NL-12-007, Response to Request for Additional Information Regarding the Inter-Unit Spent Fuel Transfer License Amendment Request (TAC Nos. ME1671, ME1672, and L24299)

Response to Request for Additional Information Regarding the Inter-Unit Spent Fuel Transfer License Amendment Request (TAC Nos. ME1671, ME1672, and L24299)

ML12074A027
Person / Time
Site:	Indian Point
Issue date:	03/02/2012
From:	Ventosa J Entergy Nuclear Northeast
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-12-007, TAC L24299, TAC ME1671, TAC ME1672
Download: ML12074A027 (37)

Text

Enteray Nuclear Northeast Indian Point Energy Center A k E450 Broadway, GSB Ente yBuchanan, P.O. Box 249 NY 10511-0249 Tel 914 254 6700 John A. Ventosa Site Vice President Administration NL-12-007 March 2, 2012 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Station O-Pl-17 Washington, DC 20555-0001

Subject:

Indian Point Nuclear Power Plant Units 2 and 3 Response to Request for Additional Information Regarding the Inter-Unit Spent Fuel Transfer License Amendment Request (TAC Nos. ME1671, ME1672, and L24299)

Indian Point Units 2 & 3 Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64

References:

1) NRC letter to Indian Point Vice President of Operations, 11/09/11, "Indian Point Nuclear Generating Unit Nos. 2 and 3 - Request for Additional Information Regarding Amendment Application for Inter-Unit Spent-F.uel Transfer (TAC Nos. ME1671, ME1672, and L24299)"

2) Entergy letter NL-09-076, 07/08/09, "Indian Point Nuclear Power Plant Units 2 and 3 - Application for Unit 2 Operating License Condition Change and Units 2 and 3 Technical Specification Changes to Add Inter-Unit Spent Fuel Transfer Requirements"

3) Entergy letter NL-09-100, 09/28/09, "Indian Point Nuclear Power Plant Units 2 and 3 - Response to Request for Supplemental Information Regarding the Spent Fuel Transfer License Amendment Request (TAC Nos. ME1671, ME1672, and L24299)"

4) Entergy letter NL-10-093, 10/05/10, "Indian Point Nuclear Power Plant Units 2 and 3 - Response to Request for Additional Information Regarding the Inter-Unit Spent Fuel Transfer License Amendment Request (TAC Nos. ME1671, ME1672, and L24299)"

5) Entergy letter NL-1 1-052, 07/28/11, "Indian Point Nuclear Power Plant Units 2 and 3 - Response to Request for Additional Information Regarding the Inter-Unit Spent Fuel Transfer License Amendment Request (TAC Nos. ME1671, ME1672, and L24299)"

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NL-12-007 Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64 Page 2 of 3

6) Entergy letter NL-11-118, 10/28/11, "Indian Point Nuclear Power Plant Units 2 and 3 - Response to Request for Additional Information Regarding the Inter-Unit Spent Fuel Transfer License Amendment Request (TAC Nos. ME1671, ME1672, and L24299)"

7) Entergy letter NL-1 1-130, 12/15/11, "Indian Point Nuclear Power Plant Units 2 and 3 - Response to Request for Additional Information Regarding the Inter-Unit Spent Fuel Transfer License Amendment Request (TAC Nos. ME1 671, ME1672, and L24299)"

Dear Sir or Madam:

This letter provides Entergy Nuclear Operations, Inc (Entergy) additional response to the NRC Request for Additional Information (RAI) (Reference 1) regarding the Entergy license amendment requests concerning inter-unit transfer of fuel (Reference 2), the supplement to the amendment request (Reference 3), and the responses to previous RAIs (References 4 through 7). In reference 7 Entergy submitted responses to all Reference 1 RAIs with the exception of RAI 14.

The response to RAI 14 is provided in Attachment 1. As identified in the response there will be changes to the previously proposed IP2 and IP3 Appendix C TS and there will also be a new proposed change to IP2 Appendix A TS LCO 3.7.13. The response also identifies changes to the licensing report. Both the TS and licensing report changes will be submitted separately.

The licensing report will also be revised to reflect the fact that Entergy is procuring a new HI-TRAC that will be used for fuel transfers instead of the existing HI-TRAC. The procurement of the new HI-TRAC has resulted in minor editorial changes to the licensing report. In addition, due to manufacturing tolerances, the lead thickness, overall weight, and overall outside diameter of the outer shell are slightly greater than for the existing HI-TRAC. These slight changes, which were within defined tolerances, have been evaluated and do not result in changes to the Holtec evaluations and reports previously submitted in support of this amendment request and there are no changes to the proposed Appendix C Technical Specifications including Part I that provide a description of the HI-TRAC and its critical dimensions. As noted in Section 1.5 of the licensing report the Part 72 HI-TRAC licensing drawings were submitted for information only.

During the fabrication of the new HI-TRAC these drawings have been the subject of two 10 CFR 72.48 evaluations. However, as these drawings were submitted for information only, it is not intended to submit new drawings for these or potentially future 10CFR72.48 evaluations.

In addition, in order to minimize or eliminate corrosion, the seal surfaces around the STC Vent and Drain Ports and the HI-TRAC Port Cover Lid will be provided with stainless steel weld overlay and for the associated bolt holes stainless steel bushings will be installed. These changes will also be included in the revised licensing report, but do not affect any of the results or conclusions of evaluations previously submitted.

NL-12-007 Docket Nos. 50-247 and 50-286 License Nos. DPR-26 and DPR-64 Page 3 of 3 Enclosures 1 and 2 contain non proprietary and proprietary Westinghouse reports, respectively.

These reports support certain of the RAI responses as identified in Attachment 1.

This submittal includes information deemed proprietary by an entity that is providing support to Entergy on this project. As such, in Enclosure 3, a 10 CFR 2.390 affidavit has been executed by the owner of the information.

There are no new regulatory commitments in this submittal.

In accordance with 10 CFR 50.91, a copy of this submittal is being provided to the designated New York State official.

