 4 different average soluble boron values so at nearly all burnups there is not agreement between the average soluble boron content and the boron letdown curve. For this analysis the average soluble boron content used matches the letdown curve.

Figure 1.2 shows that there is almost no difference in the final k between using the correct average soluble boron content and using a boron letdown curve. It is slightly conservative to use the average soluble boron content in that if they were identical, one would expect the Monte Carlo variation to produce higher k values half the time for each method. In this analysis only one point (at 38 GWd/T) produced a higher k (within expected Monte Carlo uncertainty) when using the boron letdown curve. Most of the differences were close to the Monte Carlo uncertainty but a few were a little larger (up to 200 pcm).

1400 1200 1000 E

C.

C. 800 QI bO Ill 600 QI >

ct 400 200 0

0 10 20 30 40 50 Burnup (GWd/T)

Figure 1.1: Average ppm which matches 3 cycle letdown curves (1200 to O ppm every 15 GWdff)

Enclosure NL-19-023 Page 10 of 54 1.25 1.2 1.15 0.95 0.9 0

10 15 Letdown Soluble Boron Average Soluble Boron 20 25 30 35 40 Bumup (GWd/T)

Figure 1.2: Region 2 calculated k using a letdown curve or the average ppm 45 It is important to note that the method used in NET-28091-0003-01 uses the end-of-cycle soluble boron concentration from the actual operation. Therefore, if a cycle is shut down early, the assembly has a higher average soluble boron. For assemblies which have been categorized, the correct burnup-weighted average soluble boron concentration is used (see earlier part of this response). For assemblies discharged after the categorization, Indian Point must confirm the burnup-weighted average soluble boron content is less than 950 ppm or use the bias described in the response to part g (NET-28091-0003-01 Section 8.15).

Enclosure NL-19-023 Page 11 of 54 In Attachment 1, "Analysis of Proposed Technical Specification Changes, Section 4.6 Configuration Modeling" sub-section "Normal Conditions, Item B Movement of Fuel" states:

"Section 9.1 of the CSA addresses movement of fuel assemblies in and around the pool, fuel inspection, and fuel reconstitution. An assembly is isolated if there is 20 cm of water between assemblies. There are two locations where it would be possible to place two assemblies within 20 cm of each other outside of the rack: when a fuel assembly is in the fuel elevator and another fuel assembly is vertical in the upender. However, procedures will not permit moving an assembly within 25 cm of either of those locations when another assembly is in either the fuel elevator or upender." With respect to these restrictions provide the following information:

a. How procedures will prevent operators from moving an assembly within 25 cm of the fuel elevator or upender or within 20 cm of the storage racks.

Response to RAI 2.a The following statement will be incorporated into the Precautions and Limitations Section and precede the fuel movement Section in the body of the relevant procedures.

During any fuel handling, fuel being moved by the Spent Fuel Handling Machine SHALL NOT be moved to within 25 cm (approx. 10 inches) of the new fuel elevator or the SFP upender whenever there is fuel contained in either. In addition, fuel SHALL NOT be moved in the cask load pit to within 20 cm (approx. 8 inches) of the storage racks on the north and east sides of the pit.

NOTE:

As a reference to ensure the above distances are maintained, the dimension of a fuel assembly top nozzle is 8.404 inches square. A two fuel assembly spacing would ensure proper separation in all cases.

The procedures to be updated are as follows:

2-SOP-17. 12, Spent Fuel Handling Machine And Spent Fuel Pit Operations 2-DCS-008-GEN, Unit 2 MPG Loading and Sealing Operations 2-DCS-012-GEN,Unit 2 Unloading Operations 2-FTR-001-GEN, Unit 2 STC Unloading Operations 2-REF-003-GEN, Reactor Core Refueling O-NF-203, Internal Transfer Of Fuel Assemblies And Inserts

b. Describe whether there are physical barriers that prevent moving an assembly within 25 cm of the fuel elevator or upender or within 20 cm of the storage racks.

Enclosure NL-19-023 Page 12 of 54 Response to RAI 2.b While there are no physical barriers that prevent moving an assembly within 25 cm of the fuel elevator or upender or within 20 cm of the storage racks, there are physical visual aids (e.g. fuel rack indexing system, fuel rack encoder readings, piping, gates, etc.) that will provide the fuel handling operator situational awareness and identify boundaries. In addition, the experience of the Fuel Handling Operator and the presence of the Spotter and Fuel Handling Supervisor will ensure the restrictions are maintained.

c. How fuel handling operators will know the distance between a fuel assembly being moved and the fuel elevator, upender, or the storage racks.

Response to RAI 2.c The new NOTE that will appear in all applicable procedures (see response to RA/ 2.a),

along with the use of the Spent Fuel Pool indexing system, will provide sufficient administrative and visual guidance to assist the fuel handling operator in maintaining the required distance.

The proposed new technical specifications are listed in Attachment 2. With respect to this proposed TS provide the following information:

a. Note 14 of Insert 3 states, "The edge of the spent fuel rack can be considered as a row of Water Holes." The analysis in NET-28091-0003-01 models the "edge" is of the spent fuel pool. Clarify whether the "edge" as used in the proposed TS could include the "edge" of the various racks within the spent fuel pool.

Response to RAI 3.a The "edge" was meant to be the pool wall or the cask loading area, not the edge of a rack module or region. Note 14 will be revised as follows: The edge of Region 2 next to the pool wall or cask loading area can be considered to be a row of Water Holes.

b.

Proposed TS 4.3.1.1 is not consistent with the standard technical specifications. Provide the justification for deviating from the standard technical specifications.

Enclosure NL-19-023 Page 13 of 54 Response to RAI 3.b The Standard Technical Specification (STS) contained in NUREG-1431, Volume 1 for Section 4.3.1.1 are based on a storage approach of high and low density storage racks, with the high-density storage racks containing burnup versus enrichment requirements. Given the approach taken in the criticality safety analysis, the deviations from the STS are warranted to ensure that the technical specifications for the Indian Point Unit 2 SFP are representative of the pool configuration justified in the supporting analysis and that the regulatory requirements of 10 CFR 50.68 are satisfied.

The following statement will be incorporated into Section 4.3.1.1 of the proposed Technical Specification:

The spent fuel storage racks are designed and shall be maintained with:

a. Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent and poisons, if necessary, to meet the limit for kett,

b. kett S 0.95 when flooded with borated water, kett < 1.0 if fully flooded with unborated water, and

c. Each fuel assembly categorized based on initial enrichment, burnup, cooling time, averaged assembly peaking factors, and number of Integral Fuel Burnable Absorbers (IFBA) rods with individual fuel assembly storage location within the spent fuel storage rack restricted as required by Technical Specification 3.7.13.

NET-28091-0003-01, Revision 0, Section 3.4, "Plant Operation Data," defines a new term "Tave*" For the purposes of the analysis in NET-28091-0003-01 "Tave" is defined as "... the average moderator temperature in the active fuel (not the vessel average temperature)." "Tave" is used to determine the moderator and fuel temperature profiles used in the depletion portion of the analysis. The moderator and fuel temperature the fuel experiences during depletion can have a significant impact on the post irradiation reactivity of the fuel. Therefore the moderator and fuel temperature must be modeled appropriately. With respect to the determination and use of "Tave" provide the following information:

a.

Describe how "Tave" is calculated and/or measured.

Response to RAI 4.a Tave is provided by the core designer (Westinghouse) in each cycle's core design report (NOR or NuPOP). It is the average moderator temperature in the active fuel. The Tave from the NOR is a best estimate design value at which the core is expected to operate. The cycle specific values for Tave from the NORs are shown on Tables 3.5 and 3.6 of the CSA (NET-28091-0003-01). Tave is calculated (by Westinghouse) via a heat balance using the 1iniet, Power Rating (MWt of the core), the System Pressure, the thermal properties of water, the vessel flow rate (gpm), and the percent bypass flow. The Tave used for the criticality analysis is the same Tave as used in the core design.

Enclosure NL-19-023 Page 14 of 54 For clarification, the cycle dependent Tave is used in the CSA to determine node dependent moderator temperatures. Tave has an insignificant impact on the fuel temperature which is controlled by the linear heat generation rate (input through the peaking factor) and the thermal conductivity of the gap and fuel which changes with burnup. The node dependent fuel temperatures are determined based on a limiting (highest power cycle) fuel temperature analysis from Cycle 21.

b.

Describe any biases or uncertainties in "Tave" and how they are included in the analysis.

Response to RAI 4.b Tave is a nominal, best estimate value. There is no known bias in the parameters used by Westinghouse in the calculation of Tave* The uncertainty in Tave has an insignificant effect on the total uncertainty and can be ignored. For example, a 2% power uncertainty produces about a 1 ° F uncertainty in Tave and would cause a 1 °F uncertainty in the nodal temperatures (L1T across the core is between 60 and 70 °F). This translates into a L1k of 0.0003 (the sensitivity is 0. 0052 L1k for a 1 O K difference in moderator temperature during depletion -

see Section 5. 1.5 of the CSA). Statistically combining this small effect with all the other uncertainties would not change the total uncertainty (see Tables 7.3, 7.4, and 7.5).

RAIS The licensee's analysis and TS use the term 'averaged assembly peaking factor' when determining depletion parameters and the minimum assembly average burnup required for several storage categories. "The averaged assembly peaking factor is the assembly burnup divided by the sum of the cycle burnups for the cycles the assembly was in the core." The licensee's use of the 'averaged assembly peaking factor' is an attempt to take credit for a fuel assembly that isn't at its maximum fuel and moderator temperature for its entire life. However, the licensee has not shown that using the 'averaged assembly peaking factor' is an appropriate or conservative means of capturing the post irradiated reactivity of a given fuel assembly or set of assemblies, especially if the more limiting conditions occur closer to the time when the fuel assembly is stored in the SFP. As the 'averaged assembly peaking factor' hasn't previously been shown to be an acceptable means of determining depletion parameters and the minimum assembly average burnup required for several storage categories provide the following information:

a. Evidence that using the 'averaged assembly peaking factor' is an appropriate or conservative means of capturing the post irradiated reactivity of a given fuel assembly or set of assemblies, especially if the more limiting conditions occur immediately before storage in the SFP.

Enclosure NL-19-023 Page 15 of 54 Response to RAI 5.a This CSA analyzes differing peaking factors as a function of burnup as discussed at the end of Section 5.1.5 of NET-28091-003-01. The analysis demonstrates that when the more limiting peaking factor is at the beginning of irradiation, using the average peaking factor is conservative. Since the reactivity of assemblies decreases with burnup (except early in a cycle with burnable absorbers) the limiting peaking factor is generally during the first cycle of irradiation. After reviewing the spent fuel inventory, there are no assemblies with burnups near the burnup requirements at Indian Point that have the most limiting conditions at end of life. Assemblies with significant burnup are often placed in the center position of the core but these assemblies greatly exceed the burnup requirements for the given enrichment.

The RA/ specifically addresses high power at end of life. A high end of life power would have a high equilibrium Pm-149 concentration which then decays to Sm-149. Since Sm-149 is a strong absorber, NET-28091-0003-01 assumes a 50% reduction in the peaking factor at end of life. NET-28091-0003-01 also assumes a coast down where the Pm-149 reaches equilibrium at 50% of the previously reduced power. The impact of these two assumptions is 250 pcm (see the 2nd paragraph of Section 5.8 of NET-28901-0003-01). Additional analysis has been performed where the peaking factors go up rather than down. The assembly was depleted at a peaking factor of 0.8 and 1.2 for the first and second halves of the burnup, respectively. Assuming the latter high peaking factor (1.2) decreased by 20% to a peaking factor of 1.0 at end of life in addition to the 50% coast down, the impact on k is slightly non-conservative by 150 pcm. To summarize, increasing peaking factor with burnup is rare. As stated in the first paragraph, the current inventory at in the Indian Point pool has been reviewed, and all assemblies with increasing peaking factors have significant margin to cover this 150 pcm. Even if a future assembly is loaded with increasing peaking factors, the impact on k is small relative to the margin to the 10 CFR 50.68 criteria.

b. Describe any biases and uncertainties associated with using 'averaged assembly peaking factor' and how they are included in the analysis.

Response to RAI 5.b Since the radial peaking factors are normalized there is no bias by definition. The uncertainty in the assembly average peaking factor is the same as the radial assembly power distribution uncertainty (-2%). The average assembly peaking factor is only used to determine the moderator and fuel temperatures used during depletion. From the loading curve for Z fuel, the maximum t1burnup between PF=0.8 and PF=1.2 is 1.67 GWd/T or about 0.008 t1k. A 5% error in the peaking factor is therefore only.05/0.4 x 0.008 = 0.001 t1k. Statistically combining this small effect with all the other uncertainties would not affect the total uncertainty and can therefore be neglected.

c. Describe and justify the data fitting method used to derive the equations in the proposed Technical Specifications for Reactivity Category 4.

Enclosure NL-19-023 Page 16 of 54 Response to RAI 5.c The data fitting method is a generalized least squares method that minimizes the difference between the curve fit data points and the input data points from the loading curve table. It is known that the burnup requirement is a function of enrichment (enr) and cooling time (CT).

The requirement decreases exponentially with cooling time to an asymptotic value, so the generalized form is the following:

Burnup = A(enr) + B(enr) exp [ - C(enr) CT J where the coefficients A, B, and C are functions of enrichment.

If the functions are linear with enrichment, the difference between the curve-generated burnups and the table burnups was approximately 1 GWd/T. By making the functions quadratic in enrichment, the difference between the curve-generated burnups and table burnups was less than O. 1 GWd/T. The curve fit was adjusted so that all burnups from the curve fit were either equal to or greater than the data points thereby assuring that the curve fit was conservative.

NET-28091-0003-01, Revision 0, Section 5.5, "Limiting Depletion Parameters - Control Rod Operation," describes the licensee's analysis to address the effect of reactor operation with controls inserted on the post-irradiation reactivity of fuel assemblies. With respect to the operation and modeling of control rod insertion provide the following information:

a. There is an apparent disconnect in the analysis. The analysis for I P2 states that

"... assemblies depleted with 20 rodlet Pyrex burnable absorbers conservatively bounds assemblies that were operated with Control Bank D at the bite position." IP2 fuel assemblies that were under control rod bank D with the control rods inserted to the "bite" position but were modeled as having a 20 rodlet Pyrex burnable absorber were not assigned a burnup penalty as expected. For IP3 the analysis states that IP3 always operated at the all rods out condition with no controls inserted to the "bite" position. The analysis also states that all"... IP3 (A-U) batch grouping is depleted with a 20 rodlet Pyrex burnable absorber which is removed at 20 GWdff." Based on the licensee's stated intention and the IP2 analysis, the NRC staff did not expect the IP3 (A-U) batch grouping to have a burnup penalty assigned. Yet in Table 5.14 some of those fuel assemblies do have a burnup penalty. Explain why the IP3 fuel assemblies require a burnup penalty but the IP2 fuel assemblies do not.

Enclosure NL-19-023 Page 17 of 54 Response to RAI 6.a IP2 fuel assemblies were not assigned a burnup penalty if they were under control rod bank D with the control rods inserted to the "bite" position, but were modeled as having a 20 rod/et Pyrex burnable absorber. These assemblies never contained any burnable absorber, so the full reactivity effect of the Pyrex burnable absorber offsets the reactivity effect of the control rods at the bite position. For batches A to U of IP3, the modeling included 20 rod/et Pyrex burnable absorbers. The IP3 assemblies where a burnup penalty is required actually had burnable absorbers in the assembly and were placed under a control rod in the second cycle. Since these assemblies contained a burnable absorber there is no conservatism in the depletion analysis to offset even a small amount of D bank operation. Many of the IP3 A to U batch assemblies actually contained WABA rather than Pyrex burnable absorbers and they also had less than 20 fingers, but this difference from what was modeled was conservatively ignored. Thus, these assemblies need a burnup penalty.

In order to help follow the color coding, Table 6.1 is provided on the following page, and it summarizes the combination of 0-bank and absorbers studied.

Color Code Orange White Table 6. 1: Burnup Penalty for Assemblies under D-bank Actual D Bank Operation

<2GWdff Inserted

<2GWd/T Inserted Modeled D Bank Operation None

<2GWd/T Inserted Actual Removable Burnable Absorbers None Some Burnable Absorber 20 WABA or less Modeled Removable Burnable Absorbers 20 P rex 20 WABA or 20 Pyrex 20 WABA Burnup Penalty (GWDff) 1 0

Enclosure NL-19-023 Page 18 of 54

b. There is no information provided on the analysis performed to cover Batch Grouping IP3 (V-X). Provide a description of the analysis which was performed to cover Batch Grouping IP3 (V-X).

Response to RAI 6.b Just above Table 5.14 in NET-28091-0003-01, it states that there are three batch groupings for Indian Point 3 assemblies. However, with regard to depletion analysis there were only two groups. The early cycles A-U are depleted with a 20 finger Pyrex burnable absorber and a soluble boron concentration of 560 ppm. The later cycles (V-X) use the loading curve for batch Z which is depleted with a 20 finger WABA plus 148 IFBA rods at 1.5X and a soluble boron concentration of 950 ppm. IP3 Batch Grouping V-X is distinct from batch Z only because it uses lower enrichments than the fit generated for batch Z. This is discussed below Table 8.22 of NET-28091-0003-01.

Table 5.14 of NET-28091-0003-01 erroneously has IP3 W assemblies in cycle 9 colored orange. These assemblies should be green since they used Batch Z depletion parameters which accounts for 0-bank operation. As the error is conservative, no change to NET-28091-003-01 will be performed.

Enclosure NL-19-023 Page 19 of 54 NET-28091-0003-01, Revision 0, Section 5.6, "Depletion Analysis Model," describes the licensee's analysis to justify the use of the TRITON depletion sequence in SCALE 6.1.2 as the depletion code of record for this analysis by comparing three sets of TRITON delta k-effective to three sets of CASM0-5 delta k-effective by listing the difference between them. Using the comparison of the delta between two state points from one computer code to the delta between two state points from another computer code is inadequate to determine that one code is acceptable for use based on similarity to the other code as this method of comparison says nothing about how the codes compare at the actual state points. With respect to the use of the TRITON depletion sequence in SCALE 6.1.2 as the depletion code of record for this analysis provide the following information:

a. A comparison of the TRITON depletion sequence in SCALE 6.1.2 to an NRC approved depletion code.

Response to RAI 7.a In order to match the pool analysis, three cases for the IP2 pool are analyzed using CASM0-5 and TRITON for depletion and the 2x2 Region 2 model with a 3 out of 4 loading.

