Text

License Amendment Request for Spent Fuel Pool Region I Criticality
	ML110380083
Person / Time
Site:	Palisades
Issue date:	01/31/2011
From:	Kirwin T; Entergy Nuclear Operations
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	PNP 2011-002
	Download: ML110380083 (58)
	v • d • e

Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043 Tel 269 764 2000 Thomas P Kirwin Acting Site Vice President PNP 2011-002 January 31, 2011 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

SUBJECT:

License Amendment Request for Spent Fuel Pool Region I Criticality Palisades Nuclear Plant Docket 50-255 License No. DPR-20

Dear Sir or Madam:

Pursuant to 10 CFR 50.90, Entergy Nuclear Operations, Inc. (ENO) requests Nuclear Regulatory Commission (NRC) review and approval of a proposed license amendment to amend Renewed Facility Operating License DPR-20 for the Palisades Nuclear Plant (PNP). ENO proposes to revise Appendix A, Technical Specifications (TS), as they apply to the spent fuel pool (SFP) storage requirements in TS section 3.7.16 and criticality requirements for Region I SFP and north tilt pit fuel storage racks, in TS section 4.3.

The criticality analyses supporting the proposed TS change for the Region I fuel storage racks reflect credit for fuel assembly burnup and soluble boron. Based on the analyses, the proposed change, in accordance with 10 CFR 50.68, "Criticality accident requirements," would maintain the effective neutron multiplication factor (Keff) limits for Region I storage racks to less than 1.0 when flooded with unborated water, and less than, or equal to, 0.95 when flooded with water having a minimum boron concentration of 850 ppm during normal operations, and 1350 ppm during accident conditions.

This proposed change has been evaluated in accordance with 10 CFR 50.91 (a)(1) using criteria in 10 CFR 50.92(c), and it has been determined that this change involves no significant hazards consideration. The bases for this determination are included in , which provides a description of the proposed change, a background discussion, a technical analysis, a regulatory analysis, and an environmental review. provides the revised TS pages reflecting the proposed changes. provides the annotated TS pages showing the proposed changes. provides the AREVA NP Inc. proprietary authorization affidavit supporting the proprietary nature of Attachment 5. The affidavit sets forth the basis for which the

Document Control Desk Page 2 information may be withheld from public disclosure by the NRC and addresses the specific considerations listed in 10 CFR 2.390. contains the proprietary AREVA NP Inc. report, "Palisades SFP Region 1 Criticality Evaluation with Burnup Credit." ENO requests that Attachment 5 be withheld from public disclosure in accordance with 10 CFR 2.390. Correspondence regarding the proprietary aspects of the AREVA NP Inc. report or the supporting affidavit should reference the affidavit and be addressed to George L. Pannell, Manager, Product Licensing, AREVA NP, Inc., 3315 Old Forest Road, Lynchburg, VA 24506. contains the non-proprietary version of the AREVA NP Inc. report with the proprietary information deleted.

To support fuel pool operations necessary to accommodate a full reactor core off load during the 2012 refueling outage, ENO requests approval of the proposed license amendment request by February 7, 2012. The amendment will be implemented within 60 days of approval.

In accordance with 10 CFR 50.91, ENO is notifying the State of Michigan of this proposed license amendment by transmitting a copy of this letter and non-proprietary attachments to the designated state official.

Summary of Commitments This letter identifies no new commitments and no revisions to existing commitments.

I declare under penalty of perjury that the foregoing is true and correct. Executed on January 31, 2011.

Sincerely, tpk/jlk

Document Control Desk Page 3 Attachments: 1. Description and Evaluation of Requested Change

2. Revised Technical Specification Pages and Renewed Operating License Page Change Instructions

3. Mark-up of Technical Specification Pages

4. AREVA NP Inc. Affidavit

5. AREVA NP Inc. Technical Report, Document No.

ANP-2858P-003, Palisades SFP Region 1 Criticality Evaluation with Burnup Credit

6. AREVA NP Inc. Technical Report, Document No.

ANP-2858NP-003, Palisades SFP Region 1 Criticality Evaluation with Burnup Credit cc: Administrator, Region III, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC State of Michigan

ATTACHMENT 1 DESCRIPTION AND EVALUATION OF REQUESTED CHANGE

1.0 DESCRIPTION

Entergy Nuclear Operations, Inc. (ENO) requests amending the Renewed Facility Operating License DPR-20 for Palisades Nuclear Plant (PNP) to revise Appendix A, Technical Specifications (TS), fuel storage requirements as they apply to Region I storage racks in the PNP main spent fuel pool (SFP) and the north tilt pit portion of the SFP. The license amendment would revise the spent fuel assembly storage specification in TS 3.7.16 and the criticality section in TS 4.3 for Region I fuel storage racks. The analysis that supports the proposed change takes credit for fuel assembly burnup, soluble boron, a complete loss of boron in the Carborundum plates, and complete voiding of the gaps between the SFP rack individual storage cells (flux trap).

Based on the analyses, the proposed change, in accordance with 10 CFR 50.68, "Criticality accident requirements," would maintain the effective neutron multiplication factor (Keff) limits for Region I storage racks to less than 1.0 when flooded with unborated water, and less than, or equal to, 0.95 when flooded with water having a minimum boron concentration of 850 ppm during normal operations, and 1350 ppm during accident conditions.

The arrangement of the Region I and Region II storage racks in the SFP is shown in Figure B 3.7.16-1 of the Technical Specifications Bases. The storage racks are located in the main pool area and the north tilt pit area of the SFP.

2.0 PROPOSED CHANGE

ENO proposes to modify (1) the SFP storage requirements in TS 3.7.16, by revising a limiting condition for operation (LCO) for Region I fuel and non-fissile bearing component storage and by inserting tables containing spent fuel minimum burnup for Regions 1B, 1C, 1D, and 1 E, and (2) the Region I fuel storage criticality requirements, in the TS design features, in TS section 4.3, by describing revised requirements for Regions 1B and 1E and adding requirements for new Regions 1C and 1D. Generally, requirements in TS section 4.3 for Region 1A are not changed but are reformatted to align with the format of the proposed requirements for Regions 1B, 1C, 1 D, and 1 E.

However, new interface requirements for the regions are proposed.

The supporting analysis for Regions 1A, 1B, 1C, 1D, and 1E requirements in has resulted in proposing restrictions on fuel assemblies that are unique and do not allow the use of exact verbiage from NUREG-1 432, "Standard Technical Specifications - Combustion Engineering Plants." The content of the specifications adhere to NUREG-1432 to the extent possible.

TS page numbers in TS sections 3 and 4 are also changed due to the revised text.

Page 1 of 21

TS LCO 3.7.16 would be revised to add requirements for the maximum nominal Planar average U-235 enrichment and burnup for Region I Region 1B, 1C, 1 D, and 1 E fuel assemblies, and would read as follows:

"Storage in the spent fuel pool shall be as follows:

a. Each fuel assembly and non-fissile bearing component stored in Region I shall be within the limitations in Specification 4.3.1.1 and, as applicable, within the requirements of the maximum nominal planar average U-235 enrichment and burnup of Tables 3.7.16-2, 3.7.16-3, 3.7.16-4, or 3.7.16-5; and

b. The combination of maximum nominal planar average U-235 enrichment, burnup, and decay time of each fuel assembly stored in Region II shall be within the requirements of Table 3.7.16-1."

The proposed change would add restrictions for Region I in LCO item "3.7.16a." on fuel assembly maximum nominal planar average U-235 enrichment and burnup. The proposed change would clarify Region II LCO "3.7.16b." by replacing an initial enrichment requirement with a requirement on maximum nominal planar average U-235 enrichment. The format differs from NUREG-1432 due to the unique restrictions on fuel assembly storage.

The header in the left hand column in Table 3.7.16-1 would be revised from "Initial Enrichment (Wt%)" to "Nominal Planar Average U-235 Enrichment (Wt%)."

This proposed change is for consistency and clarity.

Tables 3.17.6-2 throuah 3.17.6-5 would be added as described below.

Page 2 of 21

New Table 3.7.16-2 contains spent fuel minimum burnup requirements for Region 1B, and would read as follows:

'Table 3.17.6-2 (page 1 of 1)

Spent Fuel Minimum Burnup Requirements for Storage in Region 1 B (three-of-four loading configuration) of the Main Spent Fuel Pool Nominal Planar Burnup Burnup Average U-235 (GWD/MTU) (GWD/MTU)

Enrichment (Batches L (Batches A (Wt%) and later) through K)

<2.10 0 1.0 2.40 4.1 5.1 2.60 6.7 7.7 2.80 9.5 10.5 3.00 12.2 13.2 3.20 14.9 15.9 3.40 17.6 18.6 3.60 20.2 21.2 3.80 23.0 24.0 4.00 25.7 26.7 4.20 28.4 29.4 4.40 31.1 32.1 4.54 33.0 34.0 (a) Linear interpolation between two consecutive points for nominal planar average U-235 enrichments between 2.10 and 4.54 will yield acceptable results.

(b) Comparison of nominal assembly average burnup numbers to these in the table is acceptable if measurement uncertainty is < 10%."

Page 3 of 21

New Table 3.7.16-3 contains spent fuel minimum burnup requirements for Reaion 1C. and would read as follows:

'Table 3.7.16-3 (page 1 of 1)

Spent Fuel Minimum Burnup Requirements for Storage in Region 1C (four-of-four loading configuration) of the Main Spent Fuel Pool Nominal Planar Burnup Burnup Average U-235 (GWD/MTU) (GWD/MTU)

Enrichment (Batches L (Batches A (Wt%) and later) through K)

<1.35 0 1.0 2.40 20.7 21.7 2.60 24.5 25.5 2.75 27.5 28.5 2.80 28.2 29.2 3.00 31.0 32.0 3.20 33.9 34.9 3.40 36.7 37.7 3.60 39.5 40.5 3.80 42.4 43.4 4.00 45.2 46.2 4.20 48.0 49.0 4.40 50.8 51.8 4.54 52.8 53.8 (a) Linear interpolation between two consecutive points for nominal planar average U-235 enrichments between 1.35 and 4.54 will yield acceptable results.

(b) Comparison of nominal assembly average burnup numbers to these in the table is acceptable if measurement uncertainty is < 10%."