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

I declare under penalty of perjury that the foregoing is true and correct to the best of my knowledge. Executed on  ?-*-*o* -

Sincerely, JV/rw Attachments and

Enclosures:

Attachment 1: Response to Request for Additional Information (RAI-14) (Non Proprietary)

Enclosure 1: Attachments 1, 2 and 3 to Westinghouse Letter NF-IN-12-4 (Non Proprietary)

Enclosure 2: Attachment 4 to Westinghouse Letter NF-IN-12-4 and Cover Letter (Proprietary)

Enclosure 3: Attachment 5 to Westinghouse Letter NF-IN-12-4 (Non Proprietary)

Application for Withholding Proprietary Information from Public Disclosure cc: NRC Resident Inspector's Office Mr. John Boska, Senior Project Manager, NRC NRR DORL Mr. William M. Dean, Regional Administrator, NRC Region 1 Mr. Francis J. Murray Jr., President and CEO, NYSERDA Ms. Bridget Frymire, New York State Dept. of Public Service

ATTACHMENT 1 TO NL-12-007 RESPONSE to REQUEST for ADDITIONAL INFORMATION (NON PROPRIETARY)

Entergy Nuclear Operations, Inc.

Indian Point Units 2 and 3 Docket Nos. 50-247 and 50-286

NL-12-007 Attachment 1 Page 1 of 11 REQUEST FOR ADDITIONAL INFORMATION REGARDING SPENT FUEL TRANSFER ENTERGY NUCLEAR OPERATIONS, INC.

INDIAN POINT NUCLEAR GENERATING UNIT NOS. 2 AND 3 DOCKET NOS. 50-247 AND 50-286 By letter dated July 8, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession Nos. ML091940177 and ML091940178), as supplemented by letters dated September 28, 2009, (ADAMS Accession Nos. ML092950437 and ML093020080), October 5, 2010 (ADAMS Accession Nos. ML102910511, ML103080112, and ML103080113), July 28, 2011 (ADAMS Accession Nos. ML11220A079, ML112200258), and October 28, 2011, Entergy Nuclear Operations, Inc. (Entergy or the licensee), submitted a license amendment request for Indian Point Nuclear Generating Unit Nos. 2 and 3 (IP2 and IP3). The proposed changes are requested to provide the necessary controls and permission required for Entergy to move spent fuel from the IP3 spent fuel pool (SFP) to the IP2 SFP using a newly designed. shielded transfer canister (STC), which is placed inside a HI-TRAC 10OD cask for outdoor transport. The chapter listed below refers to the safety analysis report (SAR) for the STC, HI-2094289, Revision 4, ADAMS Accession No. ML112200258. The Nuclear Regulatory Commission (NRC) staff is reviewing the submittal and has the following question:

CHAPTER 4 - CRITICALITY EVALUATION (SRXB and CSDAB)

NRC RAI 14 In Section 4.8 of Holtec Report HI-2094289 (Reference 1), it discusses the IP2 SFP nuclear criticality design bases analysis with respect to the IP2 fuel and depletion parameters. It is not clear from the information provided that it is acceptable to store the IP3 spent nuclear fuel (SNF) in the IP2 SFP without further analysis. With respect to the comparison between the IP2 SFP nuclear criticality safety (NCS) design basis analysis (DBA) and IP3 SNF provide the following information.

a. Discuss the use of integral and fixed neutron absorbers in the IP2 SFP NCS DBA and IP3 SNF. Identify any cases where the IP2 SFP NCS DBA does not bound the IP3 SNF.

b. Discuss the use of other fuel assembly inserts such as Hafnium flux suppressors and control rods in the IP2 SFP NCS DBA and IP3 SNF. Identify any cases where the IP2 SFP NCS DBA does not bound the IP3 SNF.

c. Holtec Report HI-2094289 Table 4.8.2 indicates the IP2 SFP NCS DBA used a core exit temperature of 624 OF for the depletion analysis. Table 3.2-4 of the IP3 UFSAR Rev. 3, 2009, says the Nominal Outlet of the Hot Channel for Cycle 1 was 635.7 OF. This table also indicates the average temperature rise in the core has increased from Cycle 1, indicating the Nominal Outlet of the Hot Channel has also likely increased as well. The STC NCS analysis used an IP3 core exit temperature of 637.3 OF. The IP3 UFSAR and the STC NCS analysis indicate that using a core exit temperature of 624 OF is not bounding for the IP3 SNF. Justify using the IP2 SFP storage criteria for storing IP3 SNF given the apparent use of a non-bounding core exit temperature in the IP2 SFP NCS DBA.

Response to Request for Additional Information NL-12-007 Attachment 1 Page 2 of 11

d. Holtec Report HI-2094289 discusses the method that was used to identify the axially distributed burnup profile in the IP2 SFP NCS DBA and concludes this method would also be applicable to IP3. While the NRC staff takes no position on that assertion, Holtec Report HI-2094289 provides no information that would allow the staff to evaluate whether or not the axially distributed burnup profile actually used in the IP2 SFP NCS DBA is bounding for the IP3 SNF. Describe how the axially distributed burnup profile used in the IP2 SFP NCS DBA is bounding for the IP3 SNF. Compare the axially distributed burnup profile actually used in the IP2 SFP NCS DBA with the axially distributed burnup profile used in the STC NCS analysis and explain why that method of determining the profile in the STC NCS analysis was used over the method that was used to identify the axially distributed burnup profile in the IP2 SFP NCS DBA.

e. The NRC staff expects a thorough evaluation of the IP2 SFP NCS DBA and IP3 SNF addressing all of the topics in the STC NCS analysis at a minimum and including additional items as the licensee's due diligence identifies.

f. Based on the above, identify and justify any additional measures that need to be taken to ensure the IP3 SNF can be safely stored in the IP2 SFP.

Response to RAI 14 Section 4.8 of Holtec Report HI-2094289 discusses the IP2 SFP nuclear criticality design bases analysis with respect to the IP2 fuel and depletion parameters. In order to demonstrate the acceptability of storing IP3 SNF in the IP2 SFP additional evaluations and analyses have been perfornied that consisted of:

1. A comparison between IP2 and IP3 fuel and reactor operation.

2. Qualitative and quantative evaluations of the differences in fuel design and reactor operation.

The results of these evaluations and analyses have led Entergy to conservatively propose restricting placement of IP3 SNF into Regions 1-1 and 1-2 of the IP2 SFP. These regions are qualified for higher reactivity fuel and, as discussed below, offset differences in fuel design and reactor operation. Entergy is also proposing to restrict the IP3 SNF to be transferred to that discharged from Cycles 1 through 11 and to initial enrichments of > 3.2 and < 4.4 w/o U235.

These restrictions are explained below in the detailed RAI responses.

a. IP2 Burnable Absorbers Integrated Fuel Burnable Absorbers (IFBA), Wet Annular Burnable Absorber (WABA) assemblies and borated Pyrex glass rods (clad in stainless steel) assemblies have been incorporated into IP2 core designs. IFBAs are the only integral and fixed neutron absorbers utilized. WABAs and the Pyrex glass assemblies are the only removable absorbers utilized.

