NL-17-102, Response to Request for Additional Information Regarding Inter-Unit Transfer of Spent Fuel

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Response to Request for Additional Information Regarding Inter-Unit Transfer of Spent Fuel
ML17234A402
Person / Time
Site: Indian Point  Entergy icon.png
Issue date: 08/16/2017
From: Vitale A
Entergy Nuclear Northeast
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
CAC MF8991, CAC MF8992, NL-17-102
Download: ML17234A402 (24)
v • d • e


Text

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  • Entergy;

.-ap Entergy Nuclear Northeast Indian Point Energy Center 450 Broadway, GSB P.O. Box 249 Buchanan , NY 10511-0249 Tel (914) 254 6700 Anthony J Vitale Site Vice President NL-17-102 August 16, 2017 U.S. Nuclear Regulatory Commission Document Control Desk 11545 Rockville Pike, TWFN-2 F1 Rockville , MD 20852-2738

SUBJECT:

Indian Point Nuclear Generating Unit Nos. 2 and 3 - Response to Request for Additional Information Regarding Inter-Unit Transfer of Spent Fuel (CAC Nos. MF8991 and MF8992)

Docket Nos. 50-24 7 and 50-286 License Nos. DPR-26 and DPR-64

REFERENCES:

1) Entergy letter dated April19 , 2017, "Response to Request for Supplemental Information Needed for Acceptance of Requested Licensing Action Regarding Amendment of Inter-Unit Transfer of Spent Fuel ," (NL-17-044) (ML17114A467)
2) NRC letter dated April 11 , 2017, "Indian Point Nuclear Generating Unit Nos. 2 and 3 - Supplemental Information Needed for Acceptance of Requested Licensing Action RE : Amendment of Inter-Unit Transfer of Spent Fuel," (CAC Nos. MF8991 and MF8992) (ML 171OOA128)
3) NRC letter dated July 17, 2017 , "Request for Additional Information Regarding Inter-Unit Transfer of Spent Fuel ," (CAC Nos. MF8991 and MF8992)
4) Entergy letter dated December 14, 2016, "Indian Point Nuclear Power

' Plant Units 2 and 3 Proposed License Amendment Regarding the Inter-Unit Transfer of Spent Fuel ," (NL-16-118) (ML16355A067)

Dear Sir or Madam :

By Reference 1, Entergy Nuclear Operations, Inc. (Entergy) provided responses to the supplemental information request by the U.S. Nuclear Regulatory Commission (NRC), identified in Reference 2 pertaining to the review of the License Amendment Request for Indian Point Energy Center (IPEC) Unit Nos. 2 and 3 Inter-Unit Transfer of Spent Fuel.

By Reference 3, the NRC identified a need for additional information in order to complete its review of the proposed license amendment requested by Reference 4. The attachments to th is letter provide the information requested by Reference 3. The information provided herein does not change the conclusion that the proposed changes involve no significant hazards considerations .

NL-17-102 Docket Nos. 50-24 7 and 50-286 Page 2 of 2 , Holtec International Report Hl-2094289 Appendix 4.A, contains information that Holtec International considers to be proprietary and , therefore, exempt from public disclosure pursuant to 10 CFR 2.390. In accordance with 10 CFR 2.390 and in support of this request for withholding , an affidavit executed by Holtec International is provided in Attachment 5.

There are no new commitments being made in this submittal.

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

I declare under penalty of perjury that the foregoing is true and correct. Executed on A.% \Cz , 2017.

Sincerely, AJV/rl Attachments :

1. Reply to NRC Request for Additional Information Regarding Inter-Unit Transfer of Spent Fuel
2. Revised Proposed TS Page Markups
3. Revised Holtec Report Hl-2094289 Appendix 4.A (Proprietary Version)
4. Revised Holtec Report Hl-2094289 Appendix 4.A (Non-Proprietary Version)
5. Affidavit in Support of Request to Withhold Information cc: Mr. Daniel H. Dorman , Regional Administrator, NRC Region I Mr. Sherwin E. Turk, NRC Office of General Counsel , Special Counsel Mr. William Burton, NRC Senior Project Manager, Division of License Renewal Mr. Richard V. Guzman, NRR Senior Project Manager Ms. Bridget Frymire, New York State Department of Public Service Ms. Alicia Barton , President and CEO NYSERDA NRC Resident Inspector's Office

ATTACHMENT 1 to NL-17-102 REPLY TO NRC REQUEST FOR ADDITIONAL INFORMATION REGARDING THE LICENSE AMENDMENT REQUEST FOR INTER-UNIT TRANSFER OF SPENT FUEL ENTERGY NUCLEAR OPERATIONS , INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS. 50-247 AND 50-286

NL-17-102 Attachment 1 Page 1 of 7 Request for Additional Information (RAl)-1 (RCCB)

In the supplement dated April 19, 2017, the licensee stated that the preliminary results from the 2017 BADGER campaign demonstrate that the Boraflex degradation in Region 1-2 is still bounded by the current IP2 SFP criticality analysis of record (CAOR). In order to verify that the Boraflex degradation is bounded , the NRC staff requests that the licensee provide:

a. A comparison of the maximum local degradation recorded during the 2017 BADGER campaign to the IP2 SFP CAOR.
b. A comparison of the average panel degradation recorded during the 2017 BADGER campaign to the IP2 SFP CAOR.
c. A comparison of the maximum and average gap size recorded during the 2017 BADGER campaign to the IP2 SFP CAOR.
d. Any additional relevant information that demonstrates the Boraflex degradation is bounded by the current I P2 SFP CAOR.