Page 4 of 21

New Table 3.7.16-4 contains spent fuel minimum burnup requirements for Region 1D, and would read as follows:

'Table 3.7.16-4 (page 1 of 1)

Spent Fuel Minimum Burnup Requirements for Storage in Region 1 D (three-of-four loading configuration) of the North Tilt Pit Nominal Planar Burnup Burnup Average U-235 (GWD/MTU) (GWD/MTU)

Enrichment (Batches L (Batches A (Wt%) and later) through K)

<2.35 0 1.0 2.40 0.5 1.5 2.60 2.4 3.4 2.80 4.3 5.3 3.00 6.2 7.2 3.20 8.1 9.1 3.40 10.0 11.0 3.60 11.9 12.9 3.80 13.8 14.8 4.00 15.7 16.7 4.20 17.7 18.7 4.40 19.6 20.6 4.54 20.9 21.9 (a) Linear interpolation between two consecutive points for nominal planar average U-235 enrichments between 2.35 and 4.54 will yield acceptable results.

(b) Comparison of nominal assembly average burnup numbers to these in the table is acceptable if measurement uncertainty is _<10%."

Page 5 of 21

New Table 3.7.16-5 contains spent fuel minimum burnup requirements for Redion 1 E, and would read as follows:

'Table 3.7.16-5 (page 1 of 1)

Spent Fuel Minimum Burnup Requirements for Storage in Region 1 E (four-of-four loading configuration) of the North Tilt Pit Nominal Planar Burnup Burnup Average U-235 (GWD/MTU) (GWD/MTU)

Enrichment (Batches L (Batches A (Wt%) and later) through K)

<1.48 0 1.0 2.40 13.9 14.9 2.60 16.9 17.9 2.80 19.9 20.9 3.00 23.0 24.0 3.20 26.0 27.0 3.30 27.5 28.5 3.40 28.7 29.7 3.60 31.0 32.0 3.80 33.3 34.3 4.00 35.6 36.6 4.20 37.9 38.9 4.40 40.2 41.2 4.54 41.8 42.8 (a) Linear interpolation between two consecutive points for nominal planar average U-235 enrichments between 1.48 and 4.54 will yield acceptable results.

(b) Comparison of nominal assembly average burnup numbers to these in the table is acceptable if measurement uncertainty is < 10%."

Page 6 of 21

TS 4.3.1.1 would be revised to read as follows:

"The Region I fuel storage racks (see Figure B 3.7.16-1) incorporating Regions 1A, 1B, 1C, 1D, and 1E are designed and shall be maintained with:"

TS 4.3.1.1 contains requirements for Region I fuel storage racks. The proposed revision would add new Regions 1C and 1D within Region I. The requi.rements in Region 1A, 1B, 1C, 1D, and 1E are described below. The proposed changes within TS 4.3.1.1, including those below, would result in revision of the specification that differs from the wording in NUREG-1432.

TS 4.3.1.1 a. would be revised to read as follows:

"a. New or irradiated fuel assemblies having a maximum nominal planar average U-235 enrichment of 4.54 weight percent;"

The proposed change would reflect that the maximum nominal planar average U-235 enrichment for new or irradiated fuel assemblies is now 4.54 weight percent for Region 1A, 1 B, 1C, 1D, and 1E as determined in the analysis in Attachment 5. The maximum nominal planar average U-235 enrichment for Region 1A would be unchanged and would remain as 4.54 weight percent.

TS 4.3. 1.1 b. would be revised to read as follows:

Changed the font for "Keff."

TS 4.3.1.1d. would be revised to read as follows:

"d. Regions 1A, 1B, and 1C have a nominal 10.25 inch center to center distance between fuel assemblies;"

The proposed change would remove the description of the single Type E rack that would be inserted in the revised TS 4.3.1.le. below.

TS 4.3.1.1 e. would be revised to read as follows:

"e. Regions 1D and 1 E have a nominal 11.25 inch by 10.69 inch center to center distance between fuel assemblies;"

The proposed change removes "new or irradiated assemblies," which is included in the proposed TS 4.3.1.1 a, and inserts the description of Regions 1D and 1E in the single Type E rack from TS 4.3.1 .ld.

Page 7 of 21

TS 4.3.1 .lf. would be revised to read as follows:

"f. Region 1A is defined as a subregion of the Region I storage racks located in the main spent fuel pool and is subject to the following restrictions. Fuel assemblies (or fissile bearing components) located in Region 1A shall be in a maximum of two-of-four checkerboard loading pattern of two fuel assemblies (or fissile bearing components) and two empty cells. Designated empty cells may contain non-fuel bearing components in accordance with Section 4.3.1.1 m.2. below;"

The proposed change would reformat this specification to align with the proposed TS 4.3.1.1g., TS 4.3.1.1h., and TS 4.3.1.1i. No changes would be made to fuel assembly storage requirements in this specification.

TS 4.3.1.1.q. would be revised to read as follows:

"g. Region 1 B is defined as a subregion of the Region I storage racks located in the main spent fuel pool and is subject to the following restrictions. Fuel assemblies (or fissile bearing components) located in Region 1B shall be in a maximum of three-of-four loading pattern consisting of three fuel assemblies (or fissile bearing components) and one empty cell. Fuel assemblies in Region 1B shall meet the enrichment dependent burnup restrictions listed in Table 3.7.16-2.

Designated empty cells may contain non-fuel bearing components in accordance with Section 4.3.1.1 m.2. below;"

The proposed change reflects the analysis in Attachment 5 for Region 1B.

TS 4.3.1.lh. would be revised to read as follows:

"h. Region 1C is defined as a subregion of the Region I storage racks located in the main spent fuel pool and, is subject to the following restrictions. Fuel assemblies (or fissile bearing components) located in Region 1C may be in a maximum of four-of-four loading pattern with no required empty cells. Fuel assemblies in Region 1C shall meet the enrichment dependent burnup restrictions listed in Table 3.7.16-3;"

The proposed change reflects the analysis in Attachment 5. The existing interface requirements for the main spent fuel pool would be relocated to TS 4.3.1.1 i.

Interface requirements would be relocated from TS 4.3.1.1 h and revised. Changes to the interface requirements would reflect the analysis in Attachment 5 as related to Regions 1A, 1B, and 1C in the main pool in TS 4.3.1.1 i., and Regions 1D and 1E in the north tilt pit in TS 4.3.1.11.

Page 8 of 21

TS 4.3,1.1 i. would be revised to read as follows:

"i. Interface requirements for the main spent fuel pool between Region 1A, 1B, and 1C are as follows. Region 1A, 1B, and 1C can be distributed in Region I, in the main spent fuel pool, in any manner provided that any two-by-two grouping of storage cells and the assemblies in them correspond to the requirements of 4.3.1.1f., 4.3.1.1g., or 4.3.1.1h. above;"

The proposed change would provide interface requirements for Regions 1A, 1B, and 1C in the main spent fuel pool.

TS 4.3.1.1 i. would be revised to read as follows:

"j. Region 1 D is defined as a subregion of the Region I storage rack located in the north tilt pit and is subject to the following restrictions. Fuel assemblies (or fissile bearing components) located in Region 1D may be in a maximum of three-of-four loading pattern with no required empty cells. Fuel assemblies in Region 1 D shall meet the enrichment dependent burnup restrictions listed in Table 3.7.16-4;"

The proposed change would reflect the analysis in Attachment 5. Non-fissile bearing component requirements would be relocated to TS 4.3.1.1 m.

New TS 4.3.1.1k., which revises 4.3.1.1 i., would be added to read as follows:

"k. Region 1 E is defined as a subregion of the Region I storage rack located in the north tilt pit and is subject to the following restrictions. Fuel assemblies (or fissile bearing components) located in Region 1E may be in a maximum of four-of-four loading pattern with no required empty cells. Fuel assemblies in Region 1E shall meet the enrichment dependent burnup restrictions listed in Table 3.7.16-5;"

The proposed change would revise TS 4.3.1.li. and reflects the analysis in .

New TS 4.3.1.11. would be added to read as follows:

"1. Interface requirements for the north tilt pit between Region 1D and 1 E are as follows. Region 1D and 1E can be distributed in Region I in the north tilt pit in any manner provided that any two-by-two grouping of storage cells and the assemblies in them correspond to the requirements of 4.3.1. 1j. or 4.3.1.1 k.

above;"

The proposed change adds interface requirements for Regions 1 D and 1E in the north tilt pit.

Page 9 of 21

New TS 4.3.1.1 m., which revises TS 4.3.1.1 i., would be added to read as follows:

"im. Non-fissile bearing component restrictions are as follows:

1. Non-fissile material components may be stored in any designated fuel location in Region 1A, 1B, 1C, 1 D, or 1 E without restriction.

2. The following non-fuel bearing components (NFBC) may be stored face adjacent to fuel in any designated empty cell in Region 1A or 1B.

(i) The gauge dummy assembly and the lead dummy assembly may be stored face adjacent to fuel in any designated empty cells with no minimum required separation distance.

(ii) A component comprised primarily of stainless steel that displaces less than 30 square inches of water in any plane within the active fuel region may be stored in any designatedempty cell as long as the NFBC is at least ten locations away from any other NFBC that is in a designated empty cell, with the exception of 4.3.1.1 m.2.(i) above.

3. Control blades may be stored in both fueled and unfueled locations in Regions 1D and 1E, with no limitation on the number."

The proposed change would revise TS 4.3.1 .lj. for clarity.

The proposed change would remove from TS 4.3.1 .lj.2 the exception for interface locations described in TS 4.3.1.1 h. when storing non-fuel bearing components face adjacent to fuel in designated empty cells in Region 1A or 1B. This exception would be removed due to the analysis in Attachment 5.

The proposed change also would remove TS 4.3.1 .lj.2.b. because assemblies comprised of 216 solid stainless steel rods are not installed at Palisades.

Lastly, the proposed change would remove and revise TS 4.3.1 .lj.3., which restricted non-fissile bearing components from being stored in designated empty cells in Region 1E. This restriction would be removed because Region 1 E would no longer have designated empty cells. The revised section would provide clarification on the storage of control blades.

TS 4.3.1.2a. would be revised to read as follows:

"a. Fuel assemblies having maximum nominal planar average U-235 enrichment of 4.60 weight percent;"

For consistency with the proposed changes to TS 3.7.16, the proposed change to TS 3.1.2a. would insert "nominal" into the specification.