IP3 Burnable Absorbers The IP3 core designs have also utilized integral and fixed neutron absorbers and removable neutron absorbers. IFBAs are the only integral and fixed neutron absorbers utilized. Removable neutron absorbers include Pyrex glass assemblies and WABAs.

Response to Request for Additional Information NL-12-007 Attachment 1 Page 3 of 11 For both IP2 and IP3, the Pyrex assemblies consist of 4 to 20 absorber rodlets per assembly.

Although not burnable absorbers the IP3 cores also include the presence of hafnium flux suppressor (HFS) assemblies, to reduce neutron fluence in the vicinity of the reactor vessel inner wall. These devices, which do not appear in the IP2 core, are discussed separately in the response to RAI 14b below.

Burnable Absorbers and the IP2 SFP Criticality Analysis The IP2 SFP criticality analysis (Reference 1) explicitly modeled the effect of WABAs on spectrum hardening. In that analysis, the limiting effect from WABA insertion was calculated to be +0.00951 Akff, which was rounded to a penalty of +0.01 Akeff in the reactivity roll-up. In accordance with the then accepted methodology, the IP2 SFP criticality analysis did not explicitly model the effects of IFBAs, except to credit their presence in unirradiated fuel stored in the SFP.

In order to assess the spectrum hardening effects of the IP3 fixed and removable burnable absorbers, Westinghouse (Attachment 1) has independently performed an evaluation using current criticality modeling methodology. The evaluation included Pyrex absorber assemblies containing 20, 16 and 12 rodlets, which bound all Pyrex configurations. Also included in the analysis is the limiting IFBA / WABA configuration of 20 WABA / 80 IFBA, which bounds all IFBA / WABA combinations for IP3 Cycles 1 through 11. The calculations to determine the reactivity effect associated with burnable absorber usage were performed using the U. S. NRC licensed depletion code Paragon to develop spent fuel isotopic concentrations and KENO V.a to determine spent fuel pool rack reactivity. The reactivity effect was determined by depleting an assembly with and without burnable absorbers. The statistical uncertainty associated with the KENO V.a calculation was incorporated into the results at a 95/95 confidence level.

The Westinghouse analysis assumed the following:

• An initial fuel enrichment of 3.2 to 4.4 w/o U21. This range encompasses the initial enrichment of all fuel that operated in the IP3 core and was discharged from Cycles 1 through 11. This range has been incorporated into a new proposed revision to IP2 Appendix A TS LCO 3.7.13 Spent Fuel Pit Storage.

• IP3 SNF to be stored in the IP2 SFP is restricted to that fuel that operated in the IP3 core and was discharged from Cycles 1 through 11. This restriction has been incorporated into a new proposed revision to IP2 Appendix A TS LCO 3.7.13 Spent Fuel Pit Storage.

• IP3 SNF to be stored in the IP2 SFP is restricted to IP2 SFP Regions 1-1 and 1-2 without further restriction. In other words, there are no special designations as to which of the two regions where an IP3 fuel assembly may be stored.

• No credit in the analysis has been taken for decay of short- and medium-lived isotopes in the discharged IP3 SNF. As the cooling times of the fuel assemblies scheduled for transfer vary from 11 to 30 years, this represents a significant margin.

Response to Request for Additional Information NL-12-007 Attachment 1 Page 4 of 11 The results of the analysis (Enclosure 1) are as follows:

• The limiting configuration for Pyrex glass burnable absorbers is the 20-rodlet assembly. The limiting configuration for IFBA, WABA or IFBANWABA combo is the 20 WABA / 80 IFBA combination.

• The resultant maximum reactivity penalty was 0.02270 Akeff for Pyrex and 0.02200 Akeff for IFBA / WABA burnable absorbers. These reactivity penalties exceed the 0.01 Ake, reactivity bias used in the original (2001) IP2 SFP criticality analysis to account for spectrum hardening effects arising from the presence of WABA burnable poisons. The IP2 SFP criticality analysis used the modeling methodologies available at the time. Since the IP2 analysis included a reactivity bias of 0.01 Akeff, it is necessary to offset an additional bias of 0.01270 Ake,. The analysis determined that a significant amount of reactivity margin is available in the Region 1-1 and Region 1-2 IP2 SFP racks to provide the necessary offset.

Specifically, the minimum and maximum reactivity margins between initial enrichments of 3.2 and 4.4 w/o U235 were determined to be 0.03086 and 0.05999 Ake, which are well in excess of the 0.01270 Akeff offset required.

In summary, when considering burnable absorbers, in order to maintain the validity of the IP2 spent fuel pool criticality analysis with IP3 fuel in the pool the IP3 fuel assemblies shall be stored in Region 1-1 or Region 1-2 of the Spent Fuel Pit and:

a. The fuel assembly discharge Cycle >- 1 and < 11.

b. The fuel assembly initial enrichment > 3.2 and < 4.4 w/o U235.

Section 4.8 of Holtec Report HI-2094289 will be revised to incorporate the results of the analysis described above and the IP3 SNF storage restrictions will be incorporated into a new proposed revision to IP2 Appendix A TS LCO 3.7.13 Spent Fuel Pit Storage and Appendix TS LCO 3.1.2 Shielded Transfer Canister (STC) Loading. The TS Bases will be revised accordingly.

b. IP2 Fuel Assembly Inserts The removable inserts in the IP2 core fall into the following categories: Pyrex glass burnable absorbers, WABA burnable absorbers, Rod Control Cluster Assemblies (RCCAs), primary and secondary neutron sources, and thimble plugs.

IP3 Fuel Assembly Inserts The IP3 core designs have also utilized the removable inserts identified for IP2 and, in addition, Hafnium Flux Suppressors have been utilized.

Fuel Assembly Inserts and the IP2 SFP Criticality Analysis The effects of Pyrex absorbers and WABAs have been discussed in the response to RAI 14a.

Response to Request for Additional Information NL-12-007 Attachment 1 Page 5 of 11 Primary and secondary neutron sources and thimble plugs are not specifically accounted for in the IP2 SFP criticality analysis. Neutron sources are mounted in the guide tubes in a fashion similar to discrete burnable absorbers, such as Pyrex, WABA, and Hafnium inserts. The bounding burnable absorber reactivity effect calculated as part of the response to RAI 14a includes the presence of a 20 finger Pyrex absorber during multiple cycles. Because the neutron sources have fewer fingers and do not contain strong thermal neutron absorbers such as the boron contained in Pyrex, the spectral hardening effect of the neutron sources during depletion is bounded by the bias calculated with Pyrex. Even though the neutron sources are bounded by the Pyrex bias, Entergy has conservatively determined that IP3 neutron sources will not be transferred to the IP2 SFP at this time. Proposed Appendix C TS LCO 3.1.2 STC Loading and Chapter 7 of the licensing report will be revised accordingly.