NRC Request

a. Provide a comparison of the maximum local degradation recorded during the 2017 BADGER campaign to the IP2 SFP CAOR.

Response

Table 1 below, contains the panel average and maximum areal densities for the sixteen (16)

Region 1-2 panels tested during the 2017 BADGER campaign at Indian Point Unit 2 (IP2).

Column 4 contains the minimum measured areal density for each panel. Column 5 contains the peak change (%) in areal density relative to the minimum certified areal density of 0.028 g boron-10/cm 2 for Region 1-2 from Reference [1] § 4.5.2.1 1 Positive values in the tables indicate the areal densities were higher than the minimum certified areal density, whereas negative values indicate the measured areal densities were below the minimum certified areal density.

The peak changes in areal density for the Region 1-2 panels in the 2017 IP2 BADGER test ranged from a low of -4.7 percent for panel D21 South to a maximum decrease of-43.7 percent for panel C19 North.

1 Reference [1]: "Indian Point Unit 2 Spent Fuel Pool Increased Storage Capacity Licensing Report",

Consolidated Edison Company of New York; June 1988

NL-17-102 Attachment 1 Page 2 of 7 Table 1: BADGER Measured Peak and Average Areal Density Change for the 2017 IP2 Region 1-2 Test Panels Panel Intact Panel Intact Peak%

Average % Change BADGER Estimated Average Minimum Change in in Minimum Change in Total Boron Panel ID Measured Areal Measured Areal Minimum Certified Areal Carbide(%) (from Density Density Certified Density Minimum Areal Density)

(g ioB/cm2) (g ioB/cm2) Areal Density C19N 0.0268 -4.4% 0.0158 -43.7% -14.5 E18N 0.0256 -8.5% 0.0182 -34.9% -16.9 H18N 0.0310 10.7% 0.0215 -23.2% -0.9 H18S 0.0305 8.9% 0.0200 -28.6% -4.5 F19S 0.0327 16.7% 0.0174 -37.8% -4.9 D21S 0.0343 22.6% 0.0267 -4.7% -0.3 F19N 0.0324 15.9% 0.0189 -32.4% -2.0 D21N 0.0337 20.4% 0.0207 -26.1% -1.2 D21W 0.0328 17.2% 0.0201 -28.2% -2.5 E18W 0.0301 7.4% 0.0202 -27.7% -6.9 H18E 0.0338 20.6% 0.0189 -32.6% -3.8 H18W 0.0330 18.0% 0.0201 -28.1% -5.8 C19W 0.0333 18.8% 0.0176 -37.2% -5.8 F19W 0.0338 20.8% 0.0195 -30.5% -3. 6 F19E 0.0344 23.0% 0.0160 -43.0% -3.0 D21E 0.0321 14.6% 0.0175 -37.5% -4.9 Panel Intact Panel Intact Peak% loss Average % loss BADGER Estimated Total Average Minimum from from Minimum Boron Carbide Loss (%)

Panel ID Measured Areal Measured Areal Minimum Certified Areal (from Minimum Areal Density Density (gm b- Certified Areal Density Density)

(gm b-10/cm2) 10/cm2) Density C19N 0.0268 -4.4% 0.0158 -43.7% -14.5 E18N 0.0256 -8.5% 0.0182 -34.9% -16.9 HlBN 0.0310 10.7% 0.0215 -23.2% -0.9 H18S 0.0305 8.9% 0.0200 -28.6% -4.5 F19S 0.0327 16.7% 0.0174 -37.8% -4.9 D21S 0.0343 22.6% 0.0267 -4.7% -0.3 F19N 0.0324 15.9% 0.0189 -32.4% -2.0 D21N 0.0337 20.4% 0.0207 -26.1% -1.2 D21W 0.0328 17.2% 0.0201 -28.2% -2.5 E18W 0.0301 7.4% 0.0202 -27.7% -6.9 H18E 0.0338 20.6% 0.0189 -32.6% -3.8 H18W 0.0330 18.0% 0.0201 -28.1% -5.8 C19W 0.0333 18.8% 0.0176 -37.2% -5.8 F19W 0.0338 20.8% 0.0195 -30.5% -3.6 F19E 0.0344 23.0% 0.0160 -43.0% -3.0 D21E 0.0321 14.6% 0.0175 -37.5% -4.9

~ ------------------------------------------------

NL-17-102 Attachment 1 Page 3 of 7 Table 2, below, contains the assumed average and peak relative change(%) in areal density for the 9 Region 1-2 panel models assumed in the I P2 CAOR taken from Appendix A of Reference [2]2. Similar to Table 1, the changes in Table 2 are relative to the minimum certified areal density for Region 1-2, since this is consistent with the assumptions in the CAOR. The average change in areal density is modeled as a uniform reduction in areal density, whereas the peak change in areal density is modeled as a local reduction in areal density. With the exception of one panel (A14E) in Table 2, the peak losses assumed in the CAOR ranged from -50 to -95 percent. These were well above the peak losses estimated by BADGER in the 2017 test.