Page 10 of 21

TS 4.3.1.2c. would be revised as follows:

Changed the font for "Keff."

TS 4.3.1.2e. would be revised to read as follows:

"e. New or irradiated fuel assemblies which meet the maximum nominal planar average U-235 enrichment, burnup, and decay time requirements of Table 3.7.16-1."

For consistency with the proposed changes to TS 3.7.16, the proposed change to TS 4.3.1.2e. would replace "initial enrichment" with "maximum nominal planar average U-235 enrichment."

TS 4.3.1.3a. would be revised to read as follows:

"a. Twenty four unirradiated fuel assemblies having a maximum nominal planar average U-235 enrichment of 4.95 weight percent, and stored in accordance with the pattern shown in Figure 4.3-1, or Thirty six unirradiated fuel assemblies having a maximum nominal planar average U-235 enrichment of 4.05 weight percent, and stored in accordance with the pattern shown in Figure 4.3-1 ;"

For consistency with the proposed changes to TS 3.7.16, the proposed change to TS 4.3.1.3a would insert "nominal" into the specification.

TS 4.3.1.2, 4.3.1.3, 4.3.2, 4.3.3 and Figure 4.3-1 would be moved and pages would be repaginated.

3.0 BACKGROUND

In July 2008, ENO identified that results from Boron-1 0 Areal Density Gauge for Evaluating Racks (BADGER) testing, of the Region I SFP storage racks, indicated that the neutron absorber material contained less boron-10 than assumed in the SFP criticality analysis of record. At the time, the neutron absorber in the Region I SFP and north tilt pit storage racks, was relied on for compliance with TS 4.3.1 .lb criticality requirements. TS 4.3.1.1b required that Keff for Region I fuel racks be less than or equal to 0.95 if fully flooded with unborated water. Soluble boron would have been required to maintain a Keff less than or equal to 0.95 in the Region I fuel racks, therefore, PNP was no longer in compliance with the TS requirement or with 10 CFR 50.68 requirements. In accordance with NRC Administrative Letter 98-10, "Dispositioning'of Technical Specifications that are Insufficient to Assure Plant Safety,"

compensatory measures were implemented.

Page 11 of 21

The SFP contains storage racks that are designated as Region I and Region I1. The Region I storage racks contain Carborundum neutron absorber plates that is no longer credited in the Region I criticality calculations (refer to PNP license Amendment No. 236, Accession no. ML090160238). The Region II racks contain a neutron absorbing material, Boraflex, that is not credited in the Region II criticality calculations (refer to PNP license amendment no. 207, Accession nos. ML020590151 and ML020440048). Soluble boron, at 850 ppm, is required by TS to maintain Keff less than or equal to 0.95 in the Region I and Region IIstorage racks when fully flooded with water.

ENO submitted a letter dated August 27, 2008 (ML082410132), describing four commitments and plans for a license amendment request (LAR) to address the degraded SFP storage rack neutron absorber. Licensee Event Report 08-004, dated September 15, 2008 (ML082660584), described the noncompliance with TS 4.3.1.1 b.

and 10 CFR 50.68. The NRC issued a Confirmatory Action Letter (CAL) on September 18, 2008 (ML082630145), confirming commitments by ENO in the August 27, 2008, letter. The CAL also indicated a LAR submittal is needed to restore regulatory compliance prior to the next refueling outage.

ENO submitted a LAR dated November 25, 2008 (ML083360619 and ML083360624),

to restore regulatory compliance with TS 4.3.1.1b and 10 CFR 50.68. This LAR proposed to revise SFP storage requirements in TS section 3.7.16 and the criticality requirements for the Region I spent fuel pool and north tilt pit storage racks in TS section 4.3.1.1. The supporting criticality analysis credited soluble boron in the same manner and magnitude credited for the Region II fuel storage racks. The NRC subsequently approved the LAR by issuing Amendment No. 236 on February 6, 2009 (ML090160238).

On September 1, 2009, ENO submitted another LAR (ML092450683) to modify the TS SFP Region I requirements. Subsequent review by the NRC resulted in a request for supplemental information in a letter dated October 15, 2009 (ML092860363). ENO subsequently withdrew the LAR submittal, on October 29, 2009 (ML093030034), and proceeded with actions to address the NRC requested information. Follow-up meetings were held with the NRC staff. This submittal modifies the previous proposed LAR of September 1, 2009, and addresses the information requested by the NRC in the October 15, 2009, letter.

4.0 TECHNICAL ANALYSIS

Region I Criticality Evaluation AREVA NP Inc. report, Document No. ANP-2858P-003, "Palisades SFP Region 1 Criticality Evaluation with Burnup Credit," (Attachment 5) provides the technical analysis for the proposed change to store fuel up to a maximum nominal planar average 4.54 Page 12 of 21

weight percent U-235 in Region 1 of the SFP. ENO has reviewed and accepted this report.

Key elements of the report are as follows:

Analysis Model Conservatisms

No credit is taken for intermediate fuel assembly spacer grids or end fittings.

" No credit is taken for any absorber material in the Carborundum plates or moderator material in the gap region between the fuel assemblies.

" No credit is taken for the presence of residual gadolinia in the burned fuel assemblies.

" The conservative fuel enrichment tolerance of 0.05 weight percent is considered in the tolerance evaluation.

" The only moderator material is contained within the fuel assembly envelope, and above and below the SFP storage racks.

As stated in Table B-i, of ANP-2858P-003 (Attachment 5), the most reactive assembly configuration contains no integral gadolinium absorbers. This determination is based on sensitivity studies for a range of gadolinium concentrations and configurations over the exposure range of interest. This result reflects the importance of the gadolinium absorption relative to the increased actinide production in assemblies containing integral gadolinium absorbers. Therefore, it is conservative to assume no integral gadolinium absorbers are present.

Methodology The KENO-V.a computer code, a part of the SCALE4.4a package, was used in the computational analyses. Extensive benchmarking of KENO is described in Appendix A of the report (Attachment 5). The CASMO-3 computer code, a multi-group two dimensional transport theory program, was used to calculate isotopic compositions at specified burnups for the fuel assemblies.

Results The results of the analysis described in Attachment 5 determined that the Region I subregion 1A, 1B, 1C, 1D, and 1 E racks have a Keff of less than 1.0, with the racks loaded with a certain bounding nominal planar average enrichment, designated storage cells void of fuel, and racks flooded with unborated water at a temperature corresponding to the highest reactivity. The report demonstrated that Keff is less than or equal to 0.95 with the racks loaded with a certain bounding nominal planar average enrichment and designated storage cells void of fuel, and flooded with borated water at a temperature corresponding to the highest reactivity. Thus, compliance with 10 CFR 50.68 is maintained. Also, reactivity effects of abnormal and accident conditions (mis-loaded fuel) will not result in Keff exceeding the regulatory limit of 0.95 under borated conditions. The report analyzed the impact of the change of Region I on the Region II fuel storage racks and the resultant interfaces within and between the racks in each Region. The analysis does not affect other PNP fuel handling systems.

Page 13 of 21

Summary of Boron Dilution Evaluation On March 2, 2001, Consumers Energy (former owner and license holder) submitted an amendment request for the SFP boron concentration. The amendment request and supplements provided the basis for NRC issuance of Amendment No. 207 (February 26, 2002; ML020440048 & ML020590151) to the Palisades Operating License, allowing changes to enrichment limits in the SFP. The amendment request provided a SFP boron dilution evaluation. The evaluation has been reviewed and remains valid. In this evaluation, available dilution sources were compiled and analyzed against the calculated dilution volumes to determine the potential of a SFP boron dilution event.

For each dilution scenario, calculations were performed to define the dilution time for the SFP to reach 850 ppm.

The evaluation shows that a large volume of water (123,007 gallons) is necessary to dilute the SFP from the present TS limit of 1720 ppm to a soluble boron concentration where a Keff of 0.95 would be approached in the pool. For the limiting dilution source flow rate, the dilution time to reach a pool concentration of 850 ppm was determined to be 9.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The first 15,000 gallons of dilution water would fill the pool to its overflow level. The remaining 107,600 gallons needed to dilute the pool to 850 ppm would all flow over the pool deck and down the equipment hatch, the elevator shaft, or the stair well, all of which are located within four to ten feet of the pool.

The resulting water distribution throughout the auxiliary building and safeguards room basement would result in high sump level alarms in the control room. The large amount of water on the floor in these areas would be easily spotted by the operators whether they have specifically been sent there in response to an alarm or if they were making normal rounds through the auxiliary building and fuel pool on a shiftly basis. Therefore, it is reasonable to assume that the operators will recognize and terminate this event well before the boron concentration in the SFP would drop below 850 ppm at 9.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months into the event. A fuel pool high level alarm would give an even earlier warning of fuel pool level increases that could lead to dilution of the soluble boron concentration.

Abnormal Conditions and the Double-Contingency Principle NRC Memorandum from L. Kopp to T. Collins, dated August 19, 1998, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," includes the following summary of abnormal conditions and the double-contingency principle:

"The criticality safety analysis should consider all credible incidents and postulated accidents. However, by virtue of the double-contingency principle, two unlikely independent and concurrent incidents or postulated accidents are beyond the scope of the required analysis. The double-contingency principle means that a realistic condition may be assumed for the criticality analysis in calculating the effects of incidents or postulated accidents. For example, if soluble boron is normally present in the spent fuel pool water, the loss of soluble boron is considered as one accident condition and the second concurrent Page 14 of 21

accident need not be assumed. Therefore, credit for the presence of the soluble boron may be assumed in evaluating other accident conditions."

The proposed changes support compliance with this principle.

Human Performance The current process for moving fuel assemblies is controlled by system operating procedure SOP-28, "Fuel Handling System." This procedure provides the detailed steps associated with the equipment controls on the fuel handling machines, as well as the required communications necessary between the fuel handling machine operator and the fuel handling communicator (FHC).

Fuel movement plans are developed by experienced and qualified personnel.

Engineering Manual procedure EM-04-29, "Guidelines for Preparing Fuel Movement Plans," is the governing document for preparation of fuel move plans. The procedure requires an independent review by another qualified person and ensures that both the preparer and reviewer verify that the fuel move plan would result in approved storage patterns per Technical Specification 4.3.

A human performance work practice includes annotating unique conditions associated with specific fuel moves with clarifying notes in the comment field or elsewhere on the fuel move sheets. These notes establish additional process controls to minimize the probability of a fuel move error. These notes are discussed during the pre-job brief.