The cases represent the highest and lowest burnup requirement for the current fuel (Z batch fuel) and a low enriched historical fuel (D batch fuel). The cases are described below:

1) Z fuel, 5.0 wlo, 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> cooling, 1.2 PF, 48.19 GWdfl, PPM=950, TM0:600K, TFU:1000K

2) Z fuel, 4.2 wlo, 25 year cooling, 0.8 PF, 31.59 GWd/T, PPM=950, TM0:580K, TFU=900K

3) D fuel, 3.4 w/o, 10 year cooling, 1.2 PF, 27.82 GWdfl, PPM=570, TM0=600K, TFU=1000K where:

TMO is the moderator temperature, TFU is the fuel temperature, PPM is the soluble boron concentration, PF is the peaking factor.

The above cases utilized the current Batch Z fuel, while Batch D is older legacy fuel. The above cases were selected to span a range of fuel parameters that encompass the majority of fuel assemblies in the IP2 spent fuel pool (enrichment from 3.4 w/o to 5.0 w/o, PF from 0.8 to 1.2, Burnup from 27.82 to 48.19 GWDIMTU, soluble boron from 570 ppm to 950 ppm, moderator temperatures from 580K to 600K, fuel temperature from 900K to 1000K).

Cases 1 and 2 are depleted with 148 IFBA and a 20 finger WABA, which is never removed.

Case 3 is depleted with a 20 finger Pyrex, which is never removed. The 3 cases were depleted using both CASM0-5 and TRITON. Consistent with the CSA methodology, the fission gases and Pm-149 are reduced. The fuel was moved to a 2x2 pool model with one water hole in Region 2 (the burnable absorbers are removed from the fuel in the pool model). The pool model is identical except that TRITON number densities are used in one case and CASM0-5 number densities are used in the other case. The results are shown in Table 7. 1 below.

Enclosure NL-19-023 Page 20 of 54 Case 1

2 3

Table 7.1 Comparison of TRITON Depletion to CASM0-5 Depletion Enrichment Burnup Cooling TRITONk CASM0-5k

~k

{wt% LJ235)

(GWdff)

Time 5.0 48.19 72hr 0.9674 0.9648 0.0026 4.2 31.59 25 yr 0.9565 0.9555 0.0010 3.4 27.82 10 yr 0.9588 0.9570 0.0018 The TRITON depletion produces more reactive fuel than the CASM0-5 depletion in all three cases. Due to the conservative results when compared to a licensed fuel management tool (CASM0-5), the DSS-ISG-2010-01 acceptance of 5% of the delta k of depletion applies to this analysis.

b. Identify whether or not any additional biases and uncertainties are necessary.

Response to RAI 7.b As shown above in Table 7.1, TRITON is conservative for the Indian Point specific cases.

Therefore, no additional bias and uncertainty is needed.

RAIS Using the licensee's 'batch grouping' method sets reasonably bounding depletion parameters for each 'batch group.' The licensee's analysis identified fuel assemblies that did not meet its desired SFP storage category for those fuel assemblies. NET-28091-0003-01, Revision 0, Section 5.7, 'Special Case Depletions,' describes the licensee's analysis to use fuel assembly specific depletion parameters to qualify those fuel assemblies for its desired category. With respect to the 'Special Case Depletions' analysis provide the following information:

a. NET-28091-0003-01, Revision 0, Table 5.9 indicates the depletion soluble boron for IP2 Batch F and IP3 Batch Vis 580 PPM and 950 PPM, respectively. However, Table 5.21 indicates the special case depletions for IP2 F44 and its symmetric sisters and the IP3 V43/V48 sister used 540 PPM and 650 PPM, respectively. This reduction in the depletion soluble boron is not explained in the text. Provide the justification for using the lower depletion soluble boron.

Response to RAI a.a For F44, the actual average soluble boron experienced by this assembly (using Equation 1 from RAJ 1.a) is 503 ppm, so using 540 ppm in the depletion is conservative. For IP3 V43N48, the actual average soluble boron experienced by these assemblies (using Equation 1 from RAJ 1.a) is 525 ppm, so using 650 ppm in the depletion is conservative.

b. NET-28091-0003-01, Revision 0, states, "L48 and its sisters (L37, L38, L39, L44, L51, L52, and L64) spent two cycles on the outside corner of the core." Describe and provide the results of the radial burnup tilt of these fuel assemblies and its effect on their reactivity.

Enclosure NL-19-023 Page 21 of 54 Response to RAI 8.b The loading of these assemblies was such that the second cycle on the outside corner (opposite side of the core) reversed the direction of the flux so that the radial burnup tilt would be minimized. Therefore, the effect on reactivity is very small. To estimate a maximum reactivity effect, however, a radial tilt of 0.6 to 1.4 (from one corner to the opposite corner) in burnup gradient was applied to the k calculation of assembly L48. The gradient was applied by using 9 burnup zones (0.6, 0.7, 0.8, 0.9, 1.0, 1.1, 1.2, 1.3, and 1.4 times the average burnup). See Figure 8. 1 below. The assembly quadrant tilt for this case was 0. 76 to 1.24 which covers over 95% of the DOE data base on radial tilt (DOE/RW-0496 Horizontal Burnup Gradient Datafile for PWR Assemblies). The high quadrant tilts occur only for once burned fuel with low burnups. For this extreme case, the k increased by only 0.0024. This would be a maximum effect due to radial tilt. The k(95/95) for L48 was 0.9859 (from Table 8.24 of NET-28091-0003-01), so applying a maximum delta k of 0.0024 would increase this to 0.9883, which is still less than the target value of 0.99.

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Enclosure NL-19-023 Page 22 of 54

c. Two groups of fuel assemblies represented by assemblies W52 and X18 were depleted using their actual IFBA loading and burnup profiles. Describe and provide the results of the bias and uncertainty analysis associated with using the fuel assembly specific I FBA loading and burnup profiles.

Response to RAI 8.c The specific IFBA loading is used only in the depletion analysis (the CSA assumes no IFBA in the k calculations). If the IFBA loading uncertainty were treated separately, the effect on the total uncertainty is negligible. The separate W52 case with 1.5x loading during depletion was performed (instead of the actual 1.25x). The difference ink was only 0.0005. If the IFBA loading uncertainty were as large as.05x, the reactivity effect (LJ.k) of the IFBA uncertainty would be only.05/.25 x 0.0005 = 0.0001. Statistically combining this small effect with all the other uncertainties would not change the total uncertainty.

For these assemblies, the burnup profile is obtained by using the smallest relative power in each node from all assemblies in the group (8 assemblies). The 8 assemblies represent 8 independent measurements and, rather than use the average value at each node, the smallest value at each node is conservatively chosen. The model assumes all assemblies in the model have the limiting profile, which is also conservative. Since a bounding approach for the burnup profile is used, an additional bias or uncertainty associated with the burnup profile does not need to be included. The final margin for these assemblies is found on Table 8.24 of NET-28091-0003-01.

d. NET-28091-0003-01, Revision 0, states, "V43 and V48 did not initially meet the Category 3 fuel requirement but with special depletions, these two assemblies qualify for Category 3." If there were other changes from that 'batch group' depletion parameters than the already identified depletion soluble boron, describe and justify those changes.

Response to RAI 8.d The only changes in the depletion parameters for these two assemblies were that 60 IFBA rods are used during depletion (instead of 148) and 650 ppm soluble boron (instead of 950).

The change in the number of IFBA rods modeled is justified because only 60 IFBA rods are loaded into the V43N48 assemblies.

Determining the axial burnup profile to be used is a key component of the nuclear criticality safety analysis. The licensee's analysis has used a complex combination of familiar methods, but with new and unique variations to determine the burnup profiles for several different groupings of fuel assemblies. Those groupings will be addressed individually.

Enclosure NL-19-023 Page 23 of 54

a. The first grouping is for fuel assemblies that do not have axial blankets. The licensee's analysis indicates it used the burnup profiles from NUREG/CR-6801, "Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses." NUREG/CR-6801 evaluates approximately three thousand actual burnup profiles from numerous plant design types. The profiles are divided into twelve groups based on burnup, with the lowest numbered group having the highest burnup. The overall group profile is considered bounding for that group because they are at least three standard deviations from the average for the group. The use of NUREG/CR-6801 profiles is inconsistent with NRG interim staff guidance in DSS-ISG-2010-01, "Final Division of Safety Systems Interim Staff Guidance, DSS-ISG-2010-01, Revision O, Staff Guidance Regarding the Nuclear Criticality Safety Analysis For Spent Fuel Pools,"

(Agencywide Document and Management System (ADAMS) Accession Number ML110620086) because the licensee's analysis modified those profiles in new and unique ways. With respect to the axial burnup profiles used for non-blanketed fuel provide the following information:

1. The NET-28091-0003-01, Revision O, Section 6.2.1, "Axial Burnup Distribution,"

speculates that there was some aspect of the group profiles for Groups 2 and 4 that make the profiles for Groups 3 and 5 inappropriate for use. The analysis then eliminated the profiles for Groups 3 and 5 and instead uses the profile from Group 2 to cover the burnup ranges of Group 2 and Group 3 and the profile from Group 4 to cover the burnup ranges of Group 4 and Group 5. This is inconsistent with NUREG/CR-6801 as the statistical analysis of the groups included in NUREG/CR-6801 indicates profiles for Groups 3 and 5 are more representative of their respective groups, less of statistical outliers, than the profiles for Groups 2 and 4 are of their respective groups. Additionally, NUREG/CR-6801 indicates it is generally acceptable to use the burnup profiles from a lower burnup group for higher burnups, yet the licensee's analysis has gone in the opposite direction. Provide the justification for eliminating the profiles for Groups 3 and 5 and using the profiles from Groups 2 and 4 instead.

Enclosure NL-19-023 Page 24 of 54 Response to RAI 9.a.1 The data for NUREG/CR-6801 comes from end of life burnup of actual assemblies. The limiting profile for Group 2 would be the limiting profile for Group 3 had the reactor shutdown earlier (same for the Group 4 and 5 pairing). In order to cover this possibility, the analysis in NET-28091-003-01 uses the more conservative profile when a more reactive profile occurs for a bin having larger burnups. This is discussed in section 6.2. 1 of NET-28091-0003-01. The decision to eliminate Groups 3 and 5 makes the application more conservative than the recommended profiles. This is shown by the plots of relative axial burnups, Figures 9. 1 through 9. 18. For some of the axial nodes the approach taken in NET-28091-003-01 is non-conservative, but, as criticality requires several nodes, Figures 9. 19 through 9.26 show that the integrated burnup down from the most reactive top nodes result in lower burnups in the most reactive top of the fuel. Since NET-28091-003-01 used lower relative burnups than those recommended by NUREG/CR-6801, the analysis is conservative with respect to the NUREG.

2. The second modification to the NUREG/CR-6801 profiles was the licensee's averaging the profiles from one group with the next. The licensee indicated it started by assuming the group profile was only appropriate at the maximum burnup in the group. The licensee then performed a linear interpolation for each burnup profile node for a given assembly average burnup. The analysis claims this is conservative but provided no evidence to support that claim. NUREG/CR-6801 evaluated the profiles at the median burnup of the group range and determined the limiting group profile was conservative throughout that range. It is not clear how the licensee determined the burnup profiles that were used in the analysis. The example the licensee provides used the profile from Group 3, which the licensee had previously determined was inappropriate to use. No explanation was given for its use in this manner. Additionally, Group 1 has no upper burnup bound. It is not clear whether the linear averaging preserves the total burnup of an assembly. With respect to the averaging of the burnup profiles provide description of how the averaged profile is calculated. Provide the description and results of the analysis that demonstrates the averaging of the burnup profiles is conservative.

Response to RAI 9.a.2 The second modification was not averaging profiles but rather removing the discontinuities caused by the step function provided in NUREGICR-6801. The final burnup requirements are a smooth function of burnup. Having step changes in the axial burnup distribution as a function of burnup, does not allow linear interpolation between calculated values in a conservative manner (e.g., the top two nodes of Group 3 have higher relative burnups than the top two nodes of Group 2, even though Group 2 is for a higher burnup range). All steps are removed in a way that is conservative compared to the NUREG. The method of removing the steps is briefly described in Section 6.2. 1 of NET-28091-0003-01. Some details were not included. For example, since the highest burnup requirement is less than 46 GWd/T, Group 1 shapes (highest burnup group) are not used. The axial burnup profiles used are graphically presented in Figures 9. 1 through 9. 18 below.

Enclosure NL-19-023 Page 25 of 54 A lower relative burnup is always conservative. Figures 9.1 through 9.18 illustrate that, for each of the 18 axial nodes, the method in NET-28091-0003-01 generally uses lower relative burnups than the step function. However, in less important nodes (such as those near the bottom of the assembly) the relative burnup was sometimes higher. To better understand the impact, Figures 9. 19 through 9.26 show the integrated relative powers. From these figures, it can be seen that the method is conservative for the top half of the fuel assembly, which dominates the reactivity. Note that the last graph (integration over the entire top half) shows the NET-28091-0003-01 approach is much more conservative. This is because the gh node in the NET-28091-0003-01 method averages the bottom nodes and effectively conservatively moves the lower burned bottom up toward the middle. The bottom of the fuel is less reactive than the top of the fuel and has no effect on k. This is supported by the response in 9.b.1 (Table 9.1) and NET-28091-003-01 Table 6.4.

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Enclosure NL-19-023 Page 26 of 54

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Enclosure NL-19-023 Page 27 of 54 1.1 1.05 0.95

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Enclosure NL-19-023 Page 28 of 54

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Enclosure NL-19-023 Page 29 of 54 r I I

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Enclosure NL-19-023 Page 30 of 54 1.2 1.18 1.16 1.14 Q. E 1.12 a:i 111.l I ~ 1.08 I

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Enclosure NL-19-023 Page 31 of 54 1.25 1.2 1

10 15 20 Node 12 NUREG/CR-6801 This CSA 25 30 35 40 45 50 Burnup (GWd/T)

Figure 9.12: Axial Relative Burnup (NUREGICR-6801, NET-28091-0003-01, Node 12) 1.25 1.2 1.05 10 15 20 Node 13 NUREG/CR-6801 This CSA 25 30 Bumup (GWd/T) 35 40 45 so Figure 9.13: Axial Relative Burnup (NUREGICR-6801, NET-28091-0003-01, Node 13)

Enclosure NL-19-023 Page 32 of 54 1.25 1.2 I D.

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Enclosure NL-19-023 Page 33 of 54 I

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Enclosure NL-19-023 Page 34 of 54 I c.

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Enclosure NL-19-023 Page 35 of 54 L_

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Enclosure NL-19-023 Page 36 of 54 4.6 4.4 4.2 Q.

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Enclosure NL-19-023 Page 37 of 54 r- -

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Enclosure NL-19-023 Page 38 of 54 Top Half (Nodes 1 - 9) 8*9 NUREG/CR-6801 This CSA 8.7

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b. The second grouping is for legacy fuel assemblies that do have axial blankets. The licensee's analysis divided these fuel assemblies into 5 batch groups. The method for determining the axial burnup profile in each group is described as taking the lowest burnup for each node of all the actual profiles in the group to amalgamate one profile for each batch group. These derived profiles are then further modified by using the top eight nodes and the"... ninth node from the top is used for the ninth and all lower nodes." The profiles derived using this method are listed in Table 6.2. The analysis states this method is conservative. However, the profiles listed in Table 6.2 have a volume averaged relative burnup greater than one. This indicates they are potentially non-conservative profiles as the modeled fuel assembly would have a higher burnup than the actual fuel assembly. With respect to the axial burnup profiles listed in Table 6.2 and used for blanketed fuel provide the following information:

1. The impact on the margin in the NET-28091-0003-01, Revision O analysis due to modeling the fuel assemblies with more burnup than they have.

Enclosure NL-19-023 Page 39 of 54 Response to RAI 9.b.1 Table 6.4 of NET-28091-0003-01 shows that using only 9 nodes having a volume averaged burnup greater than 1.0 is conservative compared to using all nodes having a volume averaged burnup equal to one for the Batch Z fuel. The reactivity impact of using only 9 nodes is very small but conservative in all cases. This is expected since the burnup in the bottom nodes is always higher than the burnup in the corresponding top nodes, and, so, the reactivity is being driven by the top. The reason that the 9 node model is conservative compared to the "all nodes" model is that the center nodes below node 9 in the 9 node model are using a lower burnup than the "all nodes" model. The burnup becomes higher in the 9 node model only for the bottom nodes, but, as noted above, the bottom nodes always have more burnup than the corresponding top nodes, which make them unimportant compared to the top nodes.

The effect of using 9 nodes is the same when applied to the Batch groups on Table 6.2.

As confirmation for the shapes given on Table 6.2, selected cases are analyzed in a 3 out of 4 Region 2 model for each shape comparing the 9 node model with a model using all nodes (either 24 or 26 axial nodes). As shown in Table 9.1 below, the 9 node model nearly always produces a calculated k that is equal (to four decimal places) to the all nodes model. The reason is that the top dominates the reactivity, and the top is modeled exactly the same in both the 9 node model and the "all nodes" model.

Table 9. 1: Comparison of 9 Node Model to "All Nodes" Model Case*

9node k All node k Delta k 3.8B22S 1.0041 1.0040 0.0001 5.0B42S 0.9549 0.9549 0.0000 3.8 830 V 0.9786 0.9786 0.0000 5.0846 V 0.9489 0.9489 0.0000 3.8830 W 0.9856 0.9856 0.0000 5.0846 W 0.9575 0.9575 0.0000 3.8 830 X with WABA 0.9860 0.9860 0.0000

5. O 846 X with WABA 0.9572 0.9572 0.0000 5.0 846 X no WABA 0.9532 0.9532 0.0000

• Enrichment, Burnup (GWd/T), Batch Group

2.

Describe any biases or uncertainties associated with using these burnup profiles and how they are included in the analysis.

Enclosure NL-19-023 Page 40 of 54

RAI 10

Response to RAI 9.b.2 The process used to determine the axial burnup distribution for blanketed fuel is consistent with NE/ 12-16, Revision 3. Specifically, Option 2 in Section 5. 1.4 of NE/ 12-16 is applied to each of the five batch groups of assemblies, ensuring that a conservative axial burnup distribution is developed by utilizing the lowest relative burnup of each of the nine nodes modelled in the analysis. Because an artificially conservative profile is used for each batch group, no further bias or uncertainty needs to be applied.