Table 2: Assumed Peak and Uniform Relative Region 1-2 Panel Areal Density Losses in the IP2 CAOR.

Assumed Average Assumed Peak Change (Uniform CAORPanel Change in Minimum Thinning) in Minimum ID Certified Areal Certified Areal Density Density(%)

(%}

A14E - 17.9 0 A14N -18.4 -85 A14S -24 -95 B14N -20.8 -50 B14S - 20.8 -75 B14W -18.5 - 65.0 B15W - 23.9 -80 B17W - 19.2 -70 D9N -23.8 -80 NRC Request

b. Provide a comparison of the average panel degradation recorded during the 2017 BADGER campaign to the IP2 SFP CAOR.

Response

Referring again to Tables 1 in (a) above, only two (2) of the sixteen (16) panels tested , had average areal densities below the minimum certified areal density for Region 1-2. The average measured change in areal density for all of the Region 1-2 panels was 13.9 percent relative to the minimum certified areal density. The average change in measured areal density in the 2017 IP2 BADGER test ranged from a low of-8.5 percent (a decrease) for panel E18 North to a maximum of +23.0 percent for panel F19 East. The two panels (C19 North and E18 North) that exhibited a decrease in average areal density had relative changes of -4.4 and -

8.5 percent respectively.

2 Reference [2]: "Characterization of Boraflex Panels in the Indian Point Unit No. 2 Spent Fuel Pool Projected to the End of 2006", NET-170-02. Northeast Technology Corp.: Kingston , NY; 1 February 200 1

NL-17-102 Attachment 1 Page 4 of 7 Referring again to Table 2 in (a), Column 2 of Table 2 lists the assumed uniform thinning loss for the 9 panel models for the I P2 CAOR. The uniform loss for C 19 North and E 18 North, which are the only panels that had average areal densities below the minimum certified areal density, was less than the average losses of each of the panels modeled in the CAOR.

NRC Request

c. Provide a comparison of the maximum and average gap size recorded during the 2017 BADGER campaign to the IP2 SFP CAOR.

Response

For the 16 panels tested in the 2017 IP2 BADGER test, the maximum individual gap size measured was estimated to be 3 1/8 inches high whereas the maximum gap size assumed in the CAOR was 4 1/3 inches high. The average estimated gap size for all gaps measured in the 2017 IP2 BADGER test was 0.8 inches whereas the average gap size assumed in the 9 Region 1-2 panel models in the CAOR is approximately 1.3 inches.

Gaps less than 1/3 of an inch are indistinguishable from local anomalies when using the BADGER system and a bias is included in the CAOR to account for small anomalies that could be cracks but appear as local thin areas on the Boraflex panels.

NRC Request

d. Any additional relevant information that demonstrates the Boraflex degradation is bounded by the current IP2 SFP CAOR.

Response

Column 6 in Table 1 contains a total estimated percent boron carbide loss determined by summing up the fractional change in areal density over all of the detector points along the entire length of the panel. Each fractional change in areal density is relative to the dose corrected minimum certified areal density to get a more accurate estimate of the boron carbide loss over the entire panel. The values in Column 6 provide an overall total mass loss estimate. The average areal density change (or uniform loss) for each panel tested in Region 1-2 is less than the total estimated loss and the total estimated loss is well below the peak loss. This indicates that the peak losses are limited to a small number of detector locations along the panel and the primary contributor to degradation, for the Region 1-2 panels tested in 2007, is uniform thinning .

RAl-2 (RCCB)

NRC Request In order for the staff to have reasonable assurance that the Boraflex will continue to perform its safety function, provide a comparison between the test results from the 2017 BADGER campaign and the RACKLIFE predicted degradation (e.g. was the RACKLIFE escape coefficient adjusted, continuing and future monitoring plans, etc.).

NL-17-102 Attachment 1 Page 5 of 7

Response

Table 3, below, contains a comparison of the BADGER total boron carbide loss, calculated relative to the dose corrected nominal areal density of the IP2 Region 1-2 Boraflex Panels.

The initial nominal areal density (at the nominal Boraflex thickness) prior to densification from gamma radiation was 0.0324 gms Boron-1 O/cm 2 from Reference (1] § 4.5.2.1 3 The BADGER results in Table 3 are calculated relative to nominal areal density for consistency in comparison to RACKLIFE model projections. For each of the panels in Table 3, the RACKLIFE predicted boron carbide loss exceeds the total estimated boron carbide loss as measured by BADGER.

Table 3: BADGER Measured and RACKLIFE Predicted Boron Carbide Loss Relative to the Nominal Areal Density of the IP2 Region 1-2 Boraflex Panels.