The thimble plugs are identical for IP2 and IP3 and are considered too small to have any observable effect on SFP reactivity.

The IP2 SFP criticality analysis explicitly models the reactivity effect of operation with an RCCA inserted to the bite position for the duration of two operating cycles. For IP2, the bite position is defined as the point of insertion which first provides a differential rod worth of 2.OE-5 Akeff per step. Depending upon the fuel cycle, the bite position varies from about 207-210 steps at BOL to 217-219 steps at EOL, versus a fuel pellet height of 221 steps.

Both IP2 and IP3 have operated at full licensed power with RCCAs at the bite or the fully withdrawn position. IP3 operated with RCCAs at the bite position at full power only during the first cycle of operation, and most operation in Cycle 1 was at or below 92%

rated thermal power. Therefore, provided that IP3 fuel is neutronically equivalent to IP2 fuel, then operation of IP3 fuel under RCCAs is bounded by the IP2 analysis.

Westinghouse has qualitatively determined in its reactivity report (Enclosure 1) that the IP3 fuel is neutronically equivalent to the IP2 fuel. (See response to RAI 14d). This conclusion applies to all IP3 fuel types (LOPAR, OFA, and the series of Vantage models) that operated in the IP3 core and were discharged from Cycles 1 through 11.

Therefore, since the IP2 fuel has been explicitly evaluated in the SFP criticality analysis for the effects of multi-cycle RCCA insertion operation, then it can be concluded that the IP3 fuel, which was operated with far less of an RCCA bite than IP2, will exhibit an axial burnup profile no more limiting than that of the existing IP2 model.

The one insert assembly that has been resident in the IP3 core but not in IP2 is the Hafnium Flux Suppressor. The HFS is a device that is used exclusively at the corners of the core to shield the interior reactor vessel wall from neutron fluence. Since HFSs have significant neutron absorbing capability, and since they are operated, like other removable neutron absorbers (i.e., in the fully inserted position throughout core life),

Entergy treats these devices in a similar fashion as RCCAs for wet transfer. For storage in the IP2 SFP, Westinghouse (Enclosure 1) has explicitly analyzed the effects of HFS on resident fuel assemblies and has confirmed that any reactivity penalty resulting from the presence of a HFS in a fuel assembly is more than compensated for by the excess reactivity holddown in IP2 SFP Regions 1-1 and 1-2. In fact for IP3 fuel assemblies discharged from Cycles 1 through 11, the HFS reactivity bias was determined to be

Response to Request for Additional Information NL-12-007 Attachment 1 Page 6 of 11 0.02200 Akff which is less than the limiting 0.02270 Akeff bias associated with Pyrex burnable absorbers discussed in response to 14a.

In summary, when considering fuel assembly inserts, in order to maintain the validity of the IP2 spent fuel pool criticality analysis with IP3 fuel in the pool the IP3 fuel assemblies shall be stored in accordance with the requirements presented in response to 14a.

c. Core Exit Temperature and the IP2 SFP Criticality Analysis As reported in Holtec Report HI-2094289 a moderator temperature of 6240 F was used in the depletion analysis. This temperature is based on the standards applied in NUREG/CR-06665 / ORNL/TM-1 999/303, "Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel". This NUREG is the source of the analyzed temperatures of 1000 K for fuel and 600 K for the moderator, and these criteria apply equally to both IP2 and IP3 fuel. As noted in the RAI, the 600 K moderator limitation translates to 6240 F.

The Cycle 1 FSAR temperature of 637.3 OF identified in the RAI refers to a design basis limiting temperature used for thermal conditions prior to the initiation of an accident that could challenge DNBR limits. An explicit analysis by Westinghouse of the first 11 cycles of IP3 operation (Enclosure 2), confirmed that the IP3 assembly exit temperature has never exceeded 624 OF.

RCS flow measurement for Cycle 1 was made using measured pump power correlated with elbow tap delta-P. After that, RCS flow measurement was not an official part of reload startup testing until Cycle 4. The flow for Cycle 3 was the minimum of four measurements made during initial power ascent at 70, 80, 90 and 100% power, using the same methodology (reverse calorimetric) that has been used since then. The four flow measurements agreed to within 1.4%, and the calculation included an inherent 4.2% penalty on the results. Therefore, it is concluded that the RCS flow for Cycle 2 can be estimated to be the same as the minimum measured Cycle 3 flow for the purposes of this calculation.

This analysis therefore verifies the applicability of 624 OF as a conservative temperature for IP3 fuel reactivity depletion analysis. The calculation results for the first eleven cycles of operation are included in Enclosure 2.

It should be noted that this temperature is conservative for use in fuel depletion analysis, since no fuel assembly would ever be in the limiting hot channel for its entire operating lifetime.

In summary, when considering fuel assembly exit temperature, in order to maintain the validity of the IP2 spent fuel pool criticality analysis with IP3 fuel in the pool the IP3 fuel assemblies must have had an exit temperature that has never exceeded 624 OF. It has been confirmed by analysis that the IP3 fuel assembly exit temperatures did not exceed 624 OF during Cycles 1 through 11.

Section 4.8 of Holtec Report HI-2094289 will be revised to incorporate the results of the analysis described above.

Response to Request for Additional Information NL-12-007 Attachment 1 Page 7 of 11

d. Axially Distributed Burnup Profile and the IP2 SEP Criticality Analysis The IP2 SFP criticality analysis uses a 24-node model axial power / burnup distribution model, based on results of the CASMO-4 and SAS2H codes. As noted in the response to RAI 14c, the analysis assumes a limiting power profile based on operation of a fuel assembly in the bite position over two cycles. This model, which was developed for the 2001 IP2 SFP criticality analysis, is less refined than the 18-node and 28-node models used by Holtec for characterization of the IP3 spent fuel in the Shielded Transfer Canister. Furthermore, a direct comparison of the two models is not appropriate, because the STC profile is a very conservative synthesized profile (as discussed in Chapter 4 of the Licensing Report), whereas the 1P2 SFP criticality analysis uses an actual bounding burnup profile. Hence, use of the STC profile would not be suitable for direct comparison to the existing IP2 SFP profile.

In order to provide assurance that the axially distributed burnup profile used in the IP2 SFP criticality analysis bounds the IP3 fuel discharged from Cycles 1 through 11 Enclosure 1 compares the IP2 and IP3 fuel assemblies neutronically.