Fuel moves are coordinated and independently verified by a qualified FHC. The PNP fuel handling machine has specific storage cell coordinates that are pre-programmed into the spent fuel handling machine (SFHM) computer. This is a physical control that lessens the probability of fuel move errors. For each fuel move, these coordinates are verified by the qualified FHC. For Regions 1D and 1E, which do not have cell coordinates in the SFHM computer, SOP-28 requires the qualified FHC to verify the cell location identified by the fuel handling machine operator.

Administrative Procedure 4.00, "Operations Organization, Responsibilities and Conduct," provides the controls necessary to conduct fuel handling activities safely and effectively, and ensures adherence with fuel movement plans.

The following human performance tools are used during fuel handling activities;

1. Three-way communications between the fuel handling machine operator and the FHC is used during verification of "from" and "to" locations, and during verification of fuel handling machine mast orientation.

2. Place-keeping on the fuel moves sheets is required for each fuel move step.

3. The performance of fuel move evolutions is preceded with formal pre-job briefs.

Page 15 of 21

The above controls are considered appropriate to minimize the probability of the occurrence of a fuel misload event.

5.0 REGULATORY SAFETY ANALYSIS No Significant Hazards Consideration Entergy Nuclear Operations, Inc. (ENO) has evaluated whether or not a significant hazards consideration is involved with the proposed amendment to the spent fuel assembly storage in Technical Specification (TS) 3.7.16 and the fuel storage criticality requirements in TS section 4.3, using the three standards set forth in 10 CFR 50.92, "Issuance of Amendment," as discussed below:

1. Does the proposed amendment involve a significant increase in the probability or consequences of an accident previously evaluated?

Response: No.

There is no significant increase in the probability of an accidental misloading of fuel assemblies into the spent fuel pool (SFP) racks when considering the presence of soluble boron in the pool water for criticality control and the proposed changes. The proposed changes credit fuel burnup and voiding of the gaps between the SFP rack individual storage cells. Fuel assembly placement would continue to be controlled by approved fuel handling procedures and would be in accordance with the TS fuel storage rack configuration limitations.

There is no significant increase in the consequences of the accidental misloading of fuel assemblies into the SFP racks. The criticality analyses that credit fuel burnup and voiding of the gaps between the SFP rack individual storage cells demonstrate that the pool would remain subcritical with margin following an accidental misloading if the pool contains an adequate boron concentration. The TS 3.7.15 limitation on minimum SFP boron concentration and plant procedures together ensure that an adequate boron concentration will be maintained.

There is no significant increase in the probability of a fuel assembly drop accident in the SFP when considering the presence of soluble boron in the SFP water for criticality control, credit fuel burnup, and voiding of the gaps between the SFP rack individual storage cells. The handling of fuel assemblies in the spent fuel is performed in accordance with site procedures in borated water. The criticality analysis has shown that the reactivity increase with a fuel assembly drop accident in both a vertical and horizontal orientation is bounded by the fuel assembly misloading accident. Therefore, in addition to there being no significant increase in the probability of a fuel assembly drop accident, the consequences of a fuel assembly drop accident in the SFP would not increase significantly due to the proposed change.

Page 16 of 21

The SFP TS 3.7.15 requires a minimum boron concentration of 1720 ppm, which bounds the analysis for the proposed amendment. Soluble boron has been maintained in the SFP water as required by TS and controlled by procedures. The criticality safety analyses for Region I and Region II of the SFP credit the same soluble boron concentration of 850 ppm to maintain a Keff < 0.95 under normal conditions and 1350 ppm to maintain a Keff < 0.95 under accident scenarios as does the analysis for the proposed change for Region I, Regions 1A, 1B, 1C, 1D, and 1E. In crediting soluble boron, in Region 1A, and soluble boron and burnup, in Regions 1B, 1C, 1D, and 1E, the SFP criticality analysis would have no effect on normal pool operation and maintenance. Credit for fuel burnup and voiding of the gaps between the SFP rack individual storage cells would have no effect on the normal SFP operation and maintenance. Thus, there is no change to the probability or the consequences of the boron dilution event in the SFP.

Since soluble boron is maintained in the SFP water, implementation of the proposed changes would have no effect on normal pool operation and maintenance. Also, since soluble boron is present in the SFP, a dilution event has always been a possibility. The loss of substantial amounts of soluble boron from the SFP was evaluated as part of the analyses in support of this proposed amendment. The analyses use the same soluble boron concentrations as were used in previous analyses for the Region I and Region II spent fuel storage racks. The SFP Regions 1A, 1B, 1C, 1 D, and 1E storage racks are analyzed to allow storage of the fuel applying a burnup credit (for regions 1 B, 1C, 1D, and 1E), a complete loss of Carborundum plates and complete voiding of the gaps between the SFP individual storage cells. A minimum margin of 0.0117 is calculated for the boron dilution events with respect to 10 CFR 50.68 criteria, both borated and unborated. All abnormal conditions meet the 0.95 criterion at 1350 ppm of boron.

Therefore, the limitations on boron concentration have not changed and would not result in a significant increase in the probability or consequences of a previously evaluated accident.

There is no increase in the probability or consequences of the loss of normal cooling to the SFP water, when considering this change that credits fuel burnup, voiding of the gaps between the SFP rack individual storage cells, and the presence of soluble boron in the pool water for subcriticality control, since a high concentration of soluble boron is always maintained in the SFP.

The criticality analyses documented in AREVA NP Inc. report ANP-2858P-003, "Palisades SFP Region 1 Criticality Evaluation with Burnup Credit," show, at a 95 percent probability and a 95 percent confidence level (95/95), that Keff is less than the regulatory limit in 10 CFR 50.68 of 0.95 under borated conditions, or the limit of 1.0 with unborated water. Therefore, the consequences of accidents previously evaluated are not increased.

Therefore, it is concluded that the proposed change does not significantly increase the probability or consequences of any accident previously evaluated.

Page 17 of 21

2. Does the proposed amendment create the possibility of a new or different kind of accident from any accident previously evaluated?

Response: No.

Spent fuel handling accidents have been analyzed in Sections 14.11, "Postulated Cask Drop Accidents," and 14.19, "Fuel Handling Incident," of the Updated Final Safety Analysis Report. Criticality accidents in the SFP have been analyzed in previous criticality evaluations, which are the bases for the existing TS.

The existing TS allow storage of fuel assemblies with a maximum planar average U-235 enrichment of 4.54 weight percent in the Region 1A fuel storage rack, 4.34 weight percent in the Region 1B storage rack, and 3.05 weight percent in the 1E Region storage rack with the exception of one assembly in Region 1 E having a maximum planar average U-235 enrichment of 3.26 weight percent. The proposed specifications would allow fuel enrichment to 4.54 weight percent in existing Regions 1B, and 1 E and for new Regions 1C and 1 D with minimum enrichment dependent burnup restrictions.

The existing Region 1A enrichment of 4.54 weight percent is unchanged in the proposed specifications. The possibility of placing a fuel assembly with greater enrichment than allowed currently exists but is controlled by the fuel manufacturer's procedures and plant fuel handling procedures. These manufacturer's and plant procedural controls would remain in place. Changing the allowed enrichments does not create a new or different kind of accident.

ENO considered the effects of a mispositioned fuel assembly. The proposed loading restrictions include locations that are prohibited from containing any fuel. Administrative controls are in place to restrict fuel moves to those locations. These controls include procedures to develop the plans for fuel movement and operation of the fuel handling equipment. These procedures include appropriate reviews and verifications to ensure that TS requirements are maintained.

Furthermore, the existing TS contain limitations on the SFP boron concentration that conservatively bound the required boron concentration of the new criticality analysis.

Currently, TS 3.7.15 requires a minimum boron concentration of 1720 ppm. Since soluble boron is maintained in the SFP water, implementation of the proposed changes would have no effect on normal pool operation and maintenance. Since soluble boron is present in the SFP, a dilution event has always been a possibility. The loss of substantial amounts of soluble boron from the SFP was evaluated as part of the analysis in support of Amendment No. 207. The analysis also demonstrated that, due to the large volume of unborated water that would need to be added and displaced, and the long duration of the event, the condition would be detected and corrected promptly.

The analyses that support the current request use the same soluble boron concentrations that were used in previous analyses for the Region I and Region II spent fuel storage racks. In the unlikely event that soluble boron in the SFP is completely diluted, the fuel in Region I, Regions 1A, 1B, 1C, 1 D, and 1E of the SFP would remain subcritical by a design margin of at least 0.0117 delta K, so the Keff of the fuel in Region 1 would remain below 1.0.

Page 18 of 21

The combination of controls to prevent a mispositioned fuel assembly, the ability to readily identify and correct a dilution event, and the relatively high concentration of soluble boron supports a conclusion that a new or different kind of accident is not created.

Under the proposed amendment, no changes are made to the fuel storage racks themselves, to any other systems, or to any plant structures. Therefore, the change will not result in any other change in the plant configuration or equipment design.

Therefore, the proposed change does not create the possibility of a new or different kind of accident from any previously evaluated.

3. Does the proposed amendment involve a significant reduction in a margin of safety?

Response: No.

Detailed analysis, with approved and benchmarked methods has shown, with a 95 percent probability at a 95 percent confidence level, that the Keff of the Region I, Region 1A, 1B, 1C, 1 D, and 1E, fuel storage racks in the SFP, including biases, tolerances and uncertainties, is less than 1.0 with unborated water and is less than or equal to 0.95 with 850 ppm of soluble boron and burnup credited (for Regions 1B, 1C, 1D, and 1 E), along with complete voiding of the gaps between the individual storage cells in the SFP racks. In addition, the effects of abnormal and accident conditions have been evaluated to demonstrate that under credible conditions the Keff will not exceed 0.95 with 1350 ppm soluble boron and burnup credited. The current TS requirement for minimum SFP boron concentration is 1720 ppm, which provides assurance that the SFP would remain subcritical under normal, abnormal, or accident conditions.

The current analysis basis for the Region I and Region II fuel storage racks is a maximum Keff of less than 1.0 when flooded with unborated water, and less than or equal to 0.95 when flooded with water having a boron concentration of 850 ppm. In addition, the Keff in accident or abnormal operating conditions is less than 0.95 with 1350 ppm of soluble boron. These values are not affected by the proposed change.