BADGER Estimated Total RACKLIFE 2.1 Boron Carbide Panel ID Projected Boron Loss(%) (from Carbide Loss Nominal Areal Density)

C19N -26.2 -31.3 ElSN -27.5 -30.7 HlSN -14.1 -14.5 HlSS -15.4 -44.0 F19S -5.1 -29.9 D21S -3.1 -13.4 F19N -9.3 -16.0 D21N -5.4 -12.5 D21W -9.1 -12.2 ElSW -17.5 -26.8 HlSE -7.9 -25.1 HlSW -12.9 -28.8 C19W -9.3 -23.1 F19W -4.8 -15.4 F19E -6.6 -13.3 D21E -10.4 -11.9 The RACKLIFE escape coefficients for Region 1-2 have been maintained at constant values of 25 (volumes/day) for the past 2.5 years in order to maintain the measured bulk pool silica concentration. Prior to that, the escape coefficient was set at 20 (volumes/day) from 2008 to 2015. While the escape coefficient may overestimate the boron carbide loss for the Region 1-2 panels measured, it would not be prudent to reduce the value for several reasons. First, the current escape coefficient is conservative with respect to predicted boron carbide loss.

3 Reference [1] : "Ind ian Point Unit 2 Spent Fuel Pool Increased Storage Capacity Licensing Report",

Consolidated Edison Company of New York; June 1988

NL-17-102 Attachment 1 Page 6 of 7 Second, it would require a series of measurements in Region 1-1 (and likewise for Region 2-1) in order to determine what fraction of boron carbide loss has already occurred in these modules. This is not practical , given the large number of fuel moves required to relocate the necessary fuel assemblies for a BADGER test. Lastly, the current spent fuel pool silica concentration is maintained around 1OOppm , which is effectively the temperature and pH dependent equilibrium concentration of reactive silica and is likely retarding the silica release rate from Boraflex panels.

RACKLIFE is typically run at six-month intervals to project out approximately 12 to 18 months ahead. It is the intention to continue this process until plant shutdown or approval of a License Amendment Request to eliminate Boraflex Credit.

RAl-3 (SNPB)

NRC Request The current IP2 and IP3 TSs limit the IP3 fuel that can be moved over to the IP2 SFP to fuel that has a maximum 4.4 weight percent (w/%) enrichment of uranium-235 (U235) , minimum 3.2 w/% enrichment of U235, and from operations earlier than IP3 Cycle 12. The restriction on IP3 fuel discharged prior to Cycle 12 carries with it an implicit minimum burn up of the fuel and minimum cooling time . These TS restrictions provided margin that the NRC staff relied upon in the initial licensing of the shield transfer canister. As initially proposed the LAR of December 14, 2016, would have revised the IP2 and IP3 TSs in a way that would have permitted the transfer of fresh unpoisoned fuel with a maximum 5.0 w/% enrichment of U235 from IP3 to IP2 SFP. This would have removed margin without providing sufficient justification for doing so. In response to the NRC staffs request for supplemental information , Entergy submitted the April 19, 2017, supplemental letter, which revised the LAR in that only the lower enrichment limit in the current TSs would be removed . Additionally, the supplemental letter provided information on the remaining limited population of fuel assemblies still residing in the IP3 SFP that would meet the stipulation that they be from operations earlier than IP3 Cycle 12. However, the supplemental letter did not revise the proposed TSs from the initial LAR letter submitted on December 14, 2016. Therefore , the NRC staff requests that the licensee, submit a revision to its proposed IP2 and IP3 TSs and/or proposed license conditions (consistent with the revision in the April 10. 2017, supplemental letter) that will control the transfer of IP3 fuel to the IP2 SFP.

Response

The revised proposed TSs , with changes stated in the April 19, 2017 supplemental submittal letter is in Attachment 2 to this letter. Specifically, only fuel assemblies with initial enrichment s; 4.4 w/% and discharged prior to IP3 Cycle 12 shall be stored in the IP2 Spent Fuel Pit.

Therefore , the TSs for IP2 , IP3 and the Shielded Transfer Canister (STC) have been revised accordingly.

NL-17-102 Attachment 1 Page 7 of 7 RAl-4 (SNPB)

NRC Request The NRC staff has determined that the vast majority of the proprietary markings in Hl-2094289 Appendix 4 .A are inappropriate as they do not meet the requirements in 10 CFR. Section 2.390(b)(iv) . The analysis follows NUREG/CR-6698, "Guide for Validation of Nuclear Criticality Safety Calculational Methodology" (ADAMS Accession No. ML012000481) , which is publicly available. The licensee's April 19, 2017, supplemental letter did not rectify this discrepancy.

Therefore , the NRC staff also requests that the licensee provide a revision to Hl-2094289 Appendix 4 .A with proprietary markings that are consistent with 10 CFR 2.390(b).

Response

The revised proprietary and non-proprietary versions of Hl-2094289 Appendix 4.A are in Attachments 3 and 4, respectively , to this letter.