Westinghouse has qualitatively determined in its reactivity report (Enclosurel) that the IP3 fuel is neutronically equivalent to the IP2 fuel. All IP3 fuel types under consideration for wet transfer (LOPAR, OFA, and the Vantage series) are similarly represented in the 1P2 SFP. Furthermore, the designs of all IP3 fuel models are identical to those of IP2, with the following mechanical exception: the IP3 fuel design includes three fewer intermediate grids than the IP2 equivalent models. This is insignificant from a reactivity contribution perspective, as noted in Enclosure 1. The existing IP2 SFP criticality analysis does not draw a distinction in the number of grids on a fuel assembly when characterizing it for storage. See Table 1 for a listing of the parameters important to a criticality analysis that were considered for the Enclosurel analysis.

This, in combination with the explicit evaluation of reactivity penalties due to spectrum hardening identified in Reference 1, and in light of the significant reactivity holddown capability of Regions 1-1 and 1-2, ensures that storage of IP3 fuel, when selected in accordance with the specifics outlined in the response to RAI 14f, is suitable for unrestricted storage in these high reactivity regions of the IP2 SFP.

It should be further noted that the enrichment, burnup and decay times of the selected IP3 fuel assemblies would more typically qualify them for storage in Regions 2-1 and 2-2, were they to have been discharged from the IP2 core. However, until such time as a formal criticality re-analysis of the IP2 SFP can be performed, the Westinghouse analysis provides ample assurance that any uncertainties or reactivity penalties resulting from the application of modern analysis methods will be compensated by the reactivity holddown capability of the Region 1-1 and 1-2 racks.

In summary, when considering the axially distributed burnup profile, in order to maintain the validity of the IP2 spent fuel pool criticality analysis with IP3 fuel in the pool, the IP3 fuel assemblies shall be stored in accordance with the requirements presented in the response to 14a.

Response to Request for Additional Information NL-12-007 Attachment 1 Page 8 of 11

e. IP3 SNF and the IP2 SFP Criticality Analysis The topics covered in the existing IP2 SFP criticality analysis are summarized in Table 2.

These topics have been addressed in the responses to the previous questions based on:

• Equivalence of IP2 and IP3 fuel for criticality analysis purposes

• Explicit calculations by Westinghouse of the effects of burnable absorber, control rods and hafnium flux suppressors

• Technical Specification changes identified above and summarized in the response to RAI 14f that restrict the IP3 fuel to that discharged from Cycles 1 through 11, limit initial enrichments, and require placement of IP3 fuel assemblies in the high reactivity regions in the IP2 SFP.

Note that the design basis parameters for IP2 fuel and IP3 fuel were compared by Westinghouse and are shown on Table 1 and in Enclosure 1. This includes manufacturing dimensions and materials, cladding material, enrichment limits, burnup limits, physical configuration, effects of axial blankets (six-inch and eight-inch), pellet density and effects of burnable absorbers. Westinghouse concludes, as noted in the response to RAI 14d, that the IP3 fuel is neutronically equivalent to the IP2 fuel, and therefore the IP3 spent fuel is suitable for storage in the IP2 SFP with appropriate restrictions.

Because the IP3 fuel is neutronically equivalent to the IP2 fuel, design basis accidents identified in the IP2 SFP criticality analysis (dropped fuel assembly, misloaded fuel assembly including outside of SFP rack, abnormal heat load) are valid for both IP2 fuel and IP3 fuel in the IP2 SFP, allowing for the TS changes noted above and summarized in the response to RAI 14f.

f. Additional Measures Based on the conclusions drawn in Sections 14a through 14e, the following restrictions have been determined to be appropriate to ensure that IP3 SNF can be safely stored in the IP2 SFP:

IP3 fuel assemblies shall be stored in Region 1-1 or Region 1-2 of the Spent Fuel Pit and:

a. The fuel assembly initial enrichment > 3.2 and < 4.4 w/o U2-, and

b. The fuel assembly discharge Cycle > 1 and s 11.

These restrictions will be incorporated into a new proposed revision to IP2 TS LCO 3.7.13 Spent Fuel Pit Storage. The TS Bases will be revised accordingly.

Response to Request for Additional Information NL-12-007 Attachment 1 Page 9 of 11 Table 1 Comparison of Fuel Features and Operating Conditions of Indian Point Fuel Parameter Unit 2 Unit 3 Cycles 1-11 Manufacturing Parameters Blanket Enrich (w/o) 0.74, 2.6, 3.2,-3.4, and 0.74, 2.6, and Fully Fully Enriched Enriched Blanket Lengths (inch) 0,6,8 0,6 Theoretical Density (%) 94.5 - 95.5 94.5 - 95.5 Clad Outer Diameter (inch) 0.422 0.422 Pellet Diameter (inch) 0.3659 0.3659 Depletion Parameters Power (MWt) 2758-3216 3025 Maximum Cycle Average Soluble Boron 870 800 Concentration (ppm)

Maximum Number of Pyrex Rods Used 20 20 Maximum Number of WABA Rodlets 20 20 Maximum Number of IFBA Rods 148 80

Response to Request for Additional Information NL-12r007 Attachment 1 Page 10 of 11 Table 2 Fuel Related Parameters/Characteristics Incorporated Into the STC Criticality Analysis Burnable poison impact on reactivity (All types)

IFBA loading Radial Burnup profile Soluble Boron Credit for accident conditions Neutron absorber material for storage of fuel Eccentric fuel assembly positioning in storage location Core Operating Parameters for Burnup Credit Specific power Moderator temperature Fuel temperature Soluble boron during depletion Axial burnup profile No boron credit during normal operations Dimensional analysis of storage location (STC cell inner dimension, wall thickness, cell pitch, etc.)

Dimensional analysis of fuel model used Manufacturing tolerances of STC included B-10 density in STC panels included Fuel Related Parameters/Characteristics Incorporated Into the IP2 SFP Criticality Analysis Credits soluble boron in SFP Axial burnup effect on reactivity (+0.02945 Akeff)

IFBA credit for high enrichment fuel (new)

Reactivity allowance for "spectrum hardening" from presence of WABA (+0.01 Akff)

Tolerance included for cell pitch, cell wall thickness, U02 density, U-235 enrichment,, and asymmetric assembly position in rack Dimensional analysis of fuel model used Reactivity effects of boraflex panel degradation (i.e., gap formation, local and uniform dissolution)

B-10 density of Boraflex panels incorporated Manufacturing tolerances of SFP racks included Core Operating Parameters for Burnup Credit Moderator temperature Fuel temperature Soluble boron during depletion Region 1-1, 2-1, and 2-2 minimum burnup curves increased by 4% for uncertainty in calculated burnup

Response to Request for Additional Information NL-12-007 Attachment 1 Page 11 of 11 Reference

1. Entergy Letter NL-01-110, 9/20/01, "License Amendment Request (LAR 01-010) for Spent Fuel Storage Pit Rack Criticality Analysis with Soluble Boron Credit'.