Therefore, it is concluded that the proposed change does not involve a significant reduction in the margin of safety.

Conclusion Based on the evaluation above, ENO concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified.

Page 19 of 21

Applicable Regulatory Requirements/Criteria The SFP storage racks maintain fresh and irradiated assemblies in a safe storage condition. The federal code requirements in 10 CFR 50.68 specify the normal and accident parameters associated with maintaining fresh and irradiated fuel assemblies in a safe storage condition. 10 CFR 50.68 defines the criticality accident requirements associated with the fuel storage racks and states the following in section 50.68(b)(4): "If no credit for soluble boron is taken, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with unborated water. If credit is taken for soluble boron, the k-effective of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95 percent probability, 95 percent confidence level, if flooded with borated water, and the k- effective must remain below 1.0 (subcritical), at a 95 percent probability, 95 percent confidence level, if flooded with unborated water."

The evaluation in Attachment 5 provides results of analyses for Region I, Regions. 1A, 1B, 1C, 1 D, and 1E that, with burnup credit (for Regions 1B, 1C, 1D, and 1E) and complete voiding of the gaps between the SFP rack individual storage cells, demonstrate the Keff is less than 1.0 with the racks loaded with fuel of the highest anticipated reactivity, and flooded with unborated water at a temperature corresponding to the highest reactivity. In addition, with burnup credit (for Regions 1B, 1C, 1D, and 1 E), the analyses demonstrate that Keff is less than or equal to 0.95 with the racks loaded with fuel of the highest anticipated reactivity, and flooded with borated water at a temperature corresponding to the highest reactivity. The maximum calculated reactivity included a margin for uncertainty in reactivity calculations including manufacturing tolerances, and the Keff is shown to be less than 0.95 with a 95 percent probability at a 95 percent confidence level with boron credit. Reactivity effects of abnormal and accident conditions were also evaluated to assure that under all credible abnormal and accident conditions, the Keff will not exceed the regulatory limit of 0.95 under borated conditions or the limit of 1.0 with unborated water.

The double-contingency principle of ANS-8.1/ANSI N16.1-1975 and NRC letter, of April 14, 1978, specifies that at least two unlikely, independent and concurrent events are required before a criticality accident is possible. This principle precludes the necessity of considering the simultaneous occurrence of multiple accident conditions.

10 CFR Section 50.36(c)(4) requires, "Design features to be included are those features of the facility such as materials of construction and geometric arrangements, which, if altered or modified, would have a significant effect on safety and are not covered in categories described in paragraphs (c)(1), (2), and (3) of this section." The proposed LAR provides changes to the Design Features section of the TS to meet this requirement.

The following applicable codes, standards, regulations and guidance, or pertinent sections thereof, were used in the analyses described in Attachment 5:

Page 20 of 21

10 CFR 50, Appendix A, General Design Criterion 62, "Prevention of Criticality in Fuel Storage and Handling"

10 CFR 50.68, "Criticality accident requirements" 0 NUREG-0800 "Standard Review Plan," Section 9.1.1, "Criticality Safety of Fresh and Spent Fuel Storage and Handling," Revision 3, March 2007 0 NRC letter to all Power Reactor Licensees dated April 14, 1978, Enclosure No. 1, "OT Position for Review and Acceptance of Spent Fuel Storage and Handling Applications" (GL-78-01 1), including modification letter dated January 18, 1979 (GL-79-004) 0 NRC Memorandum from L. Kopp to T. Collins dated August 19, 1998, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants" 0 Regulatory Guide 1.13, "Spent Fuel Storage Facility Design Basis," Revision 2, March 2007

ANSI ANS-8.17-1984, "Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors"

NUREG/CR-6698, "Guide for Validation of Nuclear Criticality Safety Calculational Methodology," January 2001

NUREG/CR-6979, "Evaluation of the French Haut Taux de Combustion (HTC)

Critical Experiment Data," September 2008

6.0 ENVIRONMENTAL CONSIDERATION

The proposed amendment would change a requirement with respect to installed facility components located within the restricted area of the plant as defined in 10 CFR 20.

However, the proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

Page 21 of 21

ATTACHMENT 2 RENEWED OPERATING LICENSE PAGE CHANGE INSTRUCTIONS AND REVISED TECHNICAL SPECIFICATION PAGES 3.7.16-1 through 3.7.16-6 and 4.0-1 through 4.0-6 13 pages follow

ATTACHMENT TO LICENSE AMENDMENT NO.

RENEWED FACILITY OPERATING LICENSE NO. DPR-20 DOCKET NO. 50-255 Remove the following pages of Appendix A Technical Specifications and replace with the attached revised pages. The revised pages are identified by amendment number and contain lines in the margin indicating the areas of change.

REMOVE INSERT Page 3.7.16-1 through 3.7.16-2 Page 3.7.16-1 through 3.7.16-6 Pages 4.0-1 through 4.0-9 Pages 4.0-1 through 4.0-6

Spent Fuel Pool Storage 3.7.16 3.7 PLANT SYSTEMS 3.7.16 Spent Fuel Pool Storage LCO 3.7.16 Storage in the spent fuel pool shall be as follows:

a. Each fuel assembly and non-fissile bearing component stored in Region I shall be within the limitations in Specification 4.3.1.1 and, as applicable, within the requirements of the maximum nominal planar average U-235 enrichment and burnup of Tables 3.7.16-2, 3.7.16-3, 3.7.16-4 or 3.7.16-5; and

b. The combination of maximum nominal planar average U-235 enrichment, burnup, and decay time of each fuel assembly stored in Region II shall be within the requirements of Table 3.7.16-1.

APPLICABILITY: Whenever any fuel assembly or non-fissile bearing component is stored in the spent fuel pool or the north tilt pit.

ACTIONS LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO A.1 Initiate action to restore Immediately not met. the noncomplying fuel assembly or non-fissile bearing component within requirements.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify by administrative means each fuel assembly Prior to storing the or non-fissile bearing component meets fuel fuel assembly or storage requirements. non-fissile bearing component in the spent fuel pool Palisades Nuclear Plant 3.7.16-1 Amendment No. 41-89, 207, 236, 2xx

Spent Fuel Pool Storage 3.7.16 Table 3.7.16-1 (page 1 of 1)

Spent Fuel Minimum Burnup and Decay Requirements for Storage in Region II of the Spent Fuel Pool and North Tilt Pit Nominal Planar Average U-235 Burnup Burnup Burnup Burnup Burnup Enrichment (GWD/MTU) (GWD/MTU) (GWD/MTU) (GWD/MTU) (GWD/MTU)

(Wt%) No Decay 1 Year Decay 3 Year Decay 5 Year Decay 8 Year Decay

1.14 0 0 0 0 0

> 1.14 3.477 3.477 3.477 3.477 3.477 1.20 3.477 3.477 3.477 3.477 3.477 1.40 7.951 7.844 7.464 7.178 6.857 1.60 11.615 11.354 10.768 10.319 9.847 1.80 14.936 14.535 13.767 13.187 12.570 2.00 18.021 17.502 16.561 15.875 15.117 2.20 21.002 20.417 19.313 18.499 17.611 2.40 23.900 23.201 21.953 21.034 20.050 2.60 26.680 25.905 24.497 23.487 22.378 2.80 29.388 28.528 27.006 25.879 24.678 3.00 32.044 31.114 29.457 28.243 26.942 3.20 34.468 33.457 31.698 30.397 29.008 3.40 36.848 35.783 33.920 32.544 31.079 3.60 39.152 38.026 36.059 34.615 33.077 3.80 41.419 40.226 38.163 36.650 35.049 4.00 43.661 42.422 40.257 38.673 37.007 4.20 45.987 44.684 42.415 40.778 39.028 4.40 48.322 46.950 44.588 42.877 41.041 4.60 50.580 49.158 46.690 44.911 43.003 (a) Linear interpolation between two consecutive points will yield acceptable results.

(b) Comparison of nominal assembly average burnup numbers to these in the table is acceptable if measurement uncertainty is < 10%.

Palisades Nuclear Plant 3.7.16-2 Amendment No. 9-89, 2Q7, 2-6, 2xx

Spent Fuel Pool Storage 3.7.16 Table 3.7.16-2 (page 1 of 1)

Spent Fuel Minimum Burnup Requirements for Storage in Region 1 B (three-of-four loading configuration) of the Main Spent Fuel Pool Nominal Planar Burnup Burnup Average U-235 (GWD/MTU) (GWD/MTU)

Enrichment (Batches L (Batches A (Wt%) and later) through K)

<2.10 0 1.0 2.40 4.1 5.1 2.60 6.7 7.7 2.80 9.5 10.5 3.00 12.2 13.2 3.20 14.9 15.9 3.40 17.6 18.6 3.60 20.2 21.2 3.80 23.0 24.0 4.00 25.7 26.7 4.20 28.4 29.4 4.40 31.1 32.1 4.54 33.0 34.0 (a) Linear interpolation between two consecutive points for nominal planar average U-235 enrichments between 2.10 and 4.54 will yield acceptable results.

(b) Comparison of nominal assembly average burnup numbers to these in the table is acceptable if measurement uncertainty is < 10%.

Palisades Nuclear Plant 3.7.16-3 Amendment No. 2xx

Spent Fuel Pool Storage 3.7.16 Table 3.7.16-3 (page 1 of 1)

Spent Fuel Minimum Burnup Requirements for Storage in Region 1C (four-of-four loading configuration) of the Main Spent Fuel Pool Nominal Planar Burnup Burnup Average U-235 (GWD/MTU) (GWD/MTU)

Enrichment (Batches L (Batches A (Wt%) and later) through K)

<1.35 0 1.0 2.40 20.7 21.7 2.60 24.5 25.5 2.75 27.5 28.5 2.80 28.2 29.2 3.00 31.0 32.0 3.20 33.9 34.9 3.40 36.7 37.7 3.60 39.5 40.5 3.80 42.4 43.4 4.00 45.2 46.2 4.20 48.0 49.0 4.40 50.8 51.8 4.54 52.8 53.8 (a) Linear interpolation between two consecutive points for nominal planar average U-235 enrichments between 1.35 and 4.54 will yield acceptable results.

(b) Comparison of nominal assembly average burnup numbers to these in the table is acceptable if measurement uncertainty is < 10%.