ATTACHMENT 2 to NL-17-102 REVISED PROPOSED TS PAGE MARKUPS SEVEN (7) PAGES TOTAL ENTERGY NUCLEAR OPERATIONS , INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS. 50-247 AND 50-286

Spent Fuel Pit Storage 3.7.13 3.7 PLANT SYSTEMS 3.7.13 Spent Fuel Pit Storage LCO 3.7.13 lP2 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, bumup, 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 :

a. 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 Region 1-1 ;
b. Fuel assemblies that satisfy requirements of Figure 3.7.13-2 may be stored in any location in Region 2-2, Region 1-2 or Region 1-1 ;
c. Fuel assemblies that satisfy requirements of Figure 3.7.13-3 may be stored in any location in Region 1-2, Region 1-1 , or in locations designated as "peripheral" cells in Region 2-2; and
d. Fuel assemblies that satisfy requirements of Figure 3.7.13-4 may be stored :
1) ln any location in Region 1-2, or
2) ln a checkerboard loading configuration ( 1 out of every two cells with every other cell vacant) in Region 1-1 ; or
3) ln locations designated as "peripheral" cells in Region 2-2.

fP3 fuel assemblies shall be stored in Region 1-2 of the Spent Fuel Pit.

Only assemblies with initial enrichment > 3 .2 aAd ::; 4.4 w/o U235 and discharged prior to IP3 Cycle 12 shall be stored in the Spent Fuel Pit. IP3 fuel assemblies V43 and V48 are not approved for storage in the Spent Fuel Pit.

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

INDIAN POINT 2 3.7.13-1 Amendment No. ~TBD

STC Loading 3.1.2 3.1 INTER-UNIT FUEL TRANSFER 3.1.2 Shielded Transfer Canister (STC) Loading LCO 3.1.2 INT ACT FUEL ASSEMBLIES placed into the Shielded Transfer Canister (STC) shall be classified in accordance with Table 3.1.2-1 based on initial enrichment and burnup and shall be restricted based on the following:

a. INT ACT FUEL ASSEMBLIES classified as Type 2 may be placed in the STC basket (see Figure 3.1.2-1) with the following restrictions:

I . Post-irradiation cooling time, initial enrichment, and allowable average burnup shall be within the limits for the cell locations as specified in Table 3.1.2-3;

2. Decay heat including NON FUEL HARDWARE ::; 650 \l/atts (eells 5thr01:1gh

~ 1.2 kW (any cell );

3. Deeay heat iRel1:1diRg NON FUE L HARDWARE S 1105 \Vatts ( eell 1,2, 3 or 4-fTotal ST C Decay heat from a ll cell l o cation s including NON FUEL HARDWARE ::; 9 .621 kW ;
4. Post-irradiation cooling time and the maximum average burnup of NON FUEL HARDWARE shall be within the cell locations and limits specified in Table 3.1.2-2. In accordance with Table 3.1.2-2 RCCAs and Hafnium Flux Suppressors cannot be placed in locations 5, 6, 7, 8, 9, 10, 11 , 12 of the STC basket.

-NOTE-If one or more Type I fuel assemblies are in the STC, cells 1, 2, 3, AND 4 must be empty, with a cell blocker installed that prevents inserting fuel assemblies and/or NON-FUEL HARDWARE .

b. INT ACT FUEL ASSEMBLIES classified as Type I or Type 2 may be placed in locations 5, 6, 7, 8, 9, I 0, 11 , 12 of the STC basket (see Figure 3.1.2-1) with the following restrictions:

I . Post-irradiation cooling time, initial enrichment, and allowable average burnup shall be within the limits for the cell locations as specified in Table 3.1.2-3 ;

2. Decay heat including NON FUEL HARDWARE ::; 650 Wattsl.2 kW ;
3. Post-irradiation cooling time and the maximum average burnup of NON FUEL HARDWARE shall be within the cell locations and limits specified in Table 3.1 .2-2. In accordance with Table 3. 1.2-2 RCCAs and Hafnium Flux Suppressors cannot be placed in locations 5, 6, 7, 8, 9, 10, 11 , 12 of the STC basket.
c. Only INTACT FUEL ASSEMBLIES with initial average enrichment ~

aittl--::; 4.4 wt% U-235 and discharged prior to IP3 Cycle 12 shall be placed in the STC basket. IP3 fuel assemblies V43 and V48 shall not be selected for transfer.