(ML012680336)

ENCLOSURE 1 TO NL-12-007 ATTACHMENTS 1, 2 and 3 to WESTINGHOUSE LETTER NF-IN-12-4 (NON PROPRIETARY)

Entergy Nuclear Operations, Inc.

Indian Point Units 2 and 3 Docket Nos. 50-247 and 50-286

OWestinghouse A TTA CHMENT 1 TO NF-IN-12-4 A TTA CHMENT ] TO CE 94:

REA CTIVITY EFFECTSOF BURNABLE ABSORBER USA GE AND EXCESS B URNUP ON SPENT FUEL POOL CRITICALITY MAR GIN (NON-PROPRIETARY)

(4 PAGES)

©2012 Westinghouse Electric Company LLC All Rights Reserved

Westinghouse Non-Proprietary Class 3 CE-12-94 February 9, 2012 Attachment 1 Reactivity Effects of Burnable Absorber Usage and Excess Burnup on Spent Fuel Pool Criticality Margin (Non-Proprietary) 3 Pages Attached

©2012 Westinghouse Electric Company LLC. All rights reserved.

Westinghouse Non-Proprietary Class 3 Attachment I to CE-12-94 Page 1 of 3 February 9, 2012 Effect of Burnable Absorbers during Irradiation References 1 and 2 provide guidance on the effect of modeling integral and discrete burnable absorbers on reactivity. These documents conclude that the maximum amount of discrete absorbers and Integral Fuel Burnable Absorbers (IFBA) should be included in the depletion models used for burnup credit analysis. The absorber impact on discharge reactivity is dependent on three key parameters that harden the neutron spectrum: the amount of burnup accrued with absorbers present, the absorber loading, and the enrichment of the assembly. Increased absorber loading and increased burnup accrued with the absorbers present result in an increase in discharge reactivity. The effect is inversely related to the enrichment (i.e., the reactivity impact of the absorbers is greater with lower enrichments).

The assemblies scheduled for inter-unit transfer entered operation at Indian Point Unit 3 (IP3) during Cycles 1 through 11. For added conservatism, all IP3 fuel that was first loaded in Cycles 1 through 11 was evaluated for this analysis. Reviewing the fuel management of these IP3 cycles, the maximum number of Pyrex, Wet Annular Burnable Absorbers (WABA), IFBA and IFBA/WABA combinations used in any assembly were identified. The reactivity effect of having only IFBA or WABA in an assembly is bounded by the effect of having IFBA and WABA in the same assembly during operation. This analysis considers the highest number of WABA rodlets and the highest number of IFBA rods that operated together. Additionally, hafnium absorber inserts were used in the final cycle of operation for some assemblies that contained WABA/IFBA during their earlier operation. The burnup values listed in Table 1 are the maximum values obtained by any assembly which contained that absorber type.

The criticality safety analysis for Indian Point Unit 2 spent fuel pool, contained in Reference 3, determined a reactivity bias of 0.01 Akeff for the depletion effects of burnable absorbers. The bias was calculated by determining the difference between the keff of an infinite array of assemblies depleted with and without burnable absorbers. In order to determine the impact of burnable absorbers on depletion for the Indian Point Unit 3 fuel, a parametric study was performed using the burnup values contained in Table 1. In the case of assemblies with WABA/IFBA and hafnium inserts, the assemblies were burned to the maximum burnup any assembly received with WABA present, the WABA was then replaced with a hafnium insert, and subsequently burned for the maximum duration any assembly contained a hafnium insert. All Calculations were performed with the minimum enrichment to be transferred (3.2 wt% 235U). The results of this parametric study are also presented in Table 1 and show the reactivity bias associated with burnable absorbers for Indian Point Unit 3 fuel.

Westinghouse Non-Proprietary Class 3 Attachment I to CE-12-94 Page 2 of 3 February 9, 2012 Table 1 - Limiting Burnable Absorber Data for All Fuel Assemblies Entering Operation at 1P3 during Cycles 1-11 Maximum Burnup Maximum Accrued while Burnable Absorber Type Number of Removable Absorber Reactivity Bias (Akeff)

Rodlets Type Present (MWd/MTU).

Pyrex 20 29,624 0.02270 WABA/IFBA 20/80 33,053 Hafnium 20 6,146 ,, __ .02_00 As shown in Table 1, the maximum reactivity bias due to the presence of burnable absorbers was 0.02270 Akeff. Since the previous analysis in Reference 3 already included a reactivity bias.of 0.01 Akeff in, it is necessary to offset an additional bias of 0.01270 Akeff.

Effect of Actual Fuel Assembly Burnups on Reactivity Entergy is planning to store the fuel being transferred from IP3 in Regions 1-1 and.1-2 of the. Indian Point Unit 2 spent fuel pool. All fuel assemblies from Cycles 1-11 have burnups that are well in excess of the current technical specification burnup limits for Region 1-1. Region 1-2 allows for fresh-fuel storage, and inherently has more margin relative to Region 1-1, therefore Region 1-1 is the only storage array considered in this analysis.

To determine the minimum amount of reactivity margin available from the-storage of fuel in Region 1-1, the minimum observed burnups were identified for each of the unique nominal enrichments of the fuel assemblies first loaded at IP3 in Cycles 1-11. The reactivity of these burnup and enrichment combinations were compared to the reactivity associated with the burnups required by the Region 1-1 Technical Specifications. The reactivity differences associated with the excess bumups of the, Indian Point Unit 3 Cycles 1 through 11 fuel, as compared with the Unit 2 Region 1-1 bumup limits, are shown in Table 2.