Palisades Nuclear Plant 3.7.16-4 Amendment No. 2xx

Spent Fuel Pool Storage 3.7.16 Table 3.7.16-4 (page 1 of 1)

Spent Fuel Minimum Burnup Requirements for Storage in Region 1 D (three-of four loading configuration) of the North Tilt Pit Nominal Planar Burnup Burnup Average U-235 (GWD/MTU) (GWD/MTU)

Enrichment (Batches L (Batches A (Wt%) and later) through K)

<2.35 0 1.0 2.40 0.5 1.5 2.60 2.4 3.4 2.80 4.3 5.3 3.00 6.2 7.2 3.20 8.1 9.1 3.40 10.0 11.0 3.60 11.9 12.9 3.80 13.8 14.8 4.00 15.7 16.7 4.20 17.7 18.7 4.40 19.6 20.6 4.54 20.9 21.9 (a) Linear interpolation between two consecutive points for nominal planar average U-235 enrichments between 2.35 and 4.54 will yield acceptable results.

(b) Comparison of nominal assembly average burnup numbers to these in the table is acceptable if measurement uncertainty is < 10%.

Palisades Nuclear Plant 3.7.16-5 Amendment No. 2xx

Spent Fuel Pool Storage 3.7.16 Table 3.7.16-5 (page 1 of 1)

Spent Fuel Minimum Burnup Requirements for Storage in Region 1 E (four-of-four loading configuration) of the North Tilt Pit Nominal Planar Burnup Burnup Average U-235 (GWD/MTU) (GWD/MTU)

Enrichment (Batches L (Batches A (Wt%) and later) through K)

<1.48 0 1.0 2.40 13.9 14.9 2.60 16.9 17.9 2.80 19.9 20.9 3.00 23.0 24.0 3.20 26.0 27.0 3.30 27.5 28.5 3.40 28.7 29.7 3.60 31.0 32.0 3.80 33.3 34.3 4.00 35.6 36.6 4.20 37.9 38.9 4.40 40.2 41.2 4.54 41.8 42.8 (a) Linear interpolation between two consecutive points for nominal planar average U-235 enrichments between 1.48 and 4.54 will yield acceptable results.

(b) Comparison of nominal assembly average burnup numbers to these in the table is acceptable if measurement uncertainty is < 10%.

Palisades Nuclear Plant 3.7.16-6 Amendment No. 2xx

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The Palisades Nuclear Plant is located on property owned by Entergy Nuclear Palisades, LLC on the eastern shore of Lake Michigan approximately four and one-half miles south of the southern city limits of South Haven, Michigan. The minimum distance to the boundary of the exclusion area as defined in 10 CFR 100.3 shall be 677 meters.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor core shall contain 204 fuel assemblies. Each assembly shall consist of a matrix of zircaloy-4 or M5 clad fuel rods with an initial composition of depleted, natural, or slightly enriched uranium dioxide (U0 2) as fuel material.

Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. A core plug or plugs may be used to replace one or more fuel assemblies subject to the analysis of the resulting power distribution. Poison may be placed in the fuel bundles for long-term reactivity control.

4.2.2 Control Rod Assemblies The reactor core shall contain 45 control rods. Four of these control rods may consist of part-length absorbers. The control material shall be silver-indium-cadmium, as approved by the NRC.

4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The Region I fuel storage racks (See Figure B 3.7.16-1) incorporating Regions 1A, 1B, 1C, 1D, and 1E are designed and shall be maintained with:

a. New or irradiated fuel assemblies having a maximum nominal planar average U-235 enrichment of 4.54 weight percent; Palisades Nuclear Plant 4.0-1 Amendment No. 1-89, 207, 224, 234, 236, 2xx

Design Features 4.0 4.3 Fuel Storage 4.3.1 Criticality (continued)

b. Keff < 1.0 if fully flooded with unborated water, which includes allowances for uncertainties as described in Section 9.11 of the FSAR;

c. Keff < 0.95 if fully flooded with water borated to 850 ppm, which includes allowances for uncertainties as described in Section 9.11 of the FSAR;

d. Regions 1A, 1B, and 1C have a nominal 10.25 inch center to center distance between fuel assemblies;

e. Regions 1D and 1E have a nominal 11.25 inch by 10.69 inch center to center distance between fuel assemblies;

f. Region 1A is defined as a subregion of the Region I storage racks located in the main spent fuel pool and is subject to the following restrictions. Fuel assemblies (or fissile bearing components) located in Region 1A shall be in a maximum of two-of-four checkerboard loading pattern of two fuel assemblies (or fissile bearing components) and two empty cells.

Designated empty cells may contain non-fuel bearing components in accordance with Section 4.3.1.1 m.2. below;

g. Region 1 B is defined as a subregion of the Region I storage racks located in the main spent fuel pool and is subject to the following restrictions. Fuel assemblies (or fissile bearing components) located in Region 1 B shall be in a maximum of three-of-four loading pattern consisting of three fuel assemblies (or fissile bearing components) and one empty cell. Fuel assemblies in Region 1 B shall meet the enrichment dependent burnup restrictions listed in Table 3.7.16-2. Designated empty cells may contain non-fuel bearing components in accordance with Section 4.3.1.1m.2. below;

h. Region 1C is defined as a subregion of the Region I storage racks located in the main spent fuel pool and is subject to the following restrictions. Fuel assemblies (or fissile bearing components) located in Region 1C may be in a maximum of four-of-four loading pattern with no required empty cells. Fuel assemblies in Region 1C shall meet the enrichment dependent burnup restrictions listed in Table 3.7.16-3; Interface requirements for the mainspent fuel pool between Region 1A, 1 B, and 1C are as follows. Region 1A, 1 B, and 1C can be distributed in Region I, in the main spent fuel pool, in any manner provided that any two-by-two grouping of storage cells and the assemblies in them correspond to the requirements of 4.3.1 .lf., 4.3.1.1g., or 4.3.1.1 h. above; Palisades Nuclear Plant 4.0-2 Amendment No. 489, 2-W, 236, 2xx

Design Features 4.0 4.3 Fuel Storage 4.3.1 Criticality (continued)

j. Region 1 D is defined as a subregion of the Region I storage rack located in the north tilt pit and is subject to the following restrictions. Fuel assemblies (or fissile bearing components) located in Region 1D may be in a maximum of three-of-four loading pattern consisting of three fuel assemblies (or fissile bearing components) and one empty cell. Fuel assemblies in Region 1 D shall meet the enrichment dependent burnup restrictions listed in Table 3.7.16-4;

k. Region 1 E is defined as a subregion of the Region I storage rack located in the north tilt pit and is subject to the following restrictions. Fuel assemblies (or fissile bearing components) located in Region 1E may be in a maximum of four-of-four loading pattern with no required empty cells. Fuel assemblies in Region 1 E shall meet the enrichment dependent burnup restrictions listed in Table 3.7.16-5; Interface requirements for the north tilt pit between Region 1 D and 1 E are as follows. Region 1 D and 1 E can be distributed in Region I in the north tilt pit in any manner provided that any two-by-two grouping of storage cells and the assemblies in them correspond to the requirements of 4.3.1 .1j. or 4.3.1.1 k.

above;

m. Non-fissile bearing component restrictions are as follows:

1. Non-fissile material components may be stored in any designated fuel location in Region 1A, 1B, 1C, 1D, or 1E without restriction.

2. The following non-fuel bearing components (NFBC) may be stored face adjacent to fuel in any designated empty cell in Region 1A or 1 B.

(i) The gauge dummy assembly and the lead dummy assembly may be stored face adjacent to fuel in any designated empty cells with no minimum required separation distance.

(ii) A component comprised primarily of stainless steel that displaces less than 30 square inches of water in any plane within the active fuel region may be stored in any designated empty cell as long as the NFBC is at least ten locations away from any other NFBC that is in a designated empty cell, with the exception of 4.3.1.1 m.2.(i) above.

Palisades Nuclear Plant 4.0-3 Amendment No. 236, 2xx

Design Features 4.0 4.3 Fuel Storage 4.3.1 Criticality (continued)

3. Control blades may be stored in both fueled and unfueled locations in Regions 1 D and 1 E, with no limitation on the number.

4.3.1.2 The Region II fuel storage racks (See Figure B 3.7.16-1) are designed and shall be maintained with;

a. Fuel assemblies having maximum nominal planar average U-235 enrichment of 4.60 weight percent;

b. Keff < 1.0 if fully flooded with unborated water, which includes allowances for uncertainties as described in Section 9.11 of the FSAR.

c. Keff < 0.95 if fully flooded with water borated to 850 ppm, which includes allowance for uncertainties as described in Section 9.11 of the FSAR.

d. A nominal 9.17 inch center to center distance between fuel assemblies; and

e. New or irradiated fuel assemblies which meet the maximum nominal planar average U-235 enrichment, burnup, and decay time requirements of Table 3.7.16-1.

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

a. Twenty four unirradiated fuel assemblies having a maximum nominal planar average U-235 enrichment of 4.95 weight percent, and stored in accordance with the pattern shown in Figure 4.3-1, or Thirty six unirradiated fuel assemblies having a maximum nominal planar average U-235 enrichment of 4.05 weight percent, and stored in accordance with the pattern shown in Figure 4.3-1;

b. Keff < 0.95 when flooded with either full density or low density (optimum moderation) water including allowances for uncertainties as described in Section 9.11 of the FSAR.

c. The pitch of the new fuel storage rack lattice being > 9.375 inches and every other position in the lattice being permanently occupied by an 8" x 8" structural steel or core plugs, resulting in a nominal 13.26 inch center to center distance between fuel assemblies placed in alternating storage locations.

Palisades Nuclear Plant 4.0-4 Amendment No. 489, 207, 236, 2xx

4.3 Fuel Storage 4.3.2 Drainage The spent fuel storage pool cooling system suction and discharge piping is designed and shall be maintained to prevent inadvertent draining of the pool below elevation 644 ft 5 inches.

4.3.3 Capacity The spent fuel storage pool and north tilt pit are designed and shall be maintained with a storage capacity limited to no more than 892 fuel assemblies.