INDIAN POINT 2 3.1.2-1 Amendment ~TBD

STC Loading 3.1.2 Table 3. 1.2-3 (Sheet 1 of 2)

Allowable STC Loading Configurations Configuration<cl Cells 1, 2, 3, 4<*Xb) Cells 5, 6, 7, 8, 9, 10, 11 , 12<*Xbl Burnup :S 55 ,000 MWD/MTU Burnup :S 40,000MWD/ MTU I Cooling time ~ I 0 years Cooling time ~ 25 years Initi a l Enrichment ~ 3 .4 wt% U-235 Initial E nrichment ~ 2.3 wt% U-235 Burn up :S 45,000 M WD/MTU Burn up :S 45,000 M WD/MTU 2 Cooling time ~ 10 years Cooling ti me ~ 20 years Initi al E nrichment ~ 3.2 wt% U-235 Initial E nrichment ~ 3 .2 wt% U-235 Burnup :S 55,000 Burnup :S 45,000 MWD/MTU 3 MWD/MTU Cooling time ~ Cooling time ~ 20 years 10 years Initi al Enrichment ~ 3.2 wt% U-235 Initia l Enrichment ~ 3.4 wt% U-235 Burnup :S 45 , 000 Burnup :S40,000 MWD/MTU 4 MWD/MTU Cooling time ~ Cooling time ~ 12 years 10 years Initial Enrichment ~ 3.2 wt% U-235 Initi al E nrichment ~ 3.6 wt% U-235 Burnup :S 45 , 000 Burnup :S 40,000 MWD/MTU 5 MWD/MTU Cooling time ~ Cooling time~ 12 years 14 years Initi al E nrichment ~ 3.2 wt% U-235 Initi a l Enrichment ~ 3.4 wt% U-235 Bumup :S 45,000 MWD/MTU Burn up :S 40,000 MW D/MTU 6 Cooling time ~ 20 years Cooling time~ 20 years Initi al Enrichment ~ 3.2 wt% U-235 Initial Enrichment ~ 2.3 wt% U-235 INDIAN POINT 2 3.1.2-6 Am endment ~TBD

STC Loading 3.1. 2 Table 3 .1 .2-3 (Sheet 2 of 2)

Allowable STC Loading Configurations Configuration<cl Cells l ,2, 3, 4(aXb) Ce lls 5, 6, 7, 8, 9, 10, 11 , 12<*Xbl Burnup :S 45,000 MWD/MTU Burnup :S 45 ,000 MWD/MTU 7

Cooling time ~ I 0 years Cooling time ~ 12 years Initial Enrichment ~ 3.2 wt% U-235 Initial Enrichment ~ 3.2 wt% U-235 Burnup :S 55 ,000 MWD/MTU Burnup :S 55 ,000 MWD/MT U 8

Cooling time ~ 10 years Cooling time ~ 15 years Initial Enrichment ~ 3.4 wt% U-235 Initial Enrichment ~ 3.4 wt% U-235 Burnup :S 55,000 MWD/MTU Burnup :S 45 ,000 MWD/MTU 9

Cooling time ~ 11 years Cooling time ~ 12 years Initial Enrichment ~ 3.4 wt% U-23 5 Initial Enrichment ~ 3.2 wt% U-235 Burnup :S 45 ,000 MWD/MTU Burnup :S 55 ,000 MWD/MTU 10 Cooling time ~ 10 years Cooling time ~ 15 years Initial Enrichment ~ 3.2 wt% U-235 Initial Enrichment ~ 3.4 wt% U-235 Burnup :S 45 ,000 MWD/MT U Burnup :S 45 ,000 MWD/MTU 11 Cooling time ~ 6 years Cooling time ~ 14 years Initial Enrichment~ 3.2 wt% U-235 Initial Enrichment ~ 3.2 wt% U-235 Burnup :S 60,000 MWD/MTU Burnup :S 50,000 MWD/MTU 12 Cool ing time ~ 9 years Cooling time ~ 14 years Initial Enrichment ~ 4.2 wt% U-235 Initial Enrichment ~ 3.6 wt% U-235 (a) Initial enrichment is the assembly average enrichment. Natural or enriched uranium blankets are not considered in determining the fuel assembly average enrichment for comparison to the minimum allowed initial average enrichment.

(b) Rounding to one decimal place to determine initial enrichment is permitted.

(c) Fuel with five middle lnconel spacers are limited to cells I, 2, 3, and 4 for all loading configurations except loading configuration 6 which allows fuel with lnconel spacers in all cells.

INDIAN POINT 2 3.1.2-7 Amendment ~TB D

STC Loading 3.1.2 3.1 INTER-UNIT FUEL TRANSFER 3.1.2 Shielded Transfer Canister (STC) Loading LCO 3.1.2 INT ACT FUEL ASSEMBLIES placed into the Shielded Transfer Canister (STC) shall be classified in accordance with Table 3 .1.2-1 based on initial enrichment and burn up and shall be restricted based on the following:

a. INT ACT FUEL ASSEMBLIES classified as Type 2 may be placed in the STC basket (see Figure 3. 1.2-1 ) with the following restrictions:

I . Post-irradiation cooling time, initial enrichment, and allowable average burnup shall be within the limits for the cell locations as specified in Table 3.1.2-3;

2. Decay heat including NON FUEL HARDWARE :S 650 \Vatts (eells 5thre1:1gh

+;B 1.2 kW (any cell);

3. Deea)' heat iAel1:1diAg NON FUBb HARDWARE :::; 1105 Watts (eelll,2, 3 er 4-}Total STC Decay heat from all cell locations including NON FUEL HARDWARE :S 9 . 6 2 1 k W ;
4. Post-irradiation cooling time and the maximum average burnup of NON FUEL HARDWARE shall be within the cell locations and limits specified in Table 3.1.2-2. In accordance with Table 3.1.2-2 RCCAs and Hafuium Flux Suppressors cannot be placed in locations 5, 6, 7, 8, 9, 10, 11, 12 of the STC basket.