Note that the fuel covered in Table 6.2 of NET-28091-0003-01 is already discharged and the data used to create the profiles comes from 100% of the assemblies where the shape is to be applied.

NET-28091-0003-01, Revision 0, Section 6.1, "SCALE 2x2 Radial Models," and Section 6.5, "Convergence of the 2x2 Infinite Model Calculations," discuss the 2x2 models in general terms, but no specifics are provided. Given the various possible combinations of batch groupings, burnup profiles, nodalization, and other parameters it is unclear what the licensee modeled to reach its conclusions. With respect to the 2x2 models provide the following information:

a. Describe which combinations were modeled using a 2x2 array and the results of those calculations.

Response to RAI 1 O.a All of the 2x2 models use the same batch grouping at the same enrichment and burnup in all fueled locations. This is more limiting than modeling different batch groupings or different enrichment/burnup combinations in the 2x2 model because the reactivity is maximized when identical fuel assemblies are modeled. When different enrichments or batch groupings are combined in the same 2x2 model, the axial reactivity profile is different in different storage cells, causing the differing cells to compete for the neutrons, which lowers k. Therefore, no combinations of batch groupings were modeled, as the combinations would always be bounded by the most limiting uniform batch model.

RAI 11

In NET-28091-0003-01, Revision O, Section 6.6.2, "Convergence of the Full Pool Model," the discussion indicates the licensee performed six calculations with different starting distributions for the neutron source term. Four of those started the neutron source term near the SFP wall, leading to a significant portion of the initial source term leaking out of the SFP. Three of those started in Category 5 fuel which is intentionally required to have excess burnup to make it non-limiting, thus further reducing the portion of the initial source term than reaches the more reactive fuel. This paucity of sampling locations indicates that the models may be under sampled for such a loosely coupled system. With respect to the convergence of the full pool model provide the following information:

a. Provide at least one model with known convergence for comparison to the other cases, for both the unborated and borated cases.

Enclosure NL-19-023 Page 41 of 54 Response to RAI 11.a Section 6.6.2 of NET-28091-0003-01 shows that the convergence of k for the full pool model is robust. It shows that for the number of neutrons per generations and the number of generations used in the NET-28091-0003-01 models, the k is converged no matter where the initial source is placed. The NET-28091-0003-01 method uses a uniform initial source, unless the area producing the highest k is known and there is a desire to increase the source in that area (e.g., for a single misloaded assembly the start source is placed near the misload). To augment the information given on NET-28091-0003-01 Table 6.6, the reference analysis (uniform start source) was rerun with 20,000 generations with 40,000 neutrons per generations. The calculated k is 0. 968853 + or - 0. 000025. This is within 2 sigma of the reported (NET-28091-003-01 Table 6.6) uniform source solution done with the normal number of generations and neutrons per generation. The k for all cases on NET-28091-003-01 Table 6.6 is within 3 sigma of the calculated reference k (0.968853). This full pool k is due to the reactivity from Region 1. A case was run where Region 2 is removed leaving only water. The k of the "region 1 only" model is 0. 9684 + or - O. 000065. Notice that this is very close to the full pool model k with both regions in the model. The slight difference (0. 00045) shows that Region 2 has only a small effect on the full pool k. The Region 2 only k is 0.9584 (see Table 6.5 of NET-28091-0003-01).

Figure 6.16 of NET-28091-003-01 shows how the average k changes by generation with the start source placed in the wrong locations. The final k from all the cases are shown on Table 6.6 and all the final k's are within expected Monte Carlo uncertainty. Figure 6.16 of NET-28091-003-01 may seem like this is not possible since the final average k's are not the same. Figure 6. 16 does not account for the skipped generations, so the final average k's are not the same. After the skipped generations are removed, all the lines on Figure 6. 16 of NET-28091-003-01 would be on top of each other after the 8000 generations. Figure 11.1, below on the following page, shows the same analysis but with the average k after the skipped generations. Please note that Figure 11. 1 has an expanded scale in order to show the small differences in k.

Enclosure NL-19-023 Page 42 of 54

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• Region 1 Left Bottom

• Top Right

• Far Right Bottom Bottom Right Cask Area Figure 11.1: Average calculated k as a function of number of generations (with skipped generations removed) for various starting locations

Enclosure NL-19-023 Page 43 of 54 The RAJ also considers borated conditions. Except for the multi-misload conditions, there is a large margin in k. The most limiting large model borated case is the Region 2 misload where every assembly in Region 2 is replaced with a 5.0 w/o fuel assembly with a burnup of 24 GWd/T (water holes and control rods are maintained). This full misload case has characteristics that are very different than the normal operation cases. By design, for normal operations, the k after bias and uncertainty is nearly the same over the entire region.

This was done by requiring 11 GWd/T extra burnup for Category 5 fuel. For the multi-misload case, all of the fuel has the same burnup so the reactivity in the Category 5 areas is higher than in the Category 4 areas. Additionally, with 2000 ppm soluble boron, the negative reactivity of the water holes is much larger and the negative reactivity of the control rods is Jess. This results in the Category 4 area of the pool having sufficiently low reactivity that neutrons cannot cross this portion of the pool. Analysis similar to that supporting Table 6.6 of NET-28091-0003-01 is performed for this misload case. Since this is a Region 2 only model, the two Region 1 start source cases are replaced with start source cases at the left top and bottom of Region 2. Table 11. 1 below provides the results of this analysis. Figure 11.2 below shows the location of the start sources. The peak reactivity area is in the top right Category 5 area of Region 2. As expected, the three cases where the start source is isolated from the peak reactivity area (by the Category 4 area) have lower calculated k's because the neutrons do not get a chance to "see" the peak reactivity area.

The cases not isolated by Category 4 all yield the same k. Since the analysis of NET-28091-0003-01 uses a uniform start source with 8000 generations and 16000 neutrons per generation, convergence is assured.

Enclosure NL-19-023 Page 44 of 54 Lf GE.m O

wm MTEIUM.2

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(Region 2 Model - 8000 Generations, 16000 Neutrons per Generation)

Source Location Calculated Reported Generations k

Sigma Skiooed Uniform for Problem 0.912440 0.000059 468 Top left of Region 2 0.905064 0.000056 458 Bottom Left 0.904838 0.000054 370 Bottom Right 0.904846 0.000055 893 Cask Area Right Corner 0.912576 0.000060 1301 Top Riqht 0.912456 0.000056 376 The uniform source case was rerun with 10,000 generations and 20,000 neutrons per generation and the calculated k was k= 0.912548 +/- 0.000045 after skipping 699 generations. All of the final k's are within 2 sigma of the reference case except when the starting neutrons are isolated from the peak reactivity area.

Enclosure NL-19-023 Page 45 of 54 All of these cases demonstrate that using the SCALE default of a uniform source and having a large number of neutrons to track will assure convergence even when there are multiple peak reactivity locations isolated from each other.

RAI 12

NET-28091-0003-01, Revision 0, Section 7.3, "Eccentricity," describes the analysis the licensee performed to consider the eccentric of position of fuel assemblies within a storage cell. The base k-effective calculation models all fuel assemblies in the center of their respective storage cell. However, the fuel assemblies can be anywhere within their respective storage cell. The

'eccentricity' portion of the analysis is intended to determine the reactivity effect of the fuel assemblies being in other than the center of their storage cell. The licensee's analysis used an 8x8 storage cell array to calculate the eccentricity effect. The licensee's analysis considered sixteen fuel assemblies to be eccentrically located while maintaining the outer rows of the 8x8 array centrally located. Using periodic boundary conditions on the 8x8 array effectively models an infinite number of these 8x8 arrays. But the boundary rows serve to minimize the nuclear interaction between them, therefore the resultant k-effective is representative of a single 8x8 array. The licensee's analysis made two implicit assumptions: (1) that there is only one eccentric configuration that yields the maximum reactivity and (2) that the fuel assembly can only be in one of four distinct locations within a storage cell. With those assumptions the licensee calculated the probability of occurrence of the worst case eccentric positioning configuration to be 2.3E-10. The inverse would be the number of possible combinations. With respect to the eccentricity effect provide the following information:

a. Given the large number of possible combinations, justify the combination used to calculate the reactivity of the eccentricity effect.

Response to RAI 12.a There are a large number of possible combinations of eccentrically located fuel assemblies, but, rather than use a combination specific bias, the largest bias of the possible combinations is used. The largest bias occurs when the assemblies are placed eccentrically toward a single point. It is logical to expect the configuration that would maximize the reactivity would be to eccentrically locate assemblies at the center of the cruciform created by the 3-out-of-4 arrangement. See Figure 7.1 in NET-28091 -0003-01. In Region 2, the movement of the assemblies toward this point does maximize the reactivity with eccentric positioning. However, moving the assemblies toward this point leaves larger water gaps between the assemblies which results in a negative reactivity. Thus centered positioning is most reactive and so for Region 2 there is no eccentricity bias.

In order to confirm that aligning the assemblies closest to the center of the cruciform is the most reactive eccentric positioning for Region 2, additional calculations have been performed. They are:

Enclosure NL-19-023 Page 46 of 54

1. Perform the move to the center eccentricity calculation with low enrichmentlburnup assemblies.

2. Move the assemblies toward the midpoint of the cell wall of one side of the cruciform with high and low burnup assemblies. See Figure 12. 1 below.

3. Move the assemblies toward one of the corners of the central cell of the cruciform with low and high burnup assemblies. See Figure 12.2 below.

Table 12.1 shows all the calculations performed for the eccentricity bias. The additional analysis confirms that the centered positioning in the cells is most limiting for Region 2.

Region 1 allows a 3-out-of-4 arrangement of Category 2 assemblies. The Category 2 burnup requirement is 21 GWd/T for all enrichments, and, so, analysis is performed only with 5 wt% fuel at 21 GWd/T burnup. For Region 1, there is an eccentricity bias. The same three arrangements described above are tested to determine the maximum bias. Figure 12.3 shows the model of the eccentricity toward the corner. As seen on Table 12.1, there is a small difference in the size of the eccentricity bias between the three cases (0. 0006 range). The eccentricity to the corner is the most limiting, and this case was not considered in NET-28091-003-01. The eccentricity bias is 0.0024. This is 0.0002 higher than 0.0022, which is what is reported in NET-28091-003-01 rev. 0. The reference case k changed from the previous analysis since more neutrons (16,000 generations with 32,000 neutrons per generation) were used to reduce the uncertainty in this small delta k test. Assuming that there are not multiple short cycles (<20,3 GWDIMTU), there is currently margin in the analysis for this small change in the bias. Should multiple short fuel cycles occur in the future, an increase in reactivity of 0.0004 would exist (see Table 8.11 of NET-28091-003-01, Rev.OJ, and the resulting margin to the regulatory limit would be 0.0094 delta-k.

The eccentricity bias gets larger with more assemblies eccentric about a point. This analysis made a probability argument in Section 7.3 of NET-28091-003-01 that 16 eccentric assemblies is more than sufficient. This is consistent with other industry CSA 's, such as North Anna [ML18180A197] and Millstone Unit 2 [ML 16003AOOBJ. As can be seen on Table 12.1 below, if fewer assemblies are asymmetric the effect is smaller.

Enclosure NL-19-023 Page 47 of 54 Table 12.1: Eccentricity Results Burnup Category Calculated k

Reference (all centered) 48.19 4

0.9586 16 Eccentric Assemblies about center 48.19 4

0.9585 assembly 16 Eccentric Assemblies Shifted down 48.19 4

0.9584 16 Eccentric Assemblies Shifted to 48.19 4

0.9585 corner Reference (all centered}

31.58 4

0.9604 16 Eccentric Assemblies 31.58 4

0.9602 16 Eccentric Assemblies Shifted down 31.58 4

0.9600 16 Eccentric Assemblies Shifted to 31.58 4

0.9601 corner Reference (all centered) 21 2

0.9672 16 Eccentric Assemblies 21 2

0.9690 16 Eccentric Assemblies Shifted down 21 2

0.9693 16 Eccentric Assemblies Shifted to 21 2

0.9696 corner 2 Eccentric assemblies (corner) 21 2

0.9674 3 Eccentric assemblies (corner) 21 2

0.9676 7 Eccentric assemblies (corner) 21 2

0.9684 Sigma Lik 0.00004 0.00004

-0.0001 0.00005

-0.0002 0.00006

-0.0001 0.00005 0.00006

-0.0002 0.00006

-0.0004 0.00006

-0.0003 0.00003 0.00006 0.0018 0.00006 0.0021 0.00003 0.0024 0.00007 0.0002 0.00007 0.0004 0.00006 0.0012

Enclosure NL-19-023 Page 48 of 54 2--0 M>lAl CRO'J5 SCCTIDN Figure 12. 1: Region 2 eccentric down to middle of center cell LfGEtC>

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Enclosure NL-19-023 Page 49 of 54 Centering Point 1- 0 RADIAL CRO!i5 st.CT ION MTlAI~ 2 MTEfUA.. 4 1S Figure 12.2: Region 2 eccentric up and left to corner of center cell

Enclosure NL-19-023 Page 50 of 54 Figure 12.3: Region 1 eccentric up and left to corner of center cell Centering Point

Enclosure NL-19-023 Page 51 of 54

RAI 13

NET-28091-0003-01, Revision O, Section 8.5, "Alternate Arrangements for Region 1," describes how fresh poisoned fuel can be stored in Region 1. The analysis proposes two rules: ( 1) Each Category 1 cell must be face adjacent with at least three water holes and (2) Each Category 2 cell may not have more than one face adjacent to a Category 1 cell. With respect to the alternate arrangements for Region 1 provide the following information:

a. The analysis states that the k-effective for checker boarding Category 1 fuel in Region 1 is 0.8548. Table 8.12, "Dependence of kett on the Region 1 Arrangement," lists the k-effective for Arrangement Number 1, Reference - No Category 1 Arrangement, as 0.9665. Table 8.12 lists the k-effective for Arrangement Number 3, No Category 2 Arrangement, as 0.9651. In both Arrangement Number 1 and 3 there are Category 3 and 5 fuel assemblies in positions consistent with the "normal" configuration. Since Category 3 and Category 5 fuel are supposedly less reactive than Category 2 and the k-effective of a checkerboard of Category 1 fuel is 0.8548, replacing all the Category 2 fuel assemblies with a checkerboard of Category 1 fuel assemblies should have resulted in a larger decrease in reactivity than is being reported. Explain the apparent discrepancy.

Response to RAI 13.a The reactivity of the no Category 2 case is dominated by the boundary between Category 1 and Category 3 fuel. Replacing all the Category 2 fuel assemblies with a checkerboard of Category 1 fuel simply makes that part of the model even less important than it already was.

Therefore, the reactivity will be driven by the important part of the model, and, so, k will not decrease appreciably.

b. The analysis states that the k-effective for checker-boarding Category 1 fuel in Region 1 is 0.8548. Table 8.12 lists the k-effective for Arrangement Number 4, Maximum Category 1 Arrangement, as 0.9584. Arrangement Number 4 is a modified checker-board of Category 1 fuel assemblies with ten Category 3 fuel assemblies placed between Category 1 fuel in accordance with Rule 1, and one Category 5 fuel assembly that is replacing a Category 1 fuel assembly. While an increase in k-effective over a pure checker-board of Category 1 fuel assemblies due to a few the Category 3 fuel assemblies interspersed with the Category 1 fuel would be expected, the reported increase for Arrangement Number 4 seems excessive. Describe the models used in detail and explain the large increase.

Response to RAI 13.b The models for the alternate arrangements discussed in Section 8.5 of NET-28091-0003-01 are the same as described in Section 6.6 of NET-28091-0003-01 except for the change in assembly category placement. Thus, the model contains both Region 1 and 2. Normally, the Region 1 reactivity dominates, but, in the Maximum Category 1 Arrangement, the reactivity of Region 1 is actually less than Region 2. The calculated k of 0.9584 for the full

Enclosure NL-19-023 Page 52 of 54 pool model is the k for Region 2. See Table 6.5 of NET-28091-0003-01 for the Region 2 only k.

c. The Category 3 fuel assemblies in the Arrangement Number 4 are either on the periphery or one row from the periphery. Proposed Rule 1 would allow an additional dozen or more Category 3 fuel assemblies to be placed in more interior locations where there would be less leakage indicating that the stated bounding scenario may not be bounding. Demonstrate that the scenario analyzed remains bounding accounting for the maximum allowed number of Category 3 fuel assemblies in the Maximum Category 1 Arrangement.

Response to RAI 13.c The intent of rules for placement of Category 1 assemblies in Region 1 was to overlay a Category 1 checkerboard on top of the reference Region 1 arrangement. It was not intended to allow Category 3 to move in from the outer two rows of Region 1, nor to allow Category 5 assemblies to move away from the edge of Region 1. The following additional rule will be added to the Technical Specifications Insert 3 item 12: c. The Category 3 and 5 locations in Figure 3.7.13-1 cannot be moved.

d. The proposed rules would allow a Category 1 fuel assembly to be face adjacent with another Category 1 fuel assembly on one side. Provide the results of this case to demonstrate that it is bounded by the analyzed case.

Response to RAI 13.d The intent was that the Category 1 fuel is in a checkerboard arrangement with an exception given in the first rule. The first rule is revised to: a) Category 1 fuel assemblies must be face adjacent to at least three Water Holes and not face adjacent to another Category 1 assembly.

RAI 14

NET-28091-0003-01, Revision 0, Section 9.5, "Multiple Misleads," analyzes one scenario for Region 2. That scenario assumes that all cells that are permitted to contain fuel are modeled as misloaded with once burned 5.0 weight percent fuel with a burnup of 24 gigawatt-days per metric ton of uranium except for the water hole and 50% water hole locations, which remain empty. The analysis provides an inadequate justification for this being the limiting misleading because filling in the water hole and fresh fuel are not considered. Therefore, justify the limiting misleading analyzed in Region 2 by explaining why water hole, 50% water hole, and fresh fuel assemblies were excluded from the analyzed misleading scenario.

NET-28091-0003-01, Revision 0, Section 9.5, "Multiple Misleads," analyzes one scenario for Region 2. That scenario assumes that all cells that are permitted to contain fuel are modeled as misloaded with once burned 5.0 weight percent fuel with a burnup of 24 gigawatt-days per metric ton of uranium except for the water hole and 50% water hole locations, which remain empty. The analysis provides an inadequate justification for this being the limiting misleading

Enclosure NL-19-023 Page 53 of 54 because filling in the water hole and fresh fuel are not considered. Therefore, justify the limiting misleading analyzed in Region 2 by explaining why water hole, 50% water hole, and fresh fuel assemblies were excluded from the analyzed misleading scenario.