Westinghouse Non-Proprietary Class 3 Attachment I to CE- 12-94 Page 3 of 3 February 9, 2012 Table 2 - Reactivity Margin Generated By Excess Burnup 235 Enrichment (wt.% U) Margin to Burnup Limit Reactivity Margin (Akeff)

(MWd/MTU) 3.2 10,012 0.03497 3.3 12,698 0.04399 3.4 12,511 0.04180 3.6 15,176 0.04860 3.8 12,958 0.04198 4.0 17,359 0.05218 4.2 10,769 0.03086 4.3 20,274 0.06193 4.4 21,161 0.05999 4.5 16,359 0.04578 4.6 19,680 0.05494 4.95 10,537 0.02873 The minimum amount of reactivity margin shown due to depletion beyond the Region 1-1 burnup limits is 0.02873 Akff. It is also noted that while cooling time credit is not taken for Region 1-1, all assemblies to be transferred have at least 11 years of post irradiation cooling, creating an additional layer of reactivity margin.

References 1.) C. E. Sanders and J. C. Wagner, "Study of the Effect of Integral Burnable Absorbers for PWR Burnup Credit," NUREG/CR-6760 (ORNL/TM-2000/321), U.S. Nuclear Regulatory Commission, Washington, DC, March 2002.

2.) J. C. Wagner and C. V. Parks, "Parametric Study of the Effect of Burnable Poison Rods for PWR Burnup Credit," NUREG/CR-6761 (ORNL/TM-2000/373), U.S. Nuclear Regulatory Commission, Washington, DC, March 2002.

3.) Letter From F. Dacimo to the U.S. Nuclear Regulatory Commission, "License Amendment Request (LAR-01-010) for Spent Fuel Storage Pit Rack Criticality Analysis with Soluble Boron Credit," September 20, 2001, ADAMS Accession #ML012680336

(oWestinghouse A TTA CHMENT 2 TO NF-IN-12-4 A TTA CHMENT 2 TO CE-12-94:

SIMILARITY OFFUEL TO BE TRANSFERRED FROM INDIAN POINT UNIT 3 TO UNIT 2 SPENT FUEL POOL INVENTORY (NON-PROPRIETARY)

(3 PAGEs)

©2012 Westinghouse Electric Company LLC All Rights Reserved

Westinghouse Non-Proprietary Class 3 CE- 12-94 February 9, 2012 Attachment 2 Similarity of Fuel to Be Transferred from Indian Point Unit 3 to Unit 2 Spent Fuel Pool Inventory (Non-Proprietary) 2 Pages Attached

©2012 Westinghouse Electric Company LLC. All rights reserved.

Westinghouse Non-Proprietary Class 3 Attachment 2 to CE-1 2-94 Page 1 of 2 February 9, 2012 Similarity of Fuel to Be Transferred from Indian Point Unit 3 to Unit 2 Spent Fuel Pool Inventory To show that the Unit 3 fuel assemblies operated in Cycles 1-11 are neutronically equivalent to the Unit 2 spent fuel, a qualitative comparison of the parameters important to criticality is performed. The parameters important to criticality are divided into two broad groups, manufacturing parameters and depletion parameters, each of these groups is discussed below in relation to the fuel stored at Indian Point. Table 1 contains a listing of the important parameters and the associated values for both Unit 2 and Unit 3 fuel.

Table 1- Comparison of Fuel Features and Operating Conditions of Indian Point Fuel Parameter Unit 2 Unit 3 Cycles 1-11 Manufacturing Parameters Blanket Enrich (w/o) 0.74, 2.6, 3.2, 3.4, and 0.74, 2.6, and Fully Enriched Fully Enriched Blanket Lengths (inch) 0,6,8 0,6 Theoretical Density (%) 94.5 -95.5 94.5 -95.5 Clad Outer Diameter (inch) 0.422 0.422 Pellet Diameter (inch) 0.3659 0.3659 Depletion Parameters Power (MWt) 2758-3216 3025 Maximum Cycle Average Soluble Boron 870 S00 Concentration (ppm) 870__ 00 Maximum Number of Pyrex Rods Used 20 20 Maximum Number of WABA Rodlets 20 20 Maximum Number of IFBA Rods 148 80 With respect to manufacturing, the fuel in use at both Indian Point units is a Westinghouse 15x15 fuel assembly. All variations of the Westinghouse 15x15 assembly have the same pellet and cladding diameter, meaning that the amount of fuel and water in the lattice are identical to the design basis assembly. Over the years ZIRLO cladding has been implemented to replace earlier versions of Zircaloy.

ZIRLO cladding has minor alloying agents added to improve the mechanical performance of the cladding. The minor constituents of the cladding are mild absorbers and are present in such small quantities as to not have a significant impact on reactivity. Fuel theoretical density has increased by approximately 1.0 % since operation began at Indian Point, however, the highest theoretical density fabricated is in use at both units and remains bounded by the Unit 2 criticality analysis. There have been a number of changes to the grid designs, which are generally accompanied by a change in the name of the fuel product (HIPAR, LOPAR, OFA, etc.). Although these designs resulted in changes to the structural components of the assembly, these are of no consequence to the neutronic equivalence of the assembly, because those components are conservatively ignored in the criticality analysis.

ZIRLO is a registered trademark of Westinghouse Electric Company LLC, its Affiliates andfor its Subsidiaries in the United States of America and may be registered in other countries throughout the world. All rights reserved. Unauthorized use is strictly prohibited. Other names may be trademarks of their respective owners.

Westinghouse Non-Proprietary Class 3 Attachment 2 to CE-12-94 Page 2 of 2 February 9,2012 With respect to fuel depletion, the parameters which cause the neutron energy spectrum to harden increase the discharge reactivity of the fuel. Important parameters that impact the neutron energy spectrum are soluble boron concentration, burnable absorber usage, and moderator temperature. All of these parameters have generally increased over the years as more precise fuel management is needed to support higher power levels and longer cycles. A review of fuel management at both Indian PoInt units showed that the early Unit 3 fuel (Cycles 1-11) was operated at lower soluble boron concentrations and equivalent bounding burnable absorber loadings when compared to the fuel already resident in the Unit 2 pool. The analysis done to support RAI 14c showed all fuel in Cycles 1-11 was bounded by the 624 OF exit temperature.

As outlined above, the fuel at Unit 2 and Unit 3 was manufactured with the same dimensions important to criticality and irradiated under similar conditions. Based on this review, it is accurate to say that the fuel from Indian Point Unit 3 Cycles 1-11 is neutronically equivalent to the fuel already resident in the Indian Point Unit 2 pool.

• Westinghouse A TTA CHMENT 3 TO NF-IN-12-4 ATTACHMENT 3 TO CE-12-94:

EFFECTOFRADIAL BURNUP GRADIENTS ON SPENT FUEL POOL CRITICALITY SAFETY (NON-PROPRIETARY)

(2 PAGES)

©2012 Westinghouse Electric Company LLC All Rights Reserved

Westinghouse Non-Proprietary Class 3 CE-12-94 February 9, 2012 Attachment 3 Effect of Radial Burnup Gradients on Spent Fuel Pool Criticality Safety (Non-Proprietary) 1 Page Attached

©2012Westinghouse Electric Company LLC. All rights reserved.