Palisades Nuclear Plant 4.0-5 Amendment No. 489, 247, 236

1:O 1:0 11I00ODODO0ETO

LZOD@I01L:10Ol 0 D 0 1:

EOl0 El0Zoll:]0 OT0ZO1 CENTERLINE- LEGEND PATTERN REPEATS D 8X 8STEEL BOX BEAM Q ASSEMBLY STORAGE LOCATION (ENRICHMENT - 4.95 WT%U-235)

ASSEMBLY STORAGE LOCATION (ENRICHMENT <= 4.05 WT%U-235)

Note: If any assemblies containing fuel enrichments greater than 4.05% U-235 are stored in the New Fuel Storage Rack, the center row must remain empty.

Figure 4.3-1 (page 1 of 1)

New Fuel Storage Rack Arrangement Palisades Nuclear Plant 4.0-6 Amendment No. 2-W, 236

ATTACHMENT 3 MARK-UP OF TECHNICAL SPECIFICATIONS PAGES (showing proposed changes; additions are highlighted and deletions are strikethrough) 15 pages follow

Spent Fuel Pool Storage 3.7.16 3.7 PLANT SYSTEMS 3.7.16 Spent Fuel Pool Storage LCO 3.7.16 Storage in the spent fuel pool shall be as follows:

a. Each fuel assembly and non-fissile bearing component stored in Region I shall be within the limitations in Specification 4.3.1 .1nd the requirements of the Imo

.v,*n-miL planar averg U-235 enrichment and burnup of Tables3Z 317.i613, 17!64 or 3.7.16-5i* and

b. The combination of aximum nominal .Pjgar avýOire- 3_PtiaJ enrichment, burnup, and decay time of each fuel assembly stored in Region II shall be within the requirements of Table 3.7.16-1.

APPLICABILITY: Whenever any fuel assembly or non-fissile bearing component is stored in the spent fuel pool or the north tilt pit.

ACTIONS

NOTE- -------------------------------

LCO 3.0.3 is not applicable.

CONDITION REQUIRED ACTION COMPLETION TIME A. Requirements of the LCO A.1 Initiate action to restore Immediately not met. the noncomplying fuel assembly or non-fissile bearing component within requirements.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.16.1 Verify by administrative means each fuel assembly Prior to storing the or non-fissile bearing component meets fuel fuel assembly or storage requirements. non-fissile bearing component in the spent fuel pool Palisades Nuclear Plant 3.7.16-1 Amendment No. 1-89, 207, 236

Spent Fuel Pool Storage 3.7.16 TABLQ5ib 3.7.16-1 (page 1 of 1)

Spent Fuel Minimum Burnup and Decay Requirements for Storage in Region II of the Spent Fuel Pool and North Tilt Pit oomiina lana J~erarg

-2 4tial Burnup Burnup Burnup Burnup Burnup Enrichment (GWD/MTU) (GWD/MTU) (GWD/MTU) (GWD/MTU) (GWD/MTU)

(Wt%) No Decay 1 Year Decay 3 Year Decay 5 Year Decay 8 Year Decay

1.14 0 0 0 0 0

> 1.14 3.477 3.477 3.477 3.477 3.477 1.20 3.477 3.477 3.477 3.477 3.477 1.40 7.951 7.844 7.464 7.178 6.857 1.60 11.615 11.354 10.768 10.319 9.847 1.80 14.936 14.535 13.767 13.187 12.570 2.00 18.021 17.502 16.561 15.875 15.117 2.20 21.002 20.417 19.313 18.499 17.611 2.40 23.900 23.201 21.953 21.034 20.050 2.60 26.680 25.905 24.497 23.487 22.378 2.80 29.388 28.528 27.006 25.879 24.678 3.00 32.044 31.114 29.457 28.243 26.942 3.20 34.468 33.457 31.698 30.397 29.008 3.40 36.848 35.783 33.920 32.544 31.079 3.60 39.152 38.026 36.059 34.615 33.077 3.80 41.419 40.226 38.163 36.650 35.049 4.00 43.661 42.422 40.257 38.673 37.007 4.20 45.987 44.684 42.415 40.778 39.028 4.40 48.322 46.950 44.588 42.877 41.041 4.60 50.580 49.158 46.690 44.911 43.003 (a) Linear interpolation between two consecutive points will yield acceptable results.

(b) Comparison of nominal assembly average burnup numbers to these in the table is acceptable if measurement uncertainty is _<10%.

Palisades Nuclear Plant 3.7.16-2 Amendment No. 4-89, 207, 236

Spent Fuel Pool Storage 3.7.16 F

.6:0 U____

U64O 18.E

.0-M-0 M_ _ _

~KMO E5.2 _ _ _

_ ._ __O 1.1 E9_.]

Linear interpolation between two consecutive points for nominal planar average U-23_5

-nrichments between 2.10 and 4.54 will yield acceptable results.

b Comparison of nominal assembly average burnup numbers to these in the table is acceptable il easurement uncertainty is <

Spent Fuel Pool Storage 3.7.16 rble 3.7.16-3 (page 1 ,1 of torge

___~on ~upup POTni 14n~1 *(f~lrf*F l our Requirementsfo loa ng co nfigura ion9 f the MainSpent Fuel Poor ominal Plana urnu urn verage U-235 GWD/MTU) GWID/MTU)

Th-nrichmen Batches Btche

ýiwt%U ndllater,' hrou h K g1. 31 __El Rio E$

R._751 P7.-5'R

.40 _ _ _ E.

[g) -...inear interpolation ,etwee. , , " i points for nominal planar aV ge U-235 enrichments

_MComparison of nominalsmbly averagaburnu numbers to4these Inthe table is acceptablei

--i easurement uncertaintyI is _ 10%.

Spent Fuel Pool Storage 3.7.16 able 3.7.16-4 (page 1 of 1' pent Fuel Minimum Burnup Requirements fo tora e in Region, D (three-of-fOur Ioing configur )

of the North Tilt Pi ominal Plan, verage~I~p~

n irichm ~GD/MU GPM Wt%1 atches Bahs "nd laerU hogK

.260 H8.E4

.401 0t _

Ka) Linear interpolation between two consecutive points for nominal planar average U-235 ilnrichments between 2.35 and 4.54 will yield acceptable resultsI Kb) C~mparison of nominal ans mbly average burnup numbers to these in the table iSacceptable i*

ment uree unceI~rfy1%

Spent Fuel Pool Storage 3.7.16 able 37,1~-5(page 1 of 1, age in Re._gion 1E (four-of-four loadig configuration

- f the North Tilt Pi Nominal Plana verage U-ý23 urnu urnu

%Enrichmen D/TGWD/MTU Btche Batches f-NSIE qT fnd later) hroughtK)'

NEW -rAbLE

-3.7-1-6-5'

_ _6 _ 1 6.9 _ _ _

.8=0 19.919

___ 3. -01 RA2

.f30 E7.__ 0 V40 E8.]7

.60] ________

K._ 7=

l.9 2__

Ka)_Linear interpolation between two consecutive points for nominal planar average U-235

.nrichments between 1.48 and 4.54 will yield acceptable resultsý b Comparison of nominal assembly arverage burnup numbers to these in the table is acceptable il easurement uncertainty is _<10%,

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The Palisades Nuclear Plant is located on property owned by Entergy Nuclear Palisades, LLC on the eastern shore of Lake Michigan approximately four and one-half miles south of the southern city limits of South Haven, Michigan. The minimum distance to the boundary of the exclusion area as defined in 10 CFR 100.3 shall be 677 meters.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor core shall contain 204 fuel assemblies. Each assembly shall consist of a matrix of zircaloy-4 or M5 clad fuel rods with an initial composition of depleted, natural, or slightly enriched uranium dioxide (UO 2 ) as fuel material.

Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods, in accordance with approved applications of fuel rod configurations, may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. A core plug or plugs may be used to replace one or more fuel assemblies subject to the analysis of the resulting power distribution. Poison may be placed in the fuel bundles for long-term reactivity control.

4.2.2 Control Rod Assemblies The reactor core shall contain 45 control rods. Four of these control rods may consist of part-length absorbers. The control material shall be silver-indium-cadmium, as approved by the NRC.

4.3 Fuel Storage 4.3.1 Criticality 4.3.1.1 The Region I fuel storage racks (See Figure B 3.7.16-1) incorporating Regions 1 A, 1 B, ftij and 1 E are designed and shall be maintained with:

a. W r i fuel assemblies having a maximum nominal planar average U-235 enrichment of 4.54 weight percentq iR Region 1A, 4.34 weight percent in reg.ion 1 B, and 3.05 welight percent in Rogio 1E with. th. ..o..,Ic,,t*,

ofo *ae*Mb in Region l E, described in 41.3. 1.li0 below, having a maximum nominRal p*lanr average UI25 enrichmont of 3.226 weight PeFGe~t.

Palisades Nuclear Plant 4.0-1 Amendment No. 8, 2--, 224, 2-34, 236

4.3 Fuel Storage 4.3.1 Criticality (continued)

b. Keff < 1.0 if fully flooded with unborated water, which includes allowances for uncertainties as described in Section 9.11 of the FSAR;

c. Keff < 0.95 if fully flooded with water borated to 850 ppm, which includes allowances for uncertainties as described in Section 9.11 of the FSAR;

d. f*e:dis-,1 B, and 1C hav* a nominal 10.25 inch center to center distance between fuel assemblies With the exception of the single Type E rack which has a nominal 11.25 i by 10.69 inch canRtor to cantor distance1 between fuel assemnblfies

e. New or irradiated fuel assemblies and 10*haveo omina 11.25 inch by 10;.69 inc ce er, to center distanc -

etween fuel assemblie -

Rlegion !A is defined as the Region 1 storage rackshl*ocAted I f.

All -C I I four checkerboard loadwith empty cells a, shoWnin the figure below.. Regio;n A fue I-is .- mited hhtot assemblies having a nom~inal plana~r average W-23-5enric-hmeint of less tha or equal to 1.5 weight pern*t .I R gi *'A shall not contain any face adjacent fuel assemblies. Restriction. _s for non fi..i.

~flAI . *~ . %f*It~tfl1**

FuelLoaingPattern for Regilon lIA t 4.54 wt% U-235 Assembly Palisades Nuclear Plant 4.0-2 Amendment No. 8, 240, 236

4.3 Fuel Storage 4.3.1 Criticality (continued)

g. Region 16 is defino*da the Rogion I storago ra. c located in tho man spn pool with face adjacent fue thti iul surro-und-ed by empty face adjacent cels Rein 6fuli limited to theos assem-lies having a nominal plaRar average U-235 enrichment of Ios than or equaI tn 1.31 woight Percent.