Response to RAI 14 The multiple misload scenario analyzed in Section 9.5 is based on implementing additional procedural barriers and visual indicators to prevent more reactive misloading scenarios (e.g., fresh fuel in every location, or placement of a fuel assembly in a location intended to not contain a fuel assembly, such as water holes and 50% water holes). The visual indicators include:

Implementation of a fixed storage configuration in the Region 2 storage modules. By fixing the same storage cells to remain empty of fissile material, there is a known and unchanging configuration of the cells that do not contain fuel assemblies that is easily verified and implemented by the fuel handlers. The regular pattern of empty storage cells, or storage cells containing control rods is visually verifiable from the fuel bridge. Also, a color coded SFP map is available to the fuel handlers during fuel movement.

When the configuration of the Region 2 racks is changed to allow a checkerboard of Category 1 fuel, one row of empty cells to isolate the checkerboard from the remainder of the Region 2 storage cells is implemented to provide an additional visual discernment in Region 2. Additionally, a formal communication between move sheet developers and the fuel moves is implemented to ensure that the fuel movers understand that the checkerboard area will be isolated to allow for fresh fuel to be placed in Region 2. This will also be updated on the color coded SFP map.

Visual verification of fresh versus spent fuel by fuel handling operators during fuel movement in the Region 2 racks to prevent loading of new, unirradiated fuel assemblies in these storage rack modules. Although the criticality analysis allows for a checkerboard of Category 1 fuel in Region 2, it is important to note that this is the exception, not the rule and requires formal communication between those generating the move sheets and those performing them to recognize the significance of isolating the checkerboard area in Region 2 from the remainder of the Region 2 storage racks.

The procedural barriers and required training include:

Use of licensee controlled administrative programs as described in Section 6.3.5.1 of NE/ 12-16, such as:

o Production of reports that show acceptability of fuel assembly locations during move sheet development o

Visual, color-coded spent fuel pool maps showing acceptability of fuel assembly locations o

Pre-verification of planned fuel moves with an independent review of move sheets during development

Enclosure NL-19-023 Page 54 of 54 o

Detailed administrative procedures for implementation o

Training and qualification of engineers responsible for spent fuel assembly selection and verification o

Training of responsible engineers prior to implementation of new storage configurations or Technical Specification loading curves Qualified fuel movers are required to be trained such that they understand the requirements of the Technical Specifications and their role in ensuring they are implemented properly. Examples of elements of fuel mover training includes:

o How to identify the area of the pool that contains the 3-out-of-4 configuration and actions required in the event a move sheet incorrectly places a fourth assembly in the 3-out-of-4 configuration, o

How to identify the RCCA checkerboard area in the pool and how to identify if a move would violate Technical Specifications associated with that area, o

Actions required if a new fuel assembly is to be placed in Region 2 and no prior communication has been received about an isolated area in Region 2 being created for new fuel assemblies

Enclosure Attachment 1 NL-19-023 Technical Specification Page Markups

Spent Fuel Pit Storage 3.7.13

3. 7 PLANT SYSTEMS Replaced by revised LCO (Insert 1)

3. 7.13 Spent Fuel Pit Storage APPLICABILITY:

INDIAN POINT 2 IP2 fuel assemblies stored in the Spent Fuel Pit shall be classified in accordance with Figure 3.7.13-1, Figure 3.7.13-2, Figure 3.7.13-3, and Figure 3.7.13-4, based on initial enrichment, burnup, cooling time and number of Integral Fuel Burnable Absorbers (IFBA) rods; and, Fuel assembly storage location within the Spent Fuel Pit shall be restricted to Regions identified in Figure 3.7.13-5 as follows:

Fuel assemblies that satisfy requirements of Figure 3.7.13-1 may be stored in any location in Region 2-1, Region 2-2, Region 1-2 or gion 1-1;

b.

Fuel a semblies that satisfy requirements of Figure 3. 7.13-2 may be store

• any location in Region 2-2, Region 1-2 or Region 1-1;

c.

Fuel assembli that satisfy requirements of Figure 3.7.13-3 may be stored in any I ation in Region 1-2, Region 1-1, or in locations designated as "perip ral" cells in Region 2-2; and

d.

Fuel assemblies that satis requirements of Figure 3. 7.13-4 may be stored:

1)

In any location in Region 1-

2)

In a checkerboard loading config tion (1 out of every two cells with every other cell vacant) in

3)

In locations designated as "peripheral" eel in Region 2-2.

I P3 fuel assemblies shall be stored in Region 1-2 of the Spent el Pit.

Only assemblies with initial enrichment 2:: 3.2 and s; 4.4 w/o U235 a discharged prior to I P3 Cycle 12 shall be stored in the Spent Fuel Pit.

Whenever any fuel assembly is stored in the Spent Fuel Pit.

3.7.13-1 Amendment No.

ACTIONS INSERT the following NOTE:

- NOTE-Separate Condition entry is allowed for each fuel assembly.

Spent Fuel Pit Storage 3.7.13 CONDITION REQUIRED ACTION COMPLETION TIME A.

Requirements of the LCO not met.

A.1

- NOTE-LCO 3.0.3 is not app I icab I e. -----------------------

Initiate action to move the Immediately SR 3.7.13.1 INDIAN POINT 2 noncomplying fuel assembly to an acceptable location.

Replaced by revised Surveillance Requirements (Insert 2)

SURVEILLANCE Verify by inistrative means that the IP2 fuel assembly has ti classified in accordance with Figure 3.7.13-1, Figu

.7.13-2, Figure 3.7.13-3, or Figure 3.7.13-4 and meets intended storage location.

OR Verify by administrative means that the IP3 fuel assembly meets the requirements for the intended storage location.

3.7.13-2 FREQUENCY Prior to storing the fuel assembly in the Spent Fuel Pit.

Prior to storing the assembly in the Sp Fuel Pit.

Amendment No.

J I-

~ -

0 (9

a.

C co 60 50 40 30 20 10 0

Replaced by new Figure 3. 7.13-1 (Insert 3) 0 Years Cooling 5 Yr Cooling 10 Yr. Cooling 15Yr Cooling O Yr. Cooling Spent Fuel Pit Storage 3.7.13 Not Acceptable for Storage in Region 2-1 2

3 5

Initial Enrichment, w/o U235 Figure 3.7.13-1 IP2 Fuel Assembly Limiting Burnup and Cooling Time versus Initial Enrichmen

• INDIAN POINT 2 Acceptable for Storage in Any Location in Region 2-1, Region 2-2, Region 1-2 or Region 1-1 3.7.13-3 Amendment No.

40 30 I-

~ -

0 20 CJ Q.

J C

lo,,,.

J ca 10 0

1 Replaced by new Figure 3.7.13-2 (Insert 4) and new Tables 3.7.13-1, 3.7.13-2, and 3.7.13-3 (Insert 5) 0 Cooling Time 5 Years Cooling 10 Years Cooling 15 Years Cooling 2

3 Initial Enrichment, w/o U235 Figure 3.7.13-2 Spent Fuel Pit Storage 3.7.13 5

IP2 Fuel Assembly Limiting Burnup and Cooling Time versus Initial Enrichment:

INDIAN POINT 2 Acceptable for Storage in Any Location in Region 2-2, Region 1-2 or Region 1-1 3.7.13-4 Amendment No.

30

a.

~

20

i...

J co E

J E C

~

10 0

1 Delete Figure 3.7.13-3 Accepta le for Storage in Region -1 2

3 Spent Fuel Pit Storage 3.7.13 Not Acceptable for Storage in Region 1-1 Initial Enrichment, w/o U235 Figure 3.7.13-3 IP2 Fuel Assembly Limiting Burnup versus Initial Enrichment:

Acceptable for Storage in Any Location in Region 1-2, Region 1-1, or in locations designated as "peripheral" cells in Region 2-2.

INDIAN POINT 2 3.7.13 - 5 Amendment No.

(/)

""C 0 a:

<(

en LL -

0,._

Q) 25 20 15 Delete Figure 3.7.13-4 in Regio 1-2 or Checkerboard Loading in eg1on 1-1 Spent Fuel Pit Storage 3.7.13

..c 10 E

J z 5

0 4.5 4.6 4.7 4.8 5

Initial Enrichment, w/o U235 Figure 3.7.13-4 IP2 Fuel Assembly Minimum number of IFBA rods versus Initial Enrichment:

1)

Acceptable for Storage in Any Location in Region 1-2, or

2)

Acceptable for Storage in a checkerboard loading configuration in Region 1-1, o

3)

Acceptable for Storage in locations designated as "peripheral" cells in Region 2-2.

INDIAN POINT 2 3.7.13-6 Amendment No.

G E

C A

N.,..

PERIPHERAL CELL INDIAN POINT 2 Delete Figure 3.7.13-5 Figure 3.7.13-5 Spent Fuel Pit Rack Layout 3.7.13-7 Spent Fuel Pit Storage 3.7.13 Amendment No.

LCO 3.7.13 INSERT 1 IP2 fuel assemblies stored in the Spent Fuel Pit shall be categorized in accordance with Table 3.7.13-1 or, if pre-categorized, Table 3.7.13-2.

IP3 fuel assemblies stored in the Spent Fuel Pit shall be categorized in accordance with Table 3.7.13-1 or, if pre-categorized, Table 3.7.13-3.

IP2 and IP3 fuel assembly storage locations within the Spent Fuel Pit shall be restricted to locations allowed by Figure 3.7.13-1 and its associated notes.

---

1\\!ote----------------------------------------------

Regarding Category 5 fuel assemblies that are required by Figure 3.7.13-1 to contain a full length RCCA - The RCCA must not be placed in or removed while the assembly is in an RCCA required location unless all 8 adjacent cells are empty.

INSERT 2 SURVEILLAI\\ICE REQUIREMEI\\ITS SR 3.7.13.1 SURVEILLAI\\ICE Verify by administrative means that the IP2 fuel assembly has been categorized in accordance with Table 3.7.13-1 or, if pre-categorized, Table 3.7.13-2 and meets the requirements for the intended storage location.

OR Verify by administrative means that the IP3 fuel assembly has been categorized in accordance with Table 3.7.13-1 or, if pre-categorized, Table 3.7.13-3 and meets the requirements for the intended storage location.

FREQUEI\\ICY Prior to storing the fuel assembly in the Spent Fuel Pit.

Prior to storing the fuel assembly in the Spent Fuel Pit.

INSERT 3 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 16 17 18 19 20 l1 ll 23 24 25 26 27 28 29 30 31 H

l--+-+-+-+-+--+--+--+--+----ll---+--+----f----f----ff--1--+-+-+-+--+--il--+--+----+-+--+-

G l--+-+-+-+-+--+--+--+--+----ll---+--+---1---1---lf--l--+-+-+-+-+--ll--+--+--l--l--+-+--+-

F l--+-+-+-+-+--+--+--+--+----l l---+--+----f----f----ff--1--+-+-+-+-+--il--+--+----+-+-+-

E l--+-+-+-+-+--+--+--+--+----l l---+--+---1---1---lf--l--+-+-+-+-+--ll--+--+--l--l--+-1---+-

D l--+-+-+-+-+--+--+--+--+----ll---+--+----f----f----ff--1--+-+-+-+-+--il--+--+----+-+-+-

c l--+-+-+-+-+--+--+--+--+----ll---+--+---1---1---lf--l--+-+-+-+-+--ll--+--+--l--l--+-l---+-

B Key:

OwaterHole D 50%WaterHole category 1 Fuel D category 2 Fuel D category 3 Fuel category 4 Fuel category 5 Fuel g category 5 Fuel with a required full lengthRCCA

[!] Blocked cell 40 41 42 43 "

45 46 n u 49 sn s1 s1 54 ss 56 s1 58 s, 60,1 &2 n

64 Figure 3.7.13-1 (page 1 of 2)

I Cask Area Allowable Spent Fuel Pit Storage Locations for Category 1 through Category 5 Fuel Assemblies in Regions 1 and 2 DP DN DM DL DK DJ DH DG DF DE CP CN CM Cl CK CJ CH CG CF CE co BN BM BL BK Bl BH BG BF BE 8D BC

INSERT 3 (Continued)

-Notes-

1.

Fuel assembly Categories are ranked in order of relative reactivity, from Category 1 to 5. Category 1 fuel assemblies have the highest reactivity, and Category 5 fuel assemblies have the lowest.

2.

Fuel assembly categorization for assembly IDs after X for IP2 and after AA for IP3 must be performed in accordance with Table 3. 7.13-1.

3.

Fuel assembly Categories for IP2 assembly IDs A through X are located in Table 3.7.13-2.

4.

Fuel assembly Categories for IP3 assembly IDs A through AA are located in Table 3.7.13-3.

5.

Fuel assemblies of any higher numbered Category can be stored in any cell location that allows for a lower numbered Category. For example, a Category 5 fuel assembly can be stored in Category 1, 2, 3, 4, and 5 cells. Any cell may be empty.

6.

Category 1 fuel assemblies that contain a full length RCCA may be stored in any Category 4, 3, 2, or 1 cell.

7.

Category 2, 3 or 4 fuel assemblies that contain a full length RCCA may be stored in any Category 5 cell that does not require an inserted RCCA or in any Category 4, 3, 2, or 1 cell.

8.

A Water Hole may contain up to 50% of absorber material by volume in the active fuel area. Stainless steel and lnconel meet the definition of absorber material. There is no restriction for non-actinide material outside of the active fuel area.

9.

A 50% Water Hole may contain up to 50% of any non-actinide material by volume in the active fuel area.

Zirconium meets the definition of non-actinide material. There is no restriction for non-actinide material outside of the active fuel area.

10.

A Blocked Cell has the same requirements as a Water Hole.

11.

A checkerboard area consists of every other cell being a Water Hole.

12.

An area of Category 1 fuel assemblies may be formed in Region 1. The Category 1 area must be formed by replacing the Region 1 arrangement shown in this figure with an area of Category 1 fuel assemblies in accordance with the following criteria (see examples in Figure 3.7.13-2):

a)

Category 1 fuel assemblies must be face adjacent to at least three Water Holes and not face adjacent to adjacent to another Category 1 assembly.

b)

Category 2 fuel assemblies must not be face adjacent to more than one Category 1 fuel assembly.

c)

Catergory 3 and Category 5 locations in Figure 3.7.13-1 may not be moved.

13.

A checkerboard area of Category 1 fuel assemblies may be formed in Region 2. All four sides of the checkerboard area must be rows of Water Holes.

14.

The edge of Region 2 next to the pool wall or cask loading area can be considered to be a row of Water Holes.

Figure 3.7.13-1 (page 2 of 2)

H G

F E

D C

B H

G F

E D

C B

J H

G F

E D

C B

INSERT4 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 27 28 29 30 31

.__............. ----1--

1----+---+----+-+---+---+----+-+---+--I

>---+---+---+-+---+---+---+-+---+--<

1 2

3 4

s 6

7 8

9 w

u u 13 H

~ u

~ ~ w n n n

M

~ ~ u

~ ~ ~ n 1

2 3

4 5

6 7

Figure 3.7.13-2 Examples of Allowable Spent Fuel Pit Storage Locations for Category 1 Fuel Assemblies in Region 1

INSERT 5 Table 3.7.13-1 Fuel Assembly Reactivity Categorization for Assembly IDs after X for IP2 and after AA for IP3 Reactivity Minimum Required Burnup (MRB) (GWd/T)(a)(b)(cl Category 1

o(d) 2 21 3

28.5 4

81.2 =(al+ a2*E + a3*E2) x exp[-(a4 + aS*E + a6*E2) x CT]+ a7 + a8*E + a9*E2 Ba.a= (bl+ b2*E + b3*E 2) x exp[-(b4 + bS*E + b6*E2) x CT]+ b7 + b8*E + b9*E 2 MRB =Ba.a+ (81.2 - Baa) x (PF - 0.8)/ 0.4 5

MRB for Category 4 plus 11 Where:

and:

E is enrichment in wt% U-235(el, CT is cooling time in years(tl, and PF is the average peaking factor defined by the fuel assembly burnup divided by the sum of the cycle burnups for the cycles the fuel assembly was in the core.

Coefficient Value Coefficient Value al

-6.26824 bl 15.1405 a2 5.29367 b2

-4.81133 a3

-0.37154 b3 0.753855 a4 0.129582 b4 0.121252 as

-0.0204918 bS

-0.0150991 a6 0.00205596 b6 0.00127009 a7

-0.13331 b7

-16.2293 a8 6.9037 b8 14.0159 a9 0.122068 b9

-0.687054 (a) 2 GWd/T must be added to the MRB for any fuel assembly that had a Hafnium insert.

(b) 4 GWd/T must be added to the MRB for any fuel assembly that was reconstituted without replacing removed fuel rods with stainless steel rods.

(c) 0.2, 0.3, 0.6, and 0.9 GWd/T must be added to the MRB for Categories 2, 3, 4, and 5, respectively if the maximum multi-cycle burn up averaged soluble boron concentration of 950 ppm ffif--tRe evae is exceeded.

(d) With 64 IFBA rods or more. Assemblies with enrichments less than or equal to 4.5, 4.0, 3.5, and 3.0 require only 48, 32, 16, and O IFBA rods, respectively.

INSERT 5 (Continued)

(e) Fuel assemblies at enrichments less than 4.2 wt% U-235 must use 4.2 wt% U-235 in the Category 4 equation.

(f) Fuel assemblies with cooling times of more than 25 years must use 25 years in the Category 4 equation.