- NOTE -

If one or more Type I fuel assemblies are in the STC, cells I , 2, 3, AND 4 must be empty, with a cell blocker installed that prevents inserting fuel assemblies and/or NON-FUEL HARDWARE.

b. INTACT FUEL ASSEMBLIES classified as Type I or Type 2 may be placed in locations 5, 6, 7, 8, 9, I 0, 11 , 12 of the STC basket (see Figure 3.1.2-1) with the following restrictions:

I. Post-irradiation cooling time, initial enrichment, and allowable average burnup shall be within the limits for the cell locations as specified in Table 3.1.2-3;

2. Decay heat including NON FUEL HARDWARE :S 650 Watts I .2 kW ;
3. Post-irradiation cooling time and the maximum average burnup of NON FUEL HARDWARE shall be within the cell locations and limits specified in Table 3.1.2-2. In accordance with Table 3. I .2-2 RCCAs and Hafnium Flux Suppressors cannot be placed in locations 5, 6, 7, 8, 9, I 0, I 1, 12 of the STC basket.
c. Only INTACT FUEL ASSEMBLI ES with initial average enric hment ~

ami-:S 4.4 wt% U-235 and discharged prior to IP3 Cycle 12 shall be placed in the STC basket. IP3 fuel assemblies V43 and V48 shall not be selected for transfer.

INDIAN POINT 3 3.1.2-1 Amendment ~TBD

STC Loading

3. 1.2 Table 3.1.2-3 (Sheet I of 2)

Allowable STC Loading Configurations Configuration<cl Cells 1, 2, 3, 4<*Xbl Ce II s 5, 6, 7, 8, 9, 10, 11 , I 2<*Xbl Burnup S 55,000 MWD/MTU Burnup S 40,000MWD/MTU I Cooling time 2: I 0 years Cooling t ime 2: 25 years Initial Enrichment 2: 3.4 wt% U-235 Initial Enrichment 2: 2.3 wt% U-235 Burnup S 45,000 MWD/MTU Burnup S 45,000 MWD/MTU 2 Cooling time 2: I 0 years Cooling time 2: 20 years Initial Enrichment 2: 3.2 wt% U-235 In itial Enrichment 2: 3 .2 wt% U-235 Burnup S 55 ,000 Burnup S 45,000 MWD/MTU 3 MWD/MTU Cooling time 2: Cooling time 2: 20 years I 0 years Initi al Enrichment 2: 3.2 wt% U-235 Initial En richm ent 2: 3.4 wt% U-235 Burnup S 45 ,000 Burnup S 40,000.MWD/MTU 4 MWD/MTU Cooli ng time 2: Cooling time 2: 12 years 10 years Initial Enrichment 2: 3.2 wt% U-235 lnitial Enrichment 2: 3.6 wt% U-235 Burnup S 45 , 000 Burnup S 40,000 MWD/MTU 5 MWD/MTU Coo ling time 2: Cooling time 2: 12years 14 years Initial Enrichrnent 2: 3.2 wt% U-235 Initial Enrichment 2: 3.4 wt% U-235 Burnup S 45,000 MWD/MTU Burnup S 40,000 MWD/MTU 6 Cooling time 2: 20 years Cooling time 2: 20 years Initial Enrichment 2: 3.2 wt% U-235 Initial Enrichment 2: 2.3 wt% U-235 INDIAN POINT 3 3.1.2-6 Amendment 24&TBD

STC Loading 3.1.2 Table 3.1.2-3 (Sheet 2 of2)

Allowable STC Loading Configurations Configuration<cJ Cells I, 2, 3, 4<*Xbl Cells 5, 6, 7, 8, 9, IO, 11 , 12<*Xbl Burnup :S 45,000 MWD/MTU Burnup :S 45,000 MWD/MTU 7

Cooling time 2: 10 years Cooling time 2: 12 years Initial Enrichment 2: 3.2 wt% U-235 Initial Enrichment 2: 3.2 wt% U-235 Burnup :S 55 ,000 MWD/MTU Burnup :S 55,000 MWD/MTU 8

Cooling time 2: 10 years Cooling time 2: 15 years Initial Enrichment 2: 3.4 wt% U-235 Initial Enrichment 2: 3.4 wt% U-235 Burnup :S 55,000 MWD/MTU Burnup :S 45 ,000 MWD/MTU 9