Westinghouse Non-Proprietary Class 3 Attachment 3 to CE-12-94 Page I ofI Febrmry 9,2012 Effect of Radial Burnup Gradient on Spent Fuel Pool Criticality Safety Radial variations in assembly burnup do exist, however for several reasons the overall impact is small and it does not degrade subcritical margin in the spent fuel pool. Margin in the pool is not reduced because fuel management practices intentionally minimize burnup gradients on assemblies, it would take an extraordinary collocation of fuel to induce a significant effect, and the vast majority of assemblies are burned significantly in excess of the established burnup limits.

Radial variations in the power distribution due to fuel loading and leakage at the edge of the core result in a non-uniform horizontal core burnup distribution. As the reactor operates, the radial variations in power flatten due to fuel depletion and fission product poisoning throughout the interior of the core.

However, it is possible for fuel rods on one side of a peripheral assembly to experience notably less burnup than fuel rods on the opposite side of the same assembly because of high neutron leakage on the core periphery. Modern fuel management aims to limit this effect due to fuel and core performance issues such as peak rod burnup limits and quadrant power tilt mitigation strategies. To comply with these fuel management goals assemblies are typically shuffled across the core between operating cycles. Cross core shuffling of fuel prevents extreme radial burnup gradients from forming.

Although there are active attempts to manage the burnup gradients on fuel small gradients developed by assemblies in peripheral core locations are unavoidable. A study was performed in Reference 1 which estimated 0.009 Akeff as a bounding reactivity change due to radial bumup gradients for a cask design. The bounding value was based on the worst case loading of a shipping canister with a bounding radial gradient. This penalty is inflated because all of the under-burned portions of the assemblies were assumed to be rotated toward one another and all assemblies were assumed to have the worst case burnup gradient. In the spent fuel pool the collocation and optimum rotation of the assemblies with large radial gradients is highly unlikely. Additionally, most assemblies are burned to well in excess of the burnup limits providing additional reactivity margin.

Radial burnup gradients do not represent a significant reactivity concern for spent fuel pools because the magnitude of the gradients are limited by the in-core fuel management, the optimum location of fuel assemblies is highly unlikely and there is a large amount of uncredited burnup on most assemblies being stored in the pool.

References:

1.) J. M. Scaglione and J. C. Wagner, Review of Yucca Mountain Disposal Criticality Studies, Oak Ridge National Laboratory.

ENCLOSURE 3 TO NL-12-007 ATTACHMENT 5 to WESTINGHOUSE LETTER NF-IN-12-4 Application for Withholding Proprietary Information from Public Disclosure Entergy Nuclear Operations, Inc.

Indian Point Units 2 and 3 Docket Nos. 50-247 and 50-286

( Westinghouse A TTA CHMENT 5 TO NF-IN-12-4 CAW-12-3389:

APPLICATION FOR WITHHOLDING PROPRIETAR YINFORMA TION FROM PUBLICDISCLOSURE (8 PAGES)

This document is the property of and contains Proprietary Information owned by Westinghouse Electric Company LLC and/or its subcontractors and suppliers. It is transmitted to you in confidence and trust, and you agree to treat this document in strict accordance with the terms and conditions of the agreement under which it was provided to you.

©2012 Westinghouse Electric Company-LLC AllRights Reserved

()Westinghouse166 O esingh useWestinghouse

~

Electric Company Nuclear Services 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 LUSA U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4643 Document Control Desk Direct fax: (724) 720-0754 11555 Rockville Pike e-mail: greshaja@westingiouse.com Rockville, MD 20852 Proj letter:. NF-IN-12-4 CAW-12-3389 February 8, 2012, APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

CE-12-86, "Thermal-Hydraulic Design Results for the Indian Point Unit 3 Wet Fuel Transfer Reactivity Analysis" (Proprietary)

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-12-3389 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.

The subject document was prepared and classified as Westinghouse Proprietary Class 2. Westinghouse requests that the document be considered proprietary in its entirety. As such, a non-proprietary version will not be issued.

Accordingly, this letter authorizes the utilization of the accompanying affidavit by Entergy Nuclear Northeast Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-12-3389, and should be addressed to J. A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, Suite 428, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.

Very truly yours,

/J.A. Gresham, Manager Regulatory Compliance Enclosures

CAW-12-3389 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF BUTLER:

Before me, the undersigned authority, personally appeared J. A. Gresham, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

A. Gresham, Manager Regulatory Compliance Sworn to and subscribed before me this 8th day of February 2012

'NotayPbi COMMONWEALTH OF PENNSYLVANIA Notarial Seal Cynthia Olesky, Notary Public Man" Boro, Westmoreland County n Commission )le*sJuly 16, 2014 Member. Pennsvlvanla Assodation of Notaries

2 CAW-12-3389 (1) 1 am Manager, Regulatory Compliance, in Nuclear Services, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

(2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of

3 CAW-12-3389 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c) Its use 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 a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

There are sound policy reasons behind the Westinghouse system which include the following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

(b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

4 CAW- 12-3389 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a

.competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iii) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission.

(iv) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

(v) The proprietary information sought to be withheld in this submittal is that which is contained in CE- 12-86, "Thermal-Hydraulic Design Results for the Indian Point Unit 3 Wet Fuel Transfer Reactivity Analysis" (Proprietary), dated February 6, 2012, for submittal to the Commission, being transmitted by Entergy Nuclear Northeast letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse is that associated with Request for Additional Information Regarding the Inter-unit Spent Fuel Transfer License Amendment Request (TAC numbers NlI 671, ME 1672, L24299),

and may be used only for that purpose.

5 CAW-12-3389 This information is part of that which will enable Westinghouse to:

(a) Provide Thermal-Hydraulic design analysis results.

Further this information has substantial commercial value as follows:

(a) Westinghouse plans to sell the use of similar information to its customers for the purpose of Thermal-Hydraulic design analysis results.

(b) Westinghouse can sell support and defense of Thermal-Hydraulic design analysis results.

(c) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar Thermal-Hydraulic design analysis results and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

PROPRIETARY INFORMATION NOTICE Transmitted herewith is the proprietary version of a document furnished to the NRC in connection with requests for generic and/or plant-specific review and approval. The document is to be considered proprietary in its entirety.

COPYRIGHT NOTICE The report transmitted herewith bears a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in this report which is necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.