INSERT 5 (Continued)

Table 3. 7.13-2 (page 1 of 3)

Fuel Assembly Reactivity Categorization for Assembly IDs A through X for IP2 Indian Point Unit 2 Fuel Assembly ID Category Assembly ID Category Assembly ID Category A01-A65 4

E43-E55 4

K01-K13 4

E56 3

K14-K15 5

B01-B07 4

E57-E:60 4

K16-K57 4

B08-B13 5

K58 5

B14-823 4

FOl 3

K59-K68 4

B24-B26 5

F02-F20 4

B27-864 4

F21 3

l01-l07 4

F22-F30 4

L08-L10 5

C01-C04 4

F31-F34 5

l11-L63 4

COS-COG 5

F35 4

L64 3

C07-C12 4

F36 3

L65-L68 4

C13 5

F37-F39 4

C14 4

F40 3

M01-M04 4

C15-C18 5

F41-F49 4

MOS 5

C19-C28 4

FSO 3

M06-M08 4

C29 5

F51-F60 4

M09 5

C30-C64 4

F61 3

M10-M12 4

F62-F64 4

Ml3-M14 5

DOl-025 4

F65 3

M15-M20 4

D26 5

F66 4

M21 5

D27-060 4

F67-F68 5

M22-M23 4

D61-D68 5

M24 5

069-072 4

GOl-GOS 4

M25-M27 4

GOG 5

M28 5

E01-E14 4

G07-G37 4

M29-M30 4

ElS 3

G38 5

M31 5

E16-E19 5

G39-G72 4

M32-M34 4

E20 4

M35 5

E21-E24 5

H01-H38 4

M36-M37 4

E25-E27 4

H39-HS1 5

M38-M44 5

E28-E31 5

H52-H54 4

M45 3

E32-E33 4

HSS 5

M46 4

E34-E35 5

HSG 4

M47-M48 5

E36-E40 4

M49-MSO 4

E41-E42 5

J01-J68 4

M51-M52 5

INSERT 5 (Continued)

Table 3. 7.13-2 (page 2 of 3)

Fuel Assembly Reactivity Categorization for Assembly IDs A through X for IP2 Indian Point Unit 2 Fuel Assembly ID Category Assembly ID Category Assembly ID Category M53-M54 4

Q71-Q73 4

T42-T43 4

M55-M56 5

Q74-Q76 5

T44-T46 5

M57 4

Q77 4

T47 4

M58-M59 5

Q78 5

T48 5

M60 4

Q79-Q80 4

T49-T51 4

M61 3

T52-T53 5

M62-M63 4

R01-R07 5

T54 4

M64 3

ROS 4

TSS 5

M65 4

R09-R38 5

T56-T72 4

M66 5

R39 4

T73-T80 5

M67 3

R40-R43 5

M68 5

R44-RSO 4

M69-M71 4

R51-R69 5

U01-U04 5

M72 5

R70 4

uos 4

R71-R72 5

U06-U13 5

N01-N08 4

R73-R74 4

U14 4

N09-N12 5

R75-R79 5

U15-U16 5

N13-N14 4

R80-R81 4

U17-U21 4

N15-N16 5

R82 5

U22 5

N17-N23 4

R83-R85 4

U23 4

N24-N32 5

U24-U49 5

N33-N47 4

S01-S44 5

USO 4

N48 5

S45 4

USl 5

N49-N80 4

S46-S47 5

U52 4

S48 4

U53-U61 5

P01-P02 4

S49-S61 5

U62-U64 4

P03 3

S62 4

U65 5

P04-P47 4

S63-S65 5

U66-U68 4

P48 5

S66 4

U69-U73 5

P49-P60 4

S67-S77 5

P61-P72 5

V01-V16 5

V17-V29 4

Q01-Q65 5

T01-T32 5

V30-V35 5

Q66 4

T33-T34 4

V36 4

Q67-Q68 5

T35-T36 5

V37-V38 5

Q69 4

T37 3

V39 4

Q70 5

T38-T41 5

V40-V41 5

INSERT 5 (Continued)

Table 3. 7.13-2 (page 3 of 3)

Fuel Assembly Reactivity Categorization for Assembly IDs A through X for IP2 Indian Point Unit 2 Fuel Assembly ID Category Assembly ID Category Assembly ID Category V42-V43 4

W21 5

X01-X02 3

V44-V49 5

W22 4

X03-X04 5

vso 4

W23 5

XOS-X37 4

V51-V54 5

W24 4

X38 5

V55-VS7 4

W25 5

X39-X49 4

V58-V61 5

W26 4

XSO-XSl 5

V62 4

W27 5

X52-X53 4

V63 5

W28-W34 4

X54-XSS 5

V64-V65 4

W35 5

XS6-XS8 4

V66-V67 5

W36-W38 4

X59-X60 5

V68 4

W39 5

X61-X62 4

V69-V77 5

W40 4

X63 5

V78-V79 4

W41-W43 5

X64-X65 4

V80-V81 5

W44-W45 4

X66 5

V82 4

W46 5

X67 4

V83 5

W47 4

X68-X69 5

V84 4

W48-W49 5

X70-X73 4

V85 5

wso 4

X74 5

V86 4

WSl 5

X75 4

V87-V88 5

W52-WSS 4

X76 5

V89 4

W56-W58 5

X77 4

V90-V91 5

WS9-W60 4

X78 5

V92 4

W61 5

X79 4

W62 4

X80-X93 5

W01-W10 4

W63-W67 5

X94-X95 4

Wll 5

W68 4

X96 5

W12-W15 4

W69-W71 5

W16 5

W72 4

FRSB1 4

W17 4

W73-W83 5

W18-W19 5

W84 4

W20 4

W85-W93 5

1 FRSB is the Fuel Rod Storage Basket

INSERT 5 (Continued)

Table 3.7.13-3 Fuel Assembly Reactivity Categorization for Fuel Assembly IDs A through AA for IP3 Indian Point Unit 3 Fuel Assembly ID I Category I Assembly ID I Category I Assembly ID I Category V43 I

3 I

V48 I

3 I

I All other Fuel Assembly IDs A through AA are Category 4

4.0 DESIGN FEATURES 4.1 Site Location Design Features 4.0 Indian Point 2 is located on the East bank of the Hudson River at Indian Point, Village of Buchanan, in upper Westchester County, New York. The site is approximately 24 miles north of the New York City boundary line. The nearest city is Peekskill which is 2.5 miles northeast of Indian Point.

The minimum distance from the reactor center line to the boundary of the site exclusion area and the outer boundary of the low population zone, as defined in 1 O CFR 100.3, is 520 meters and 1100 meters, respectively. For the purpose of satisfying 1 O CFR Part 20, the "Restricted Area" is the same as the "Exclusion Area" shown in UFSAR, Figure 2.2-2.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircalloy-4 or ZIRLO fuel rods. Fuel shall have a U-235 enrichment of

5.0 weight percent.

Limited substitutions of Zircalloy-4, ZIRLO or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used.

Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases.

A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

4.2.2 Control Rod Assemblies The reactor core shall contain 53 control rod assemblies.

The control rod material shall be silver indium cadmium, clad with stainless steel, as approved by the NRC.

4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

and poisons, if necessary, to meet the limit for kett, INDIAN POINT 2

a.

Fuel assemblies having a maximum U-235 enrichment of 5.0 wei ht 4.0 - 1 Amendment No.

Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued) categorized

b. __ kett < 1.0 if fully flooded with un orated water, and

~----------==::;--...-r kett ::. 0.95 when flooded

c.

Each fuel assembly e based on initial enrichment, burnup,

.__w_it_h_b_o_r_at_e_d_w_a_t_e_r, ___

cooling~

and number of Integral Fuel Burnable Absorbers

~

rods with individual fuel assembly storage location within

, averaged assembly peaking

factor, the spent fuel storage rack restricted as required by Technical Specification 3.7.13.

4.3.2 4.3.3 4.3.1.2 Drainage The new fuel storage racks are designed and shall be maintained with:

a.

Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent, and poisons, if necessary, to meet the limit for

kett,

b.

kett ~ 0.95 if fully flooded with unborated water, and

c.

A 20.5 inch center to center distance between fuel assemblies placed in the storage racks to meet the limit for ke11*

The spent fuel pit is designed and shall be maintained to prevent inadvertent draining of the pit below a nominal elevation of 88 feet, 6 inches.

Capacity The spent fuel pit is designed and shall be maintained with a storage capacity limited to no more than 269 fuel assemblies in Region I and 1105 fuel assemblies in Region II.

INDIAN POINT 2 4.0- 2 Amendment No.

Enclosure Attachment 2 NL-19-023 Technical Specification Bases Page Markups (Information Only)

Spent Fuel Pit Boron Concentration B 3.7.12 B 3.7 PLANT SYSTEMS B 3.7.12 Spent Fuel Pit Boron Concentration BASES BACKGROUND INSERT 1 APPLICABLE SAFETY ANALYSES INSERT 2 INDIAN POINT 2 The Spent Fuel Pit (SFP) is used to store spent fuel removed from the reactor and new fuel ready for insertion into the reactor. The SFP has been evaluated to meet the requirements of option (b) of 1 O CFR 50.68, "Criticality Accident Requirements" (Ref. 1 ). IP2 compliance with 1 O CFR 50.68(b)(4) was confirmed by an analysis docu No1theast 2 Spent Fuel Storage Reeks" (Ref. 2). This analysis demonstrated that 10 CFR 50.68(b)(4) will be met during normal SFP operation and all credible accident scenarios (including the affects of boraflex degradation) if the following requirements are met:

a) b)

Spent Fuel Pit boron concentration is maintained within the limits of LCO 3.7.12, "Spent Fuel Pit Boron Concentration," whenever fuel is stored in the SFP; and, Fuel assembly storage location within the Spent Fuel Pit is restricted in accordance with LCO 3. 7.13, "Spent Fuel Pit Storage..,_" based on the fuel assembly's initial enrichment, burnup, deeay of Plutonium 241 (i.e., cooling time), and number of Integral Fuel Burnable Absorbers (I FBA) rods.

A detailed description of how this combination of minimum boron concentration and restrictions on fuel assembly storage location is presented in the Bases for LCO 3. 7.13.

-173-01, "Criticality Analysis for Soluble Boron and Burnup Credit in the Edison Indian Point Unit No. 2 Spent Fuel Storage Racks" (Ref. 2) ev ated non-accident conditions in the SFP including the affects of the p

• cted boraflex degradation through the year 2006.

Based upon BADG esting in calendar years 2003, 2006 and 201 O and RACKLIFE code proje

• ns, the validity of the criticality and boron dilution analysis documented r eferences 2, 3, 5 and 6 can be extended through the end of the cur license (September 28, 2013).

Reference 7 allowed BADGER testing to be rformed in 2013, to confirm the progression of localized Boraflex dissolution.

e continued validity of the criticality and boron dilution analysis will be ve *

• d based on the boron monitoring program as defined in the License Renew plication.

Reference 2 determined that if storage location requirements in

• LCO are met then the B3.7.12-1 Revision

BASES Spent Fuel Pit Boron Concentration B 3.7.12 APPLICABLE SAFETY ANALYSES (continued)

INDIAN POINT 2 FP will have a keff of~ 0.95 if filled with a soluble boron concentration of

~ 86 ppm and will have a keff of < 1.0 if filled with unborated water.

Refer nee 2 also evaluated credible abnormal occurrences in accordance with A I/ ANS-57.2-1983. This evaluation considered the effects of the following. a) a dropped fuel assembly or an assembly placed alongside a rack; b) misloaded fuel assembly; and, c) abnormal heat loads.

Reference 2 etermined that the SFP will maintain a keff of ~ 0.95 under the worst-cas accident scenario if the SFP is filled with a soluble boron 1495 ppm.

NET-173-02, "India Point Unit 2 Spent Fuel Pool (SFP) Boron Dilution Analysis" (Ref. 3) e luated postulated unplanned SFP boron dilution scenarios assuming an *nitial SFP boron concentration within the limits of LCO 3. 7.12. The evalu ion considered various scenarios by which the SFP boron concentration y be diluted and the time available before the minimum boron concentrati necessary to ensure subcriticality for the non-accident condition (i.e. it

• not assumed an assembly is misloaded concurrent with the Spent el Pit dilution event).

Reference 3 determined that an unplanned o inadvertent event that could dilute the SFP boron concentration from 20 O ppm to 786 ppm is not a credible event because of the low frequenc of postulated initiating events and because the event would be readily detected and mitigated by plant personnel through alarms, flooding, and perator rounds through the SFP area.

References 2 and 3 are based on conservat1 e projections of amount of Boraflex absorber panel degradation assumed I each sub-region. These projections are valid through the end of the ar 2006. Based upon BADGER testing in calendar years 2003, 2006 an 2010 and RACKLIFE code projections, the validity of the criticality and b on dilution analysis documented in References 2, 3, 5 and 6 can be ex nded through the end of the current license (September 28, 2013). Ref ence 7 allowed BADGER testing to be performed in 2013, to confirm the regression of localized Boraflex dissolution. The continued validity of the riticality and boron dilution analysis will be verified based on the boro monitoring program as defined in the License Renewal Applicatio. These compensatory measures for boraflex degradation in the S evaluated by the NRC in Reference 4.

The concentration of dissolved boron in Criterion 2 of 1 O CFR 50.36 (c)(2)(ii).

B 3.7.12 - 2 Revision

BASES LCO APPLICABILITY ACTIONS SURVEILLANCE REQUIREMENTS INDIAN POINT 2 Spent Fuel Pit Boron Concentration B 3.7.12 The Spent Fuel Pit boron concentration is required to be

~ 2000 ppm. The specified concentration of dissolved boron in the Spent Fuel Pit preserves the assumptions used in the analyses of the potential critical accident scenarios as described in Reference 2.

This concentration of dissolved boron is the minimum required concentration for fuel assembly storage and movement within the Spent Fuel Pit.

This LCO applies whenever fuel assemblies are stored in the Spent Fuel Pit.

A.1 and A.2 The Required Actions are modified by a Note indicating that LCO 3.0.3 does not apply.

When the concentration of boron in the Spent Fuel Pit is less than required, immediate action must be taken to preclude the occurrence of an accident or to mitigate the consequences of an accident in progress.

This is most efficiently achieved by immediately suspending the movement of fuel assemblies. The concentration of boron is restored simultaneously with suspending movement of fuel assemblies.

If the LCO is not met while moving irradiated fuel assemblies in MODE 5 or 6, LCO 3.0.3 would not be applicable.

If moving irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the fuel movement is independent of reactor operation.

Therefore, inability to suspend movement of fuel assemblies is not sufficient reason to require a reactor shutdown.

SR 3.7.12.1 This SR verifies that the concentration of boron in the Spent Fuel Pit is within the required limit.

As long as this SR is met, the analyzed accidents are fully addressed.

The 7 day Frequency is appropriate because no major replenishment of Spent Fuel Pit water is expected to take place over such a short period of time.

B 3.7.12 - 3 Revision

BASES REFERENCES INSERT 3 INDIAN POINT 2

1.

2.

3.

Spent Fuel Pit Boron Concentration B 3.7.12 1 O CFR 50.68, "Criticality Accident Requirements."

ortheast Technology Corporation report NET-173-01, "Criticality An sis for Soluble Boron and Burnup Credit in the Con Edison India oint Unit No. 2 Spent Fuel Storage Racks."

Northeast chnology Corporation report N ET-173-02, "Indian Point Unit 2 S nt Fuel Pool (SFP) Boron Dilution Analysis."

4.

Safety Evaluation 6 the Office of Nuclear Reactor Regulation Related to Amendmen

o. 227 to Facility Operating License No.

DPR-26, May 29, 2002.

5.

NETCO Letter to M. R. Hans r from E. Lindquist, Northeast Technology Corp. dated 12/19/06, ubject - Reference 2 and 3 extension.

6.

NETCO Letter to Floyd Gumble from Matt Harris dated 12/22/2009, titled "Indian Point 2 RACKLIFE jections Through 2010 and 2012 BADGER Tests with RACKLIFE sion 2.1"

7.

NETCO Letter to Giancarlo Delfini from Matt Ha *s 12/12/2012, titled "Update of IP2 RACKLIFE Model -

(I fulfillment of Entergy Contract 10351857, Change Order Task-2A)".

B 3.7.12 - 4 Revision

INSERT 1 Curtiss-Wright Nuclear Division, NETCO report NET-28091-003-01, Rev. O "Criticality Safety Analysis for the Indian Point Unit 2 Spent Fuel Pool with No Absorber Panel Credit."

INSERT2 Curtiss-Wright Nuclear Division, NETCO report NET-28091-003-01, Rev. O "Criticality Safety Analysis for the Indian Point Unit 2 Spent Fuel Pool with No Absorber Panel Credit" (Ref. 2) evaluated non-accident conditions in the SFP. Reference 2 determined that if storage location requirements in this LCO are met then the SFP will have a kett of

0.95 if filled with a soluble boron concentration of ~ 700 ppm and will have a kett of <

1.0 if filled with unborated water.

Reference 2 evaluated abnormal occurrences and accidents.

This evaluation considered the effects of the following: a) multiple misleads, b) an assembly placed alongside a rack; c) a dropped assembly, d) a misloaded assembly; e) SFP over temperature, f) a seismic event, and, g) a SFP boron dilution accident. Reference 2 determined that the most limiting fuel handling accident is the multiple mislead accident and determined keff of :::; 0.95 if the SFP is filled with a soluble boron concentration of

2'. 2000 ppm.

NET-173-02, "Indian Point Unit 2 Spent Fuel Pool (SFP) Boron Dilution Analysis" (Ref.

3) evaluated postulated unplanned SFP boron dilution scenarios assuming an initial SFP boron concentration within the limits of LCO 3.7.12. The evaluation considered various scenarios by which the SFP boron concentration may be diluted and the time available before the minimum boron concentration necessary to ensure subcriticality for the non-accident condition (i.e. it is not assumed an assembly is misloaded concurrent with the Spent Fuel Pit dilution event).

Reference 3 determined that an unplanned or inadvertent event that could dilute the SFP boron concentration from 2000 ppm to 786 ppm is not a credible event because of the low frequency of postulated initiating events and because the event would be readily detected and mitigated by plant personnel through alarms, flooding, and operator and security rounds through the SFP area.

The concentration of dissolved boron in the spent fuel pit satisfies Criterion 2 of 1 O CFR 50.36 (c)(2)(ii).

INSERT 3

2.

Curtiss-Wright Nuclear Division, NETCO report NET-28091-003-01, Rev. O "Criticality Safety Analysis for the Indian Point Unit 2 Spent Fuel Pool with No Absorber Panel Credit."

3.

Northeast Technology Corporation report NET-173-02, "Indian Point Unit 2 Spent Fuel Pool (SFP) Boron Dilution Analysis."

B 3. 7 Plant Systems Spent Fuel Pit Storage B 3.7.13 B 3.7.13 Spent Fuel Pit Storage BASES BACKGROUND INDIAN POINT 2 n issue has been identified with the degradation of boraflex used in the s

nt fuel pool to meet the licensing basis. To address this degradation Pro dure O-NF-203R16, Attachment 3, Transfer Form Checklist, discu es the administrative controls used to mitigate the effects of the boraflex egradation.