Cooling time 2: 11 years Cooling time 2: 12 years Initial Enrichment 2: 3.4 wt% U-235 Initial Enrichment 2: 3.2 wt% U-235 Burnup :S 45,000 MWD/MTU Burnup :S 55,000 MWD/MTU 10 Cooling time 2: 10 years Cooling time 2: 15 years Initial Enrichment 2: 3.2 wt% U-235 Initial Enrichment 2: 3.4 wt% U-235 Burnup :S 45 ,000 MWD/MTU Burnup :S 45 ,000 MWD/MTU 11 Cooling time 2: 6 years Cooling time 2: 14 years Initial Enrichment 2: 3.2 wt% U-235 Initial Enrichment 2: 3.2 wt% U-235 Burnup :S 60,000 MWD/MTU Burnup :S 50,000 MWD/MTU 12 Cooling time 2: 9 years Cooling time 2: 14 years Initial Enrichment 2: 4.2 wt% U-235 Initial Enrichment 2: 3.6 wt% U-235 (a) Initial enrichment is the assembly average enrichment. Natural or enriched uranium blankets are not considered in determining the fuel assembly average enrichment for comparison to the minimum allowed initial average enrichment.

(b) Rounding to one decimal place to determine initial enrichment is permitted.

(c) Fuel with five middle lnconel spacers are limited to cells 1, 2, 3, and 4 for all loading configurations except loading configuration 6 which allows fuel with lnconel spacers in all cells.

INDIAN POINT 3 3.1.2-7 Amendment ~TSO

ATTACHMENT 5 to NL-17-102 AFFIDAVIT IN SUPPORT OF REQUEST TO WITHHOLD INFORMATION FIVE (5) PAGES TOTAL ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 & 3 DOCKET NOS. 50-247 AND 50-286

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk AFFIDAVIT PURSUANT TO 10 CFR 2.390 I, Kimberly Manzione, being duly sworn, depose and state as follows:

(1) I have reviewed the information described in paragraph (2) which is sought to be withheld, and am authorized to apply for its withholding.

(2) The information sought to be withheld is information provided in APPENDIX 4.A (BENCHMARK CALCULATIONS) of LICENSING REPORT ON THE INTER-UNIT TRANSFER of SPENT NUCLEAR FUEL at THE INDIAN POINT ENERGY CENTER (HI-2094289 Rev. 8).

This document contains Holtec Proprietary information.

(3) In making this application for withholding of proprietary information of which it is the owner, Holtec International relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4) and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10CFR Part 9.17(a)(4), 2.390(a)(4), and 2.390(b)(l) for "trade secrets and commercial or financial information obtained from a person and privileged or confidential" (Exemption 4). The material for which exemption from disclosure is here sought is all "confidential commercial information",

and some portions also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission , 975F2d871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704F2dl280 (DC Cir. 1983).

I of 5

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk AFFIDAVIT PURSUANT TO 10 CFR 2.390 (4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by Holtec's competitors without license from Holtec International constitutes a competitive economic advantage over other companies;
b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product.
c. Information which reveals cost or price information, production, capacities, budget levels, or commercial strategies of Holtec International, its customers, or its suppliers;
d. Information which reveals aspects of past, present, or future Holtec International customer-funded development plans and programs of potential commercial value to Holtec International;
e. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs 4.a and 4.b above.

(5) The information sought to be withheld is being submitted to the NRC in confidence. The information (including that compiled from many sources) is of a sort customarily held in confidence by Holtec International, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by Holtec International. No public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to the NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for 2 of5

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk AFFIDAVIT PURSUANT TO 10 CFR 2.390 maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge. Access to such documents within Holtec International is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his designee ), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside Holtec International are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information classified as proprietary was developed and compiled by Holtec International at a significant cost to Holtec International. This information is classified as proprietary because it contains detailed descriptions of analytical approaches and methodologies not available elsewhere. This information would provide other parties, including competitors, with information from Holtec International's technical database and the results of evaluations performed by Holtec International. A substantial effort has been expended by Holtec International to develop this information. Release of this information would improve a competitor's position because it would enable Holtec's competitor to copy our technology and offer it for sale in competition with our company, causing us financial mJury.

3 of5

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk AFFIDAVIT PURSUANT TO 10 CFR2.390 (9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to Holtec International's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of Holtec International's comprehensive spent fuel storage technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical methodology, and includes development of the expertise to determine and apply the appropriate evaluation process.

The research, development, engineering, and analytical costs comprise a substantial investment of time and money by Holtec International.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

Holtec Intemational's competitive advantage will be lost if its competitors are able to use the results of the Holtec International experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to Holtec International would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive Holtec International of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing these very valuable analytical tools.

4 of5

U.S. Nuclear Regulatory Commission AT1N: Document Control Desk AFFIDAVIT PURSUANT TO 10 CFR 2.390 STATE OF NEW JERSEY )

) ss:

COUNTY OF CAMDEN )

Kimberly Manzione, being duly sworn, deposes and says :

That she has read the foregoing affidavit and the matters stated therein are true and correct to the best of her knowledge, information, and belief.

Executed at Camden, New Jersey, this 10th day of August, 2017.

~

? /I Vl6..) c~

Kimberly Manzione Licensing Manager Holtec International Subscribed and sworn before me this 10th day of August, 201 7.

- MARIA C. MASSI NOT'ARV PUBLIC OF NEW JERSEY My Commission Expires Aprll 25, 2020 5of5

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