The Spent Fue it (SFP) is used to store spent fuel removed from the reactor and new f I ready for insertion into the reactor. Spent fuel racks (SFRs) are erected n the SFP floor to hold the fuel assemblies. The SFRs have been eva ated to meet the requirements of option (b) of 1 O CFR 50.68, "Critical!

Accident Requirements" (Ref. 1) when: a)

Spent Fuel Pit boron cone tration is maintained within the limits of LCO 3.7.12, "Spent Fuel Pit Bo n Concentration," and, b) fuel assembly storage location within the Sp t Fuel Pit is restricted in accordance with LCO 3.7.13, "Spent Fuel Pit Sto e," based on the fuel assembly's initial enrichment, burnup, decay of P tonium-241 (i.e., cooling time), and number of Integral Fuel Burnable Ab rbers (IFBA) rods.

In 1990, Spent Fuel Pit storage capac1 980 fuel assemblies to 1376 fuel assemblies by e installation of high-density racks that reduced the distance between ad cent fuel assemblies. This was possible because the k-effective of the P was maintained within the limits of 1 O CFR 50.68(b) (Ref. 1) by the llowing: 1) the use of boraflex absorber panels (i.e., neutron absorbers between SFR cells; and, 2) restrictions on fuel assembly storage loca *on within the SFP based on initial enrichment and burnup. The original esign of the high density racks met the requirements of 1 O CFR 50.68(b) ithout crediting soluble boron.

The use of high-density SFRs that depend on boraflex absor r panels between cells requires that IP2 adhere to a long-term inspection rogram to monitor the performance of the boraflex panels. Requirements r the boraflex inspection program are specified in I P2 Amendment 150 (Re 2) and Generic Letter 96-04, "Boraflex Degradation in Spent Fuel P I

Storage Racks" (Ref. 3).

83.7.13-1 Revision

BASES Spent Fuel Pit Storage 83.7.13 BACKGROUND (continued)

INDIAN POINT 2 uring an inspection of the SFRs in 2000, it was determined that the a sumptions regarding the boraflex panels used in the criticality analysis for the SFP were no longer valid because of thinning and gaps in the bora ex panels. This degradation of the boraflex panels between SFR cells quired that IP2 adopt the "use of soluble boron" option in 10 CFR 50.68(0 4) which specifies that:

"... If credit is taken for soluble boron, the k-effective of the spent fuel orage racks loaded with fuel of the maximum fuel assembly reactiv must not exceed 0.95, at a 95 percent probability, 95 perc t confidence level, if flooded with borated water, and the k-effective must remain below 1.0 (subcritical), at a 95 percent probability, 5 percent confidence level, if flooded with unborated water."

Based on the results an inspection and analysis that conservatively projected the condition o the boraflex panels through the end of 2006, IP2 compliance with 10 C R 50.68(b)(4) was confirmed by an analysis documented in Northeast chnology Corporation report NET-173-01, "Criticality Analysis for Solu e Boron and Burnup Credit in the Con Edison Indian Point Unit No.

Spent Fuel Storage Racks" (Ref. 4).

Based upon BADGER testing in c lendar years 2003, 2006 and 201 O and RACKLIFE code projections, the validity of the criticality and boron dilution analysis documented in R ferences 4, 5, 7 and 10 can be extended through the end of the cur nt license (September 28, 2013).

Based on Reference 11, BADGER te ting was performed in 2013, to confirm the progression of localized Sor flex dissolution. The continued validity of the criticality and boron dilution nalysis will be verified based on the boron monitoring program as defin d in the License Renewal Application. This analysis demonstrated that O CFR 50.68(b)(4) will be met for all normal and credible accident s narios if the following requirements are met:

a)

Spent Fuel Pit boron concentration is mainta* ed within the limits of LCO 3.7.12, "Spent Fuel Pit Boron Gonce fuel is stored in the SFP; and, b)

Fuel assembly storage location within the Spe Fuel Pit is restricted in accordance with LCO 3.7.13, "Spe t Fuel Pit Storage," based on the fuel assembly's initial enrichme t, burnup, decay of Plutonium-241 (i.e., cooling time), and number Integral Fuel Burnable Absorbers (IFBA) rods.

Fuel assembly storage location within the Spent Fuel Pit is an ess element for the validity of the analysis because the storage racks in areas designated Region 1 have a different design than the storage rac B3.7.13-2 Revision

BASES Spent Fuel Pit Storage B 3.7.13 BACKGROUND (continued)

INDIAN POINT 2 areas designated Region 2.

These design differences have a sr nificant impact on criticality calculations. Additionally, each of the two reg* ns is sub-divided into two parts based on the extent of the boraflex degr dation. Therefore, the SFP is divided into four distinct regions based on rac design and boraflex degradation. Figure 3. 7.13-5 identifies the four re *ons as Region 1-1, Region 1-2, Region 2-1 and Region 2-2.

Additiona, selected cells located on the perimeter of Region 2-2 have higher neu on leakage rates than other cells in the Region and are designated a "peripheral" cells.

Each SFP regi n and sub-region is shown in Figure 3. 7.13-5 and is described below eginning with the region that can be used to store only the least reactive el and ending with the region that must be used to store the most reacti fuel.

Region 2, consisting o nine racks that use the egg-crate design, can store 1105 fuel assembli and two failed fuel canisters. Region 2 racks consist of boxes welded

• to a "checkerboard" array with a storage location in each square. On Boraflex absorber panel is held to one side of each cell wall by picture rame sheathing.

Region 2 racks were originally designed to store uel assemblies that have undergone significant burnup (e.g., ~ 5.0 wei ht percent (w/o} U235 with a burnup of at least 40,900 megawatt days perm ric ton (MWD/MT)) or fuel assemblies with a relatively low initial enrichme t and low burnup (i.e., ~ 1.764 w/o U235 at zero burnup).

Region 2 is subdivided into two regions (

Region 2-1 is assumed to have susta* ed a 100% loss of Boraflex (i.e., none of the boraflex in the nels is assumed to be available}. Figure 3.7.13-1 shows the el assembly criteria that will meet the requirements of 10 CFR 0.68(b}(4) if stored in Region 2-1. As shown on Figure 3.7.13-the maximum initial enrichment that can be stored in Region 2-with no burnup is 1.06 w/o U235.

Figure 3.7.13-1 shows an al wance permitting storage of fuel assemblies with higher initial enri ments based on the reactivity reduction due to the cumulative bu up of the fuel assembly in the core and the decay of Pu241 (expres ed as cooling time) after a fuel assembly is discharged.

Region 2-2 is assumed to have sustained only a 30 Boraflex (i.e., 70% of the boraflex in the panels is assum d to be available). Figure 3.7.13-2 shows the fuel assembly criter that will meet the requirements of 1 O CFR 50.68(b)(4) if stor in Region 2-2. As shown on Figure 3.7.13-2, the maximum in ial enrichment that can be stored in Region 2-2 with no burnup B 3.7.13 - 3 Revision

BASES Spent Fuel Pit Storage 83.7.13 BACKGROUND (continued)

INDIAN POINT 2 1.80 w/o U235. Additionally, Figure 3.7.13-2 shows an allowance permitting storage of fuel assemblies with higher initial enrichments based on the reactivity reduction due to the umulative burnup of the fuel assembly in the core and the decay

( pressed as cooling time) of Pu241 after a fuel assembly is di harged.

Region 1, c sisting of three racks that use the flux trap design, can store 269 new r irradiated fuel assemblies. The flux trap design used in Region 1

uses spacer plate in the axial direction to separate the cells. Boraflex absorber panels ar held in place adjacent to each side of the cell by picture-frame sheath1 g. The spacer plates between cells form a flux trap between the boraflex bsorber panels. Region 1 racks were originally designed to store new fu I with enrichments up to 5.0 w/o U235*

Region 1 is subdivided into o regions (Region 1-1 and Region 1-2):

Region 1-1 is assume o have sustained a 100% loss of Boraflex (i.e., none of the bora ex in the panels is assumed to be available). Figure 3.7.13-shows the fuel assembly criteria that will meet the requirements of 10 CFR 50.68(b)(4) if stored in Region 1-1. As shown on

• ure 3.7.13-3, the maximum initial enrichment that can be store in Region 1-1 with no burnup is 1.95 w/o U235. Additionally, Figu 3.7.13-3 shows an allowance permitting storage of fuel as emblies with higher initial enrichments based on the reac *vity reduction due to the cumulative burn up of the fuel assemb in the core. Figure 3. 7.13-3 does not provide any allowance from he minimum required fuel assembly burnup based on the decay of u241 *

(Fuel assemblies that do not meet the crite ia in Figure 3.7.13-3 may be stored in Region 1-1 if the followin two conditions are met: a) the fuel assemblies are stored in a ch kerboard loading configuration (1 out of every two cells with very other cell vacant); and, b) fuel assemblies meet the criteria o Figure 3.7.13-4.)

Region 1-2 is assumed to have sustained a 50% loss f Boraflex (i.e., 50% of the boraflex in the panels is assume available). Region 1-2 can accommodate unirradiated fu up to 5.0 w/o U235 assuming the presence of a minimum number o FBA rods as specified in Figure 3.7.13-4. As shown on Figure 3. 13-4, the maximum initial enrichment that can be stored in Region 2

when there are no I FBA rods is 4.50 w/o U235*

Figure 3. 7.13-83.7.13-4 Revision

BASES Spent Fuel Pit Storage B 3.7.13 BACKGROUND (continued)

APPLICABLE SAFETY ANALYSES INSERT 2 INDIAN POINT 2 does not provide any allowance from the minimum required IFBA rods based on the decay of PU241

• Peri hera " Cells, consisting of six select cells along the SFP west wall in Region 2-2, re shown in Figure 3. 7.13-5. These six "peripheral" cells may be used to s e fuel that meets the requirements for storage in any other location in the P. Cells between and adjacent to the "peripheral" cells may be filled with el assemblies that meet the requirements of Figure 3.7.13-2 (i.e., meet requirements for storage in Region 2-2).

The two prematurely discharge uel assemblies meet the requirements of Figure 3. 7.13-4 and qualify for sto e in the "peripheral" cells.

I P3 Fuel Assemblies The SFP is also used to store spent fuel transfe d from the IP3 SFP.

The IP3 fuel assembly storage location is also restn ed in accordance with LC03.7.13 that limits IP3 fuel assemblies to Regio

-2 of the IP2 SFP. The NRC has issued Amendment 268 for the inter-un ransfer of spent nuclear fuel (Ref. 8). The Amendment is based on ev ations conducted for each aspect of the inter-unit transfer of fuel as docume in Reference 9.

As req

• ed by 1 O CFR 50.68, "Criticality Accident Requirements" (Ref.

1 ), if the S nt Fuel Pit takes credit for soluble boron, then "the k-effective of the spent el storage racks loaded with fuel of the maximum fuel assembly reactiv1 must not exceed 0.95, at a 95 percent probability, 95 percent confidence I I, if flooded with borated water, and the k-effective must remain below 1.

subcritical), at a 95 percent probability, 95 percent confidence level, if oded with unborated water."

NET-173-01, "Criticality Analysis Soluble Boron and Burnup Credit in the Con Edison Indian Point Unit No.

Spent Fuel Storage Racks" (Ref.

4) and NET-173-02, "Indian Point Unit pent Fuel Pool (SFP) Boron Dilution Analysis," (Ref. 5) determined that O CFR 50.68(b)(4) will be met during normal SFP operation and all ere 'ble accident scenarios (including the affects of boraflex degradation) if: a ent Fuel Pit boron concentration is maintained within the limits of LCO..12, "Spent Fuel Pit Boron Concentration," and, b) fuel assembly storage cation within the Spent Fuel Pit is restricted based on the fuel asse ly's initial enrichment, burnup, decay of Pu241 (i.e., cooling time) and n ber of Integral Fuel Burnable Absorbers (IFBA) rods.

B 3.7.13 - 5 Revision

BASES Spent Fuel Pit Storage 83.7.13 APPLICABLE SAFETY ANALYSES (continued)

INDIAN POINT 2 eference 4 evaluated non-accident conditions in the SFP including the a ects of the projected boraflex degradation through the year 2006.

Ba ed upon BADGER testing in calendar years 2003, 2006 and 201 O and RA LIFE code projections, the validity of the criticality and boron dilutio analysis documented in References 4, 5, 7 and 1 O can be extende through the end of the current license (September 28, 2013).

Based o Reference 11 BADGER testing was performed in 2013, to confirm th progression of localized Boraflex dissolution. The continued validity of th criticality and boron dilution analysis will be verified based on the boron monitoring program as defined in the License Renewal Application. Re rence 4 determined that if storage location requirements in this LCO are t then the SFP will have a ke of :::; 0.95 if filled with a soluble boron concentration of

~ 786 ppm and will H ve a keff of< 1.0 if filled with unborated water.

Reference 4 also evalu ed credible abnormal occurrences in accordance with ANSI/ ANS-57.2-198. This evaluation considered the effects of the following: a) a dropped fue assembly or an assembly placed alongside a rack; b) a misloaded fuel assembly; and, c) abnormal heat loads.

Reference 4 determined that e SFP will maintain a keff of :::; 0.95 under the worst-case accident scenar if the SFP is filled with a soluble boron concentration of~ 1495 ppm.

Therefore, reference 4 confirmed th t the requirements in 1 O CFR 50.68, "Criticality Accident Requirements," ( ef. 1) will be met for both normal SFP operation and credible abnormal o currences if:

a)

Spent Fuel Pit boron concentration is maintained within the limits of LCO 3.7.12, "Spent Fuel Pit Bor n Concentration," whenever fuel is stored in the SFP; and, b)

Fuel assembly storage location within the Spent Fuel Pit is restricted in accordance with LCO 3. 13, "Spent Fuel Pit Storage," based on the fuel assembly's initia enrichment, burnup, decay of Plutonium-241 (i.e., cooling time), an number of Integral Fuel Burnable Absorbers (IFBA) rods.

Reference 5 evaluated postulated unplanned oron dilution scenarios assuming an initial SFP boron concentration with1 the limits of LCO 3.7.12. The evaluation considered various scenarios which the SFP boron concentration may be diluted and the time available efore the minimum boron concentration necessary to ensure subcriticali for the non-accident condition (i.e. it is not assumed an assembly is mi oaded concurrent with the Spent Fuel Pit dilution event).

Refere e 5 determined that an unplanned or inadvertent event that could dilute the SFP boron concentration from 2000 ppm to 786 ppm is not a credi le event because of the low frequency of postulated initiating events an B 3.7.13-6 Revision

BASES Spent Fuel Pit Storage 8 3.7.13 APPLICABLE SAFETY ANALYSES (continued)

INDIAN POINT 2 ecause the event would be readily detected and mitigated by plant p sonnel through alarms, flooding, and operator rounds through the SFP Referen e 4 and 5 are based on conservative projections of amount of Boraflex a sorber panel degradation assumed in each sub-region. These projections re valid through the end of the year 2006.

These compensate measures for boraflex degradation in the SFP were evaluated by th NRC in Reference 6. Based upon BADGER testing in calendar years 2 03, 2006 and 2010 and RACKLIFE code projections, the validity of the iticality and boron dilution analysis documented in References 4, 5, 7 an 1 O can be extended through the end of the current license (September 28, 013). Based on Reference 11, BADGER testing was performed in 2013, confirm the progression of localized Boraflex dissolution. The continue validity of the criticality and boron dilution analysis will be verified ba d on the boron monitoring program as defined in the License Renewa pplication.

IP3 Fuel Assemblies An analysis, documented in Reference evaluated the effect of modeling IP3 integral and discrete burnable absoro rs on reactivity in the IP2 spent fuel pool using current methodologies. A r ctivity bias was determined.

In order to offset this bias, and maintain e validity of the IP2 SFP criticality analysis, it was determined that IP fuel assemblies can be stored in the IP2 SFP with the following restriction :

a.

I P3 fuel assemblies shall be stored in Re Spent Fuel Pit, and

b.

The fuel assembly initial enrichment ~ 3.2 U235,and

c.

The fuel assembly discharge Cycle > 1 and ~ 11.

The configuration of fuel assemblies in the Spent Fuel Pit Criterion 2 of 10 CFR 50.36(c)(2)(ii).

83.7.13-7 Revision

BASES LCO INDIAN POINT 2 Spent Fuel Pit Storage B3.7.13 Thi LCO establishes restrictions on fuel assembly storage location within the P to ensure that the requirements of 1 O CFR 50.68 are met. This LCO r quires that each fuel assembly stored in the Spent Fuel Pit is classifie in accordance with Figure 3.7.13-1, Figure 3.7.13-2, Figure 3.7.13-3, nd Figure 3.7.13-4, based on initial enrichment, burnup, cooling tim and number of Integral Fuel Burnable Absorbers (IFBA) rods; and, that fu I assembly storage location within the Spent Fuel Pit is restricted to R gions identified in Figure 3.7.13-5 as follows:

a. Fuel assembli s that satisfy requirements of Figure 3.7.13-1 may be stored in any lo ation in Region 2-1, Region 2-2, Region 1-2 or Region 1-1.

As shown on Figure.7.13-1, the maximum initial enrichment that can be stored in Region 1 with no burnup is 1.06 w/o U235. Additionally, Figure 3.7.13-1 show an allowance permitting storage of fuel assemblies with higher *nitial enrichments based on the reactivity reduction due to the cumu tive burnup of the fuel assembly in the core and the decay of Pu241 afte a fuel assembly is discharged (expressed as cooling time).

b. Fuel assemblies that satisfy re irements of Figure 3. 7.13-2 may be stored in any location in Region 2-, Region 1-2 or Region 1-1.

As shown on Figure 3.7.13-2, them imum initial enrichment that can be stored in Region 2-2 with no burn is 1.80 w/o U235

• Additionally, Figure 3.7.13-2 shows an allowanc permitting storage of fuel assemblies with higher initial enrichme ts based on the reactivity reduction due to the cumulative burnup of t e fuel assembly in the core and the decay (expressed as cooling ti

) of Pu241 after a fuel assembly is discharged.

c. Fuel assemblies that satisfy requirements of Fi re 3.7.13-3 may be stored in any location in Region 1-2 or Region 1-1.

As shown on Figure 3.7.13-3, the maximum initial en *chment that can be stored in Region 1-1 with no burnup is 1.95 w/o U2 5. Additionally, Figure 3.7.13-3 shows an allowance permitting st age of fuel assemblies with higher initial enrichments based on t e reactivity reduction due to the cumulative burnup of the fuel asse bly in the core.

Figure 3.7.13-3 does not provide any allowance rom the minimum required fuel assembly burnup based on the decay o u241.

(Fuel assemblies that do not meet the criteria in Figure 3.7.13-3 be stored in Region 1-1 if the fuel assemblies are stored i a

checkerboard loading configuration (1 out of every two cells with ever B 3.7.13-8 Revision

BASES APPLICABILITY ACTIONS Spent Fuel Pit Storage 83.7.13 vacant) and fuel assemblies meet the criteria of Figure

d. Fuel as emblies that satisfy requirements of Figure 3.7.13-4 may be stored as ollows:

1) In any location in Region 1-2; or, 2) In a checkerboar loading configuration (1 out of every two cells with every other cell vac t) in Region 1-1; or, 3) In locations designated as "peripheral" cells Region 2-2 of Figure 3. 7.13-5.

As shown on Figure 3..13-4, the maximum initial enrichment that can be stored in Region 1-2 ith when there are no IFBA rods is 4.50 w/o U235*

Figure 3.7.13-4 do s not provide any allowance from the minimum required IFBA rods sed on the decay of Pu241.

The six "peripheral" cells may be sed to store fuel that meets the requirements for storage in any lo tion in the SFP (i.e., meets requirements for storage in Region 1-1, 2, 2-1 or 2-2). Cells between and adjacent to the "peripheral" cells may filled with fuel assemblies that meet the requirements of Figure

.13-2 (i.e., meet the requirements for storage in Region 2-2).

he two prematurely discharged fuel assemblies meet the requiremen of Figure 3. 7.13-4 and qualify for storage in canisters that are loaded I Module H in the southeast corner of the SFP. Module H is in the uppe right corner of the SFP in Figure 3. 7.13-5.

I P3 Fuel Assemblies This LCO establishes restrictions on fuel assembly storage location wit the SFP to ensure that the requirements of 10 CFR 50.68 are met.

This LCO applies whenever any fuel assembly is stored in the Spent Fuel Pit.

INSERT: A Note has been added to the ACTIONS to clarify the application of Completion Time rules. The Condition of this Specification may be entered independently for each fuel assembly stored in a location not meeting the requirements of the LCO. This is acceptable since the Required Action for the Condition provides the appropriate compensatory action for each noncomplying fuel assembly.

INDIAN POINT 2 83.7.13-9 Revision

BASES ACTIONS (continued)

SURVEILLANCE REQUIREMENTS INSERT 4 REFERENCES INDIAN POINT 2 A.1 Spent Fuel Pit Storage 83.7.13 Required Action A.1 is modified by a Note indicating that LCO 3.0.3 does not apply.

When the configuration of fuel assemblies stored in the Spent Fuel Pit is not in accordance with the rules established by LCO 3.7.13, the immediate action is to initiate action to make the necessary fuel assembly movement(s) to bring the configuration into compliance with the rules established by LCO 3.7.13.

If unable to move irradiated fuel assemblies while in MODE 5 or 6, LCO 3.0.3 would not be applicable. If unable to move irradiated fuel assemblies while in MODE 1, 2, 3, or 4, the action is independent of reactor operation.

Therefore, inability to move fuel assemblies is not sufficient reason to require a reactor shutdown.

SR 3.7.13.1

• SR verifies by administrative means that the IP2 fuel assembly has been c

• ied based on initial enrichment, burnup, cooling time and number of In I Fuel Burnable Absorbers (IFBA) rods in the fuel assembly in accorda with Figure 3.7.13-1, Figure 3.7.13-2, Figure 3.7.13-3, or Figure 3.7.13-nd that the fuel assembly meets the requirements for the intended stora ocation defined on Figure 3. 7.13-5.

This SR also verifies by administrative m s that the IP3 fuel assembly meets the requirements for storage in the IP

. This administrative verification must be completed prior to placing any ssembly in the SFP. This SR ensures that this LCO and Specification 4.3.1.

• 11 be met after the fuel assembly is inserted in the SFP.

1 O CFR 50.68, "Criticality Accident Requirements."

2.

Safety v

• on by the Office of Nuclear Reactor Regulation Related to Amen No. 150 to Facility Operating License No.

DPR-26, April 19, 1990.

3.

Generic Letter 96-04, "Boraflex Degradation

• Storage Racks."

B 3.7.13-10 Revision

BASES Spent Fuel Pit Storage B 3.7.13 REFERENCES (continued)

INDIAN POINT 2 Northeast Technology Corporation report NET-173-01, "Criticality Analysis for Soluble Boron and Burnup Credit in the Con Edison Indian Point Unit No. 2 Spent Fuel Storage Racks."

5.

N theast Technology Corporation report NET-173-02, "Indian Poin Unit 2 Spent Fuel Pool (SFP) Boron Dilution Analysis."

6.

Safety luation by the Office of Nuclear Reactor Regulation Related to endment No. 227 to Facility Operating License No.

DPR-26, May 9, 2002.

7.

NETCO Letter to Hansler from E. Lindquist, Northeast Technology Corp. da d 12/19/06, Subject - Reference 4 and 5 extension.

8.

Safety Evaluation by the ice of Nuclear Reactor Regulation Related to Amendment No. 26 to Facility Operating License No.

DPR-26, July 13, 2012.

9.

Holtec Report Hl-2094289, Licensin Report on the Inter-Unit Transfer of Spent Nuclear Fuel at Ind, Point Energy Center, Revision 6.

1 O.

NETCO Letter to Floyd Gumble from tt Harris dated 12/22/2009, titled "Indian Point 2 RACKLIFE Proj tions Through 2010 and 2012 BADGER Tests with RACKLIFE Vers* n 2.1".

11.

NETCO Letter to Giancarlo Delfini from Matt Harn 12/12/2012, titled "Update of IP2 RACKLIFE Model -

(In artial fulfillment of Entergy Contract 10351857, Change Order No. 1, Task 2A)".

B 3.7.13-11 Revision

INSERT 1 The Spent Fuel Pit (SFP) is used to store IP2 spent fuel removed from the reactor and new fuel ready for insertion into the reactor. It is also used to store IP3 fuel that has been transferred from the I P3 SFP prior to it being placed into dry cask storage (Refs 1 and 2). Spent fuel racks (SFRs) are erected on the SFP floor to hold the fuel assemblies.

The SFRs have been evaluated to meet the requirements of option (b) of 1 O CFR 50.68, "Criticality Accident Requirements" (Ref. 3).

IP2 compliance with 10 CFR 50.68(b)(4) was confirmed by an analysis documented in Curtiss-Wright Nuclear Division, NETCO report NET-28091-003-01, Rev. O "Criticality Safety Analysis for the Indian Point Unit 2 Spent Fuel Pool with No Absorber Panel Credit." (Ref. 4) and was approved by the NRG in Amendment TBD (Ref. 5). This analysis demonstrates that 10 CFR 50.68(b)(4) will be met during normal SFP operation and all credible accident scenarios if the following requirements are met:

a)

Spent Fuel Pit boron concentration is maintained within the limits of LCO 3.7.12, "Spent Fuel Pit Boron Concentration," whenever fuel is stored in the SFP; and, b)

Fuel assembly storage location within the Spent Fuel Pit is restricted in accordance with LCO 3.7.13, "Spent Fuel Pit Storage," based on the fuel assembly's initial enrichment, burnup, decay of Plutonium-241 (i.e., cooling time),

averaged assembly peaking factor, and number of Integral Fuel Burnable Absorbers (IFBA) rods. Note that historic fuel assemblies have been pre-categorized (Tables 3.7.13-2 and 3.7.13-3).

Fuel assembly storage location within the Spent Fuel Pit is an essential element of the criticality analysis. The storage racks in the areas designated Region 1 have a different design than the storage racks in the areas designated Region 2 and this design difference has a significant impact on criticality calculations. Regions 1 and 2 are shown on the insert to Figure 3.7.13-1.

Region 1 consists of three racks that use the flux trap design and has 269 cell locations for the storage of fuel assemblies. The flux trap design used in Region 1 uses spacer plates in the axial direction to separate the cells. Boraflex panels are held in place adjacent to each side of the cell by picture-frame sheathing.

In addition, due to a damaged cell, there is one cell blocker in Region 1.

Region 2 consists of nine racks that use an egg-crate design and has 1105 cell locations for the storage of fuel assemblies and two failed fuel canisters. The Region 2 racks consist of boxes welded into a "checkerboard" array with a storage location in each square. One Boraflex absorber panel is held to one side of each cell wall by picture frame sheathing. In addition, there is one cell blocker in Region 2 that is credited in the analysis to allow higher reactive fuel assemblies to be stored in two of its adjacent cells.

This cell blocker cannot be moved while fuel is stored in the Spent Fuel Pit.

With two installed cells blockers the total number of cell locations available for the storage of fuel assemblies is 1372. However, the number of fuel assemblies that can actually be stored in Regions 1 and 2 is dependent on the fuel assembly categorization and the SFP locations allowable by Figure 3.7.13-1 and its associated notes.

The criticality analysis defines five fuel assembly reactivity categories based on minimum required burnup and the number of IFBA rods. The minimum required burnups are chosen to optimize the storage of fresh and once burnt fuel assemblies in Region 1 (Categories 1, 2, and 3), and to optimize the storage of permanently discharged assemblies in Region 2 (Categories 4 and 5).

The fuel categories are numbered from most reactive fuel (Category 1) to least reactive fuel (Category 5). The assembly's enrichment, burnup, cooling time, averaged assembly peaking factor, and number of IFBA rods are used to determine the reactivity category.

The reactivity categories and their associated burnup requirements are shown in order of decreasing reactivity in Table 3.7.13-1. The averaged assembly peaking factor is the assembly burnup divided by the sum of the cycle burnups for the cycles the assembly was in the core.

The depletion analysis treats historic fuel and current and future fuel differently. Historic fuel is defined as fuel assemblies with identifiers (IDs) A through X for IP2 and A through AA for IP3. Current and future fuel is defined as fuel assemblies with IDs after X for IP2 and after AA for IP3. As the fuel designs and operating condition of historic fuel are known this fuel can be pre-categorized. This pre-categorization in included in Tables 3.7.13-2 and 3.7.13-3.

a. Categorization of historic fuel assemblies For historic fuel assemblies, the depletion calculations utilize actual fuel assembly depletion conditions instead of bounding conditions that would penalize most assemblies for which the depletion conditions are known. This data includes use of inserts, burnup achieved while the insert was in the assembly, use of control rods, soluble boron level, averaged assembly peaking factor, and use of axial blankets. The fuel categorization of historic fuel is accomplished by use of batch groupings with similar characteristics. The depletion analysis is performed for each batch grouping using bounding depletion parameters. For some historic fuel assemblies the bounding parameters are overly conservative.

For such assemblies, the actual depletion parameters are used. Historic fuel assemblies have been pre-categorized as shown in Appendix B to the CSA and TS Tables 3.7.13-2 and 3.7.13-3.

b. Categorization of current and future fuel assemblies For current and future fuel assemblies bounding depletion assumptions are used.

Table 3.7.13-1 provides the minimum required burnup for each reactivity category for these fuel assemblies. The notes to Table 3.7.13-1 specify the circumstances under which burnup penalties must be applied to the minimum required burnup.

Should an assembly not meet any of the fuel assembly operating requirements specified in the analysis, then the fuel must be categorized as Category 1 (or Category 4 if a full length RCCA is inserted).

c. SFP storage location by fuel assembly category The analysis determines acceptable storage locations for each fuel assembly reactivity category as shown in Figures 3.7.13-1 and 3.7.13-2.

Figure 3.7.13-1 shows the base case arrangement of the fuel categories. Note that the base case arrangement only shows four reactivity categories since it is the most limiting reactivity arrangement.

d. Storage of non-fuel assemblies The analysis addresses the use of the failed fuel containers, the fuel rod storage basket, assemblies with missing fuel rods, and the storage of miscellaneous materials.

Failed fuel containers The southeast corner of the spent fuel pool contains two 16" circular pipes that are used as failed fuel containers. The criticality analysis permits 16 fuel rods in each of these failed fuel containers. The maximum of 16 fuel rods in each of the failed fuel containers will be controlled by procedure.

Fuel rod storage basket There is a movable fuel rod storage basket that can be used to store fuel rods.

This basket can fit in a storage cell and has 52 holes for storing fuel rods. The fuel rod storage basket is classified as reactivity Category 4 and this categorization is included in Table 3. 7.13-2.

Assemblies with missing fuel rods Reconstituted fuel assemblies may be stored in the SFP provided a 4GWdff burnup penalty is added to the MRS for those assemblies that did not have stainless steel rods installed. This requirement is included in the notes to Table 3.7.13-1.

Storage of miscellaneous materials Water Holes and 50% Water Holes may be used to store miscellaneous materials with certain restrictions. Miscellaneous non-actinide materials, for example, empty or full trash baskets, can be stored in fuel positions of any category. However, there are some special cases where some of the material may be stored in a water hole or 50% water hole. If the miscellaneous material is any type of steel, lnconel, or absorber material (e.g., absorber coupons, stainless steel coupon trees, control rods, unburned burnable absorbers) it may displace up to 50% of the water volume at the active fuel zone (144 inches) of a water hole or 50% water hole (there are no restrictions on material above or below the active fuel zone). If the miscellaneous material is a very low absorbing material such as a void, zirconium, aluminum, cloth, plastic, concrete, etc., it cannot be placed in a water hole but may be placed in a 50% water hole so long as the 50% water hole still has 50% water volume in the active fuel zone. The restrictions that apply to Water Holes and 50% Water Holes are included in the notes to Figure 3.7.13-1.

INSERT 2 Curtiss-Wright Nuclear Division, NETCO report NET-28091-003-01, Rev. O "Criticality Safety Analysis for the Indian Point Unit 2 Spent Fuel Pool with No Absorber Panel Credit" (Ref. 4) evaluated non-accident and accident conditions in the SFP.

Reference 4 determined that if storage location requirements in this LCO are met then then the SFP will have a keff of s 0.95 if filled with a soluble boron concentration of.::

700 ppm and will have a keff of < 1.0 if filled with unborated water.

Reference 4 evaluated abnormal occurrences and accidents.

This evaluation considered the effects of the following: a) multiple misleads, b) an assembly placed alongside a rack; c) a dropped assembly, d) a misloaded assembly; e) SFP over temperature, f) a seismic event, and, g) a SFP boron dilution accident. Reference 4 determined that the most limiting fuel handling accident is the multiple mislead accident and determined keff of :::; 0.95 if the SFP is filled with a soluble boron concentration of

~ 2000 ppm.

NET-173-02, "Indian Point Unit 2 Spent Fuel Pool (SFP) Boron Dilution Analysis" (Ref.

6) evaluated postulated unplanned SFP boron dilution scenarios assuming an initial SFP boron concentration within the limits of LCO 3.7.12. The evaluation considered various scenarios by which the SFP boron concentration may be diluted and the time available before the minimum boron concentration necessary to ensure subcriticality for the non-accident condition (i.e. it is not assumed an assembly is misloaded concurrent with the Spent Fuel Pit dilution event).

Reference 6 determined that an unplanned or inadvertent event that could dilute the SFP boron concentration from 2000 ppm to 786 ppm is not a credible event because of the low frequency of postulated initiating events and because the event would be readily detected and mitigated by plant personnel through alarms, flooding, and operator rounds through the SFP area. The criticality analysis conservatively credits 700 ppm (versus an already not credible 786 ppm in the AOR) of soluble boron.

The configuration of fuel assemblies in the Spent Fuel Pit satisfies Criterion 2 of 10 CFR 50.36(c)(2)(ii).

INSERT 3 This LCO requires that each IP2 fuel assembly stored in the Spent Fuel Pit is categorized in accordance with Table 3.7.13-1 or, if pre-categorized, Table 3.7.13-2, that each IP3 fuel assembly stored in the Spent Fuel Pit is categorized in accordance with Table 3.7.13-1 or, if pre-categorized, Table 3.7.13-3, and, IP2 and IP3 fuel assembly storage locations within the Spent Fuel Pit shall be restricted to locations allowed by Figure 3.7.13-1 and its associated notes as follows:

a. Categorized fuel assemblies may be stored in any cell location of the same or lower numbered category.

b. Category 1 fuel assemblies may be stored in any cell location in Regions 1 and 2 in accordance with Figure 3.7.13-1 notes 12 and 13, respectively.

c. Category 5 fuel assemblies with an installed full length RCCA take the reactivity credit provided by the presence of the RCCA. This credit may also be taken for Category 1 fuel assemblies that contain a full length RCCA. These assemblies may be stored in any Category 4, 3, 2, or 1 cell. Likewise, Category 2, 3, or 4 fuel assemblies that contain a full length RCCA may be stored any Category 5, 4, 3, 2, or 1 cell.

d. Category 5 fuel assemblies that are required to have a full length RCCA installed are the subject of the LCO note. Because reactivity credit is taken for the installed RCCA it may not be placed in, or removed from, the fuel assembly while the assembly is in the RCCA credited location. Movement of the RCCA in the credited location would be a violation of the criticality analysis.

This LCO establishes restrictions on fuel assembly storage location within the SFP to ensure that the requirements of 1 O CFR 50.68 are met.

INSERT 4 This SR verifies by administrative means that the IP2 fuel assembly has been classified based on initial enrichment, burnup, cooling time, averaged assembly peaking factor, and number of IFBA rods in the fuel assembly in accordance with Figure 3.7.13-1, Figure 3.7.13-2, Figure 3.7.13-3, or Figure 3.7.13-4 and that the fuel assembly meets the requirements for the intended storage location defined on Figure 3.7.13-5. This SR also verifies by administrative means that the IP3 fuel assembly meets the requirements for storage in the IP2 SFP. This administrative verification must be completed prior to placing any fuel assembly in the SFP. This SR ensures that this LCO and Specification 4.3.1.1 will be met after the fuel assembly is inserted in the SFP.

INSERT 5

1.

Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. 268 to Facility Operating License No. DPR-26, July 13, 2012.

2.

Holtec Report Hl-2094289, Licensing Report on the Inter-Unit Transfer of Spent Nuclear Fuel at Indian Point Energy Center, Revision 6.

3.

1 O CFR 50.68, "Criticality Accident Requirements."

4.

Curtiss-Wright Nuclear Division, NETCO report NET-28091-0003-01, Rev. O "Criticality Safety Analysis for the Indian Point Unit 2 Spent Fuel Pool with No Absorber Panel Credit."

5.

Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment No. TBD to Facility Operating License No. DPR-26, TBD.

6.

Northeast Technology Corporation report NET-173-02, "Indian Point Unit 2 Spent Fuel Pool (SFP) Boron Dilution Analysis."

Enclosure Attachment 3 NL-19-023 Retyped Technical Specification Pages

Spent Fuel Pit Storage 3.7.13 3.7 PLANT SYSTEMS 3.7.13 Spent Fuel Pit Storage LCO 3.7.13 APPLICABILITY:

INDIAN POINT 2 IP2 fuel assemblies stored in the Spent Fuel Pit shall be categorized in accordance with Table 3.7.13-1 or, if pre-categorized, Table 3.7.13-2.

IP3 fuel assemblies stored in the Spent Fuel Pit shall be categorized in accordance with Table 3.7.13-1 or, if pre-categorized, Table 3.7.13-3.

IP2 and IP3 fuel assembly storage locations within the Spent Fuel Pit shall be restricted to locations allowed by Figure 3. 7.13-1 and its associated notes.

---

Note-----------------------------------------------

Regarding Category 5 fuel assemblies that are required by Figure 3. 7.13-1 to contain a full length RCCA - The RCCA must not be placed in or removed while the assembly is in an RCCA required location unless all 8 adjacent cells are empty.

Whenever any fuel assembly is stored in the Spent Fuel Pit.

3.7.13-1 Amendment No.

ACTIONS

- NOTE-Spent Fuel Pit Storage 3.7.13 Separate Condition entry is allowed for each fuel assembly.

CONDITION A.

Requirements of the LCO not met.

A. 1 REQUIRED ACTION COMPLETION TIME

- NOTE-LCO 3.0.3 is not applicable.-----------------------

Initiate action to move the Immediately noncomplying fuel assembly to an acceptable location.

SURVEILLANCE REQUIREMENTS SR 3.7.13.1 INDIAN POINT 2 SURVEILLANCE Verify by administrative means that the IP2 fuel assembly has been categorized in accordance with Table 3.7.13-1 or, if pre-categorized, Table 3.7.13-2 and meets the requirements for the intended storage location.

OR Verify by administrative means that the IP3 fuel assembly has been categorized in accordance with Table 3.7.13-1 or, if pre-categorized, Table 3.7.13-3 and meets the requirements for the intended storage location.

3.7.13-2 FREQUENCY Prior to storing the fuel assembly in the Spent Fuel Pit.

Prior to storing the fuel assembly in the Spent Fuel Pit.

Amendment No.

Spent Fuel Pit Storage 3.7.13 1

2 3

4 5

6 7

8 9 W U U U U li tt U U tt W U ll il M B U U ll B

~ ll H

1--+--+--+--+--+--+--+--+--+--i G

1--+--+--+--+--+--+--+--+--+--i F

1--+--+--+--+--+--+--+--+--+--i E

l--l--l--l--l--l--l--l--1--1--1 D

l--l--l--l--l--l--l--1--1--1--1 C

l--l--l--l--l--l--l--l--1--1--1 B

Key:

D WaterHole D 50% Water Hole category lfuel D category 2 Fuel D category3Fuel category 4 Fuel category 5 Fuel 11 category 5 Fuel with a required full lengthRCCA

[!] Blocked cell I Cask Area Figure 3.7.13-1 (page 1 of 2)

Allowable Spent Fuel Pit Storage Locations for Category 1 through Category 5 Fuel Assemblies in Regions 1 and 2 INDIAN POINT 2 3.7.13 - 3 Amendment No.

DP DN DM DL DK DJ DH DG DF DE CP CN CM CL CK a

CH CG CF CE CD 8N BM BL BK Bl BH BG Bf BE BD BC

-Notes-Spent Fuel Pit Storage 3.7.13

1.

Fuel assembly Categories are ranked in order of relative reactivity, from Category 1 to 5. Category 1 fuel assemblies have the highest reactivity, and Category 5 fuel assemblies have the lowest.

2.

Fuel assembly categorization for assembly IDs after X for IP2 and after AA for IP3 must be performed in accordance with Table 3.7.13-1.

3.

Fuel assembly Categories for IP2 assembly IDs A through X are located in Table 3.7.13-2.

4.

Fuel assembly Categories for IP3 assembly IDs A through AA are located in Table 3.7.13-3.

5.

Fuel assemblies of any higher numbered Category can be stored in any cell location that allows for a lower numbered Category. For example, a Category 5 fuel assembly can be stored in Category 1, 2, 3, 4, and 5 cells. Any cell may be empty.

6.

Category 1 fuel assemblies that contain a full length RCCA may be stored in any Category 4, 3, 2, or 1 cell.

7.

Category 2, 3 or 4 fuel assemblies that contain a full length RCCA may be stored in any Category 5 cell that does not require an inserted RCCA or in any Category 4, 3, 2, or 1 cell.

8.

A Water Hole may contain up to 50% of absorber material by volume in the active fuel area. Stainless steel and lnconel meet the definition of absorber material. There is no restriction for non-actinide material outside of the active fuel area.

9.

A 50% Water Hole may contain up to 50% of any non-actinide material by volume in the active fuel area.

Zirconium meets the definition of non-actinide material. There is no restriction for non-actinide material outside of the active fuel area.

10.

A Blocked Cell has the same requirements as a Water Hole.

11.

A checkerboard area consists of every other cell being a Water Hole.

12.

An area of Category 1 fuel assemblies may be formed in Region 1. The Category 1 area must be formed by replacing the Region 1 arrangement shown in this figure with an area of Category 1 fuel assemblies in accordance with the following criteria (see examples in Figure 3.7.13-2):

a)

Category 1 fuel assemblies must be face adjacent to at least three Water Holes and not face adjacent to adjacent to another Category 1 assembly.

b)

Category 2 fuel assemblies must not be face adjacent to more than one Category 1 fuel assembly.

c)

Catergory 3 and Category 5 locations in Figure 3.7.13-1 may not be moved.

13.

A checkerboard area of Category 1 fuel assemblies may be formed in Region 2. All four sides of the checkerboard area must be rows of Water Holes.

14.

The edge of Region 2 next to the pool wall or cask loading area can be considered to be a row of Water Holes.

Figure 3.7.13-1 (page 2 of 2)

INDIAN POINT 2 3.7.13-4 Amendment No.

H G

F E

D C

B H

G F

E D

C B

J H

G F

E Spent Fuel Pit Storage 3.7.13 1

2 3

4 s

s 1

s g

w u u u u H

u v a

~ w n n n

M B

~ v u

~ ~ n t---+---+--

t---+---+---+----+-t---+---+---+----+---1 t---+---+--+----+-t---+--+--1----+---1 1

2 3

4 5

6 7

8 9

W ll U ll M ll U V

U

~ W ll ll n M

B

~ V U

~ ~ ll 1

2 3

4 S

6 7

8 9

W ll U ll M U

U V

U

~ W ll ll n M

B

~ V U

~ ~ ll t---+--+--+----+-

t---+--+--1----+---1

>---+----+--

t---+--+--

D C

B t---+--+--+----+-t---+--+--1---+--l

>---+----+---+----+-+---+----+---+--->----<

Figure 3.7.13-2 Examples of Allowable Spent Fuel Pit Storage Locations for Category 1 Fuel Assemblies in Region 1 INDIAN POINT 2 3.7.13-5 Amendment No.

Table 3.7.13-1 Spent Fuel Pit Storage 3.7.13 Fuel Assembly Reactivity Categorization for Assembly IDs after X for IP2 and after AA for IP3 Reactivity Minimum Required 8urnup (MR8) (GWd/T)(a)(b)(cl Category 1

o(d) 2 21 3

28.5 4

812 =(al+ a2*E + a3*E2) x exp[-(a4 + aS*E + a6*E2) x CT]+ a7 + a8*E + a9*E2 80_8 =(bl+ b2*E + b3*E 2) x exp[-(b4 + bS*E + b6*E2) x CT]+ b7 + b8*E + b9*E 2 MR8 = 80.s + (812 - 80.s) x (PF - 0.8)/ 0.4 5

MR8 for Category 4 plus 11 Where:

E is enrichment in wt% U-23S(el, CT is cooling time in years11l, and PF is the average peaking factor defined by the fuel assembly burnup divided by the sum of the cycle burnups for the cycles the fuel assembly was in the core.

and:

Coefficient Value Coefficient Value al

-6.26824 bl 15.1405 a2 5.29367 b2

-4.81133 a3

-0.37154 b3 0.753855 a4 0.129582 b4 0.121252 as

-0.0204918 bS

-0.0150991 a6 0.00205596 b6 0.00127009 a7

-0.13331 b7

-16.2293 a8 6.9037 b8 14.0159 a9 0.122068 b9

-0.687054 (a) 2 GWd/T must be added to the MR8 for any fuel assembly that had a Hafnium insert.

(b) 4 GWd/T must be added to the MR8 for any fuel assembly that was reconstituted without replacing removed fuel rods with stainless steel rods.

(c) 0.2, 0.3, 0.6, and 0.9 GWd/T must be added to the MR8 for Categories 2, 3, 4, and 5, respectively, if the multi-cycle burn up averaged soluble boron concentration of 950 ppm is exceeded.

(d) With 64 IF8A rods or more. Assemblies with enrichments less than or equal to 4.5, 4.0, 3.5, and 3.0 require only 48, 32, 16, and O IF8A rods, respectively.

INDIAN POINT 2 3.7.13-6 Amendment No.

Spent Fuel Pit Storage 3.7.13 (e) Fuel assemblies at enrichments less than 4.2 wt% U-235 must use 4.2 wt% U-235 in the Category 4 equation.

(f) Fuel assemblies with cooling times of more than 25 years must use 25 years in the Category 4 equation.

INDIAN POINT 2 3.7.13-7 Amendment No.

Table 3. 7.13-2 (page 1 of 3)

Spent Fuel Pit Storage 3.7.13 Fuel Assembly Reactivity Categorization for Assembly IDs A through X for IP2 Indian Point Unit 2 Fuel Assembly ID Category Assembly ID Category Assembly ID Category A01-A65 4

E43-ESS 4

K01-K13 4

ESG 3

K14-K1S 5

B01-B07 4

E57-E60 4

K16-K57 4

B08-B13 s

KS8 s

Bl4-B23 4

FOl 3

KS9-K68 4

B24-826 s

F02-F20 4

B27-B64 4

F21 3

L01-L07 4

F22-F30 4

L08-L10 s

C01-C04 4

F31-F34 s

lll-L63 4

COS-COG s

F3S 4

L64 3

C07-C12 4

F36 3

L65-L68 4

Cl3 s

F37-F39 4

C14 4

F40 3

M01-M04 4

C1S-C18 s

F41-F49 4

MOS s

C19-C28 4

FSO 3

M06-M08 4

C29 s

F51-F60 4

M09 5

C30-C64 4

F61 3

Ml0-Ml2 4

F62-F64 4

M13-M14 s

001-025 4

FGS 3

M15-M20 4

026 s

F66 4

M21 s

027-060 4

F67-F68 s

M22-M23 4

061-068 s

M24 s

069-072 4

GOl-GOS 4

M25-M27 4

GOG 5

M28 s

E01-E14 4

G07-G37 4

M29-M30 4

ElS 3

G38 s

M31 s

E16-E19 s

G39-G72 4

M32-M34 4

E20 4

M3S s

E21-E24 s

H01-H38 4

M36-M37 4

E25-E27 4

H39-HS1 5

M38-M44 5

E28-E31 s

H52-H54 4

M4S 3

E32-E33 4

HSS s

M46 4

E34-E3S s

HSG 4

M47-M48 s

E36-E40 4

M49-M50 4

E41-E42 s

J01-J68 4

M51-MS2 s

INDIAN POINT 2 3.7.13-8 Amendment No.

Table 3. 7.13-2 (page 2 of 3)

Spent Fuel Pit Storage 3.7.13 Fuel Assembly Reactivity Categorization for Assembly IDs A through X for IP2 Indian Point Unit 2 Fuel Assembly ID Category Assembly ID Category Assembly ID Category MS3-MS4 4

Q71-Q73 4

T42-T43 4

M55-M56 5

Q74-Q76 5

T44-T46 5

MS7 4

Q77 4

T47 4

M58-M59 5

Q78 5

T48 5

M60 4

Q79-Q80 4

T49-TS1 4

M61 3

T52-TS3 5

M62-M63 4

R01-R07 5

TS4 4

M64 3

ROS 4

TSS 5

M65 4

R09-R38 5

TS6-T72 4

M66 5

R39 4

T73-T80 5

M67 3

R40-R43 5

M68 5

R44-RSO 4

M69-M71 4

R51-R69 5

U01-U04 5

M72 5

R70 4

uos 4

R71-R72 5

U06-U13 5

N01-N08 4

R73-R74 4

U14 4

N09-N12 5

R75-R79 5

U15-U16 5

N13-N14 4

R80-R81 4

U17-U21 4

N15-N16 5

R82 5

U22 5

N17-N23 4

R83-R85 4

U23 4

N24-N32 5

U24-U49 5

N33-N47 4

S01-S44 5

USO 4

N48 5

S45 4

USl 5

N49-N80 4

S46-S47 5

US2 4

S48 4

U53-U61 5

P01-P02 4

S49-S61 5

U62-U64 4

P03 3

S62 4

U65 5

P04-P47 4

S63-S65 5

U66-U68 4

P48 5

S66 4

U69-U73 5

P49-P60 4

S67-S77 5

P61-P72 5

V01-V16 5

V17-V29 4

Q01-Q65 5

T01-T32 5

V30-V35 5

Q66 4

T33-T34 4

V36 4

Q67-Q68 5

T35-T36 5

V37-V38 5

Q69 4

T37 3

V39 4

Q70 5

T38-T41 5

V40-V41 5

INDIAN POINT 2 3.7.13-9 Amendment No.

Table 3. 7.13-2 (page 3 of 3)

Spent Fuel Pit Storage 3.7.13 Fuel Assembly Reactivity Categorization for Assembly IDs A through X for IP2 Indian Point Unit 2 Fuel Assembly ID Category Assembly ID Category Assembly ID Category V42-V43 4

W21 5

X01-X02 3

V44-V49 5

W22 4

X03-X04 5

vso 4

W23 5

XOS-X37 4

V51-V54 5

W24 4

X38 5

V55-V57 4

W25 5

X39-X49 4

V58-V61 5

W26 4

XSO-XSl 5

V62 4

W27 5

X52-X53 4

V63 5

W28-W34 4

X54-XSS 5

V64-V65 4

W35 5

X56-X58 4

V66-V67 5

W36-W38 4

X59-X60 5

V68 4

W39 5

X61-X62 4

V69-V77 5

W40 4

X63 5

V78-V79 4

W41-W43 5

X64-X65 4

V80-V81 5

W44-W45 4

X66 5

V82 4

W46 5

X67 4

V83 5

W47 4

X68-X69 5

V84 4

W48-W49 5

X70-X73 4

V85 5

wso 4

X74 5

V86 4

WSl 5

X75 4

V87-V88 5

W52-WSS 4

X76 5

V89 4

W56-W58 5

X77 4

V90-V91 5

WS9-W60 4

X78 5

V92 4

W61 5

X79 4

W62 4

X80-X93 5

WOl-WlO 4

W63-W67 5

X94-X95 4

Wll 5

W68 4

X96 5

W12-W15 4

W69-W71 5

W16 5

W72 4

FRSB1 4

W17 4

W73-W83 5

W18-W19 5

W84 4

W20 4

W85-W93 5

1 FRSB is the Fuel Rod Storage Basket INDIAN POINT 2 3.7.13-10 Amendment No.

Table 3.7.13-3 Spent Fuel Pit Storage 3.7.13 Fuel Assembly Reactivity Categorization for Fuel Assembly IDs A through AA for IP3 Indian Point Unit 3 Fuel Assembly ID I Category I Assembly ID I Category I Assembly ID I Category V43 I

3 I

V48 I

3 I

I All other Fuel Assembly IDs A through AA are Category 4 INDIAN POINT 2 3.7.13-11 Amendment No.

4.0 DESIGN FEATURES 4.1 Site Location Design Features 4.0 Indian Point 2 is located on the East bank of the Hudson River at Indian Point, Village of Buchanan, in upper Westchester County, New York. The site is approximately 24 miles north of the New York City boundary line. The nearest city is Peekskill which is 2.5 miles northeast of Indian Point.

The minimum distance from the reactor center line to the boundary of the site exclusion area and the outer boundary of the low population zone, as defined in 1 O CFR 100.3, is 520 meters and 1100 meters, respectively. For the purpose of satisfying 1 O CFR Part 20, the "Restricted Area" is the same as the "Exclusion Area" shown in UFSAR, Figure 2.2-2.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 193 fuel assemblies. Each assembly shall consist of a matrix of Zircalloy-4 or ZIRLO fuel rods. Fuel shall have a U-235 enrichment of

~ 5.0 weight percent.

Limited substitutions of Zircalloy-4, ZIRLO or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used.

Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRG staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases.

A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions.

4.2.2 Control Rod Assemblies The reactor core shall contain 53 control rod assemblies.

The control rod material shall be silver indium cadmium, clad with stainless steel, as approved by the NRG.

4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:

INDIAN POINT 2 4.0 - 1 Amendment No.

Design Features 4.0 4.0 DESIGN FEATURES 4.3 Fuel Storage (continued) 4.3.2

a.

Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent, and poisons, if necessary, to meet the limit for

kett,

b.

kett ~ 0.95 when flooded with borated water, kett < 1.0 if fully flooded with unborated water, and

c.

Each fuel assembly categorized based on initial enrichment, burnup, cooling time, averaged assembly peaking factor, and number of Integral Fuel Burnable Absorbers (IFBA) rods with individual fuel assembly storage location within the spent fuel storage rack restricted as required by Technical Specification 3.7.13.

4.3.1.2 The new fuel storage racks are designed and shall be maintained with:

Drainage

a.

Fuel assemblies having a maximum U-235 enrichment of 5.0 weight percent, and poisons, if necessary, to meet the limit for

keff,

b.

kett :s; 0.95 if fully flooded with unborated water, and

c.

A 20.5 inch center to center distance between fuel assemblies placed in the storage racks to meet the limit for kett*

The spent fuel pit is designed and shall be maintained to prevent inadvertent draining of the pit below a nominal elevation of 88 feet, 6 inches.

4.3.3 Capacity The spent fuel pit is designed and shall be maintained with a storage capacity limited to no more than 269 fuel assemblies in Region I and 1105 fuel assemblies in Region II.

INDIAN POINT 2 4.0 - 2 Amendment No.