Rogion IA Ioils that are diagonally adjacent to Regin 11 may contin uelassemblies provided conditions; of Sectionm 1.3.14l, 4.3.41lg.1 and 41.3.41lg.2 are met. Restrictons for non fiqssilo bearing compoen tsardecieinscin131ljbow Additional geo-metriccodtinso Region1Bae

1. Up to four face adja~ent fuel asseAmblies4 R inA; single contiguo*6s Frw are allowed as shoWn in the figures below.

All other facR-e adjacent cesIIQ shal.l be empty Or cotini non3fissij bearnl cmoensawdsrbe-nseto Aý l i - - I=_

ElmCIn IN 'rairnam. Tar rourII olr rawo~r nco* t~iacog*nT m~uuarnomau Ifl Un IEmpty cull

5 4.54 wt% U-235 Assembly S54.34 wt% U-235 Assembly Palisades Nuclear Plant 4.0-3 Amendment No. 236

4.3 Fuel Storage 4.3.1 Criticality (continued)

X *i f f , i f ....

-:2.'I fr'e, 'aGc alja*conT*uel a,;mqel-lHli 8 orMing an L I-- -- I-- - All pattern

]*__ ____

- .- I -ý I,. IN 1ý.Tý T-1 adjacent cells .urrounding tho two by toe bl*ck containing the L pa-.ern shall bh ompty-or cOF ntain

-- non-fissile bea*rin*g coME~enents as doscribed in soction 1.3.1lbl:

Reo.a.n 1 Pattorn. for Throo in an L I I I 4.54 -2 .... 7

_ - 4.54 wt% U-235 Assembly _.

64.04AO W206 okesmisiv 1 Palisades Nuclear Plant 4.0-4 Amendment No. 236

4.3 Fuel Storage 4.3.1 Criticality (continued)

h. Interface Requirements for the Main Spent Fuel Pool 1

_ ~ 4: Rnn'nR 1 f.eiri ark.L' that hatip Gn' p that nnnm19 in IniFln GVA~ 24O TnrouonQ1 um0 or tp Pain Mnent i-uoi 12on iare ania;ern ;o fuol rac in,1; Regio0Rn II Of the Main S*pnt Fuel Pool. Theso cells s~hall ho loAdodC with At least one =MWt ceAll ho~twoen~q

~-----....::i::: ::........................-

-*-j . i I l" " -- -- I* .... L L- LI-

d*nn n V+ it Vn t. L I ýtt n ,.+ + % ý tA .L ti^ 'Tn t contain, the nlver and insnectinn tMtnn T incarions are 11 10 it!Q Lean It Q~ Q~ W4 q Di Do~ and r" TrL-*4. n h11.*k II I- rtA..A I÷*.4 ...I 4 .

botwoen each fuel assombly withi, ;hn higoup

of cel*I.

Ro.i.n F= *.. d.fin.d A,; the Pnnion. I ctor.. rac.. Inn.At. d in the N 8 1 z Ri.

nrI+h +I+ n; Dnnrtn I C ch;l-l +',n~ab,+hr enlrtnf;* I,fin rB nn pattern as shown nthe figure below. This selectivo laqin*g Pattorn allowis for onAe fuel assombly having a nominal planar avore U 235 enrishmont of leos than or equal to 3.26 weight percent. This assembly shall bo placed in tho southwost cor lcoation of the rack. All 34 other allowed fuel locations in the figuro bolow are limited to a nominal planar avorage U.235 ohnrihent of loss than or o-qua! to 3.05 weight percent. The remaining fifteen cells shall be ernb:; and Region 1 E Allowod Fuol Storago Paottrn Nmrh )I EMPty-cGO"

- .5':i'J25Asm~

305 wig U'-235 Accombly Loainof a tinglo arromblyI Palisades Nuclear Plant 4.0-5 Amendment No. 236

4.3 Fuel Storage 4.3.1 Criticality (continued) j.Non-Fiseilo Boaring Comnpononts and- re-stric-tionsG are doefinod a6 INT PPTT]

1.on isslomaterial cOMPonont may be stored in any designated fuel location in Region 1A,11B, orF 1Ewihu es6t~Gti9R.

2. The following non fuel bearinRg com~ponents (NF=BC) may be storedd fa-e -adjacentto fuel in designated empty cellS in Region 1!A or 11B, except for interfarco locations described aboave in 4.1..1.h.

a. The gauge dummy assembly and the lead dummy assembly may be stored anywhere in Region 1!A or

h. An assembly comprised of up to 216-solid-stainles steel (SS) reds may be stoareddfc adjacent to fue ina designated emnpty cell as long as the NFBG is at least ten locations away from another NE=BO tha isface adja-ent to a fueRl assembly. L-ocations wi*thfin this NFB-C- assembly net containing SS red(-s) s-hall be left empty, Or G. A component com~prised primqarily of SS that displaces less-than 3-0 square inches of water i any horizontal plane within the active fuel region may be stored face adjacent to fuel, in a designate emnpty cell, as lon~g as the NFBCr is at least ten locations away 4fro another NIFBC that is face adjacent to a fue'Alaseb.

3. fisslo bearing c. Mpenbnts, shall

,on e*t be to*red in designated empty cells in Region 1E.

FIT 7ýI R -r b Palisades Nuclear Plant 4.0-6 Amendment No. 236

4.3 Fuel Storage 4.3.1 Criticality (continued) 4.3.1.2 The Region II fuel storage racks (See Figure B 3.7.16-1) are designed and shall be maintained with;

a. Fuel assemblies having maximum o planar average U-235 enrichment of 4.60 weight percent;

b. Keff < 1.0 if fully flooded with unborated water, which includes allowances for uncertainties as described in Section 9.11 of the FSAR.

c. p <5 0.95 if fully flooded with water borated to 850 ppm, which includes allowance for uncertainties as described in Section 9.11 of the FSAR.

d. A nominal 9.17 inch center to center distance between fuel assemblies; and

e. New or irradiated fuel assemblies which meet the ffaximum nominal planar average U-235, enrichment, burnup, and decay time requirements of Table 3.7.16-1.

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

a. Twent four unirradiated fuel assemblies having a maximum K aq7rn~a planar average U-235 enrichment of 4.95 weight percent, and stored in accordance with the pattern shown in Figure 4.3-1, or Thirty six unirradiated fuel assemblies having a maximum jiii-_7 j planar average U-235 enrichment of 4.05 weight percent, and stored in accordance with the pattern shown in Figure 4.3-1;

b. Keff < 0.95 when flooded with either full density or low density (optimum moderation) water including allowances for uncertainties as described in Section 9.11 of the FSAR.

Palisades Nuclear Plant 4.0-7 Amendment No. 4-18, 207, 236

FNSERT-11 f. Region 1A is defined as a subregion of the Region I storage racks located in the main spent fuel pool and is subject to the following restrictions. Fuel assemblies (or fissile bearing components) located in Region 1A shall be in a maximum of two-of-four checkerboard loading pattern of two fuel assemblies (or fissile bearing components) and two empty cells. Designated empty cells may contain non-fuel bearing components in accordance with Section 4.3.1.1 m.2. below; FN9ERT g. Region 1B is defined as a subregion of the Region I storage racks located in the main spent fuel pool and is subject to the following restrictions. Fuel assemblies (or fissile bearing components) located in Region 1B shall be in a maximum of three-of-four loading pattern consisting of three fuel assemblies (or fissile bearing components) and one empty cell. Fuel assemblies in Region 1B shall meet the enrichment dependent burnup restrictions listed in Table 3.7.16-2. Designated empty cells may contain non-fuel bearing components in accordance with Section 4.3.1.1 m.2. below;

h. Region 1C is defined as a subregion of the Region I storage racks located in the main spent fuel pool and is subject to the following restrictions. Fuel assemblies (or fissile bearing components) located in Region 1C may be in a maximum of four-of-four loading pattern with no required empty cells. Fuel assemblies in Region 1C shall meet the enrichment dependent burnup restrictions listed in Table 3.7.16-3; iNSERT 2 Interface requirements for the main spent fuel pool between Region 1A, 1B and 1C are as follows. Region 1A, 1B, and 1C can be distributed in Region I, in the main spent fuel pool, in any manner provided that any two-by-two grouping of storage cells and the assemblies in them correspond to the requirements of 4.3.1.1f., 4.3.1.1g., or 4.3.1.1h. above; NSERT 4, j. Region 1D is defined as a subregion the Region I storage rack located in the north tilt pit and is subject to the following restrictions. Fuel assemblies (or fissile bearing components) located in Region 1D may be in a maximum of three-of-four loading pattern consisting of three fuel assemblies (or fissile bearing components) and one empty cells. Fuel assemblies in Region 1D shall meet the enrichment dependent burnup restrictions listed in Table 3.7.16-4; MNSERT 5* k.. Region 1E is defined as a subregion the Region I storage rack located in the north tilt pit and is subject to the following restrictions. Fuel assemblies (or fissile bearing components) located in Region 1E may be in a maximum of four-of-four loading pattern with no required empty cells. Fuel assemblies in Region 1E shall meet the enrichment dependent burnup restrictions listed in Table 3.7.16-5;

FNSERT T Interface requirements for the north tilt pit between Region 1D and 1E'are as follows. Region 1D and 1E can be distributed in Region I in the north tilt pit in any manner provided that any two-by-two grouping of storage cells and the assemblies in them correspond to the requirements of 4.3.1 .lj. or 4.3.1.1 k. above; p

INSERT 8 m. Non-fissile bearing component restrictions are as follows:

1. Non-fissile material components may be stored in any designated fuel location in Region 1A, 1B, 1C, 1D, or 1E without restriction.

2. The following non-fuel bearing components (NFBC) may be stored face adjacent to fuel in any designated empty cell in Region 1A or lB.

(i) The gauge dummy assembly and the lead dummy assembly may be stored face adjacent to fuel in any designated empty cells with no minimum required separation distance.

(ii) A component comprised primarily of stainless steel that displaces less than 30 square inches of water in any plane within the active fuel region may be stored in any designated empty cell as long as the NFBC is at least ten locations away from any other NFBC that is in a designated empty cell, with the exception of 4.3.1.1 m.2.(i) above.

3. Control blades may be stored in both fueled and unfueled locations in Regions. 1D and 1E, with no limitation on the number.

ATTACHMENT 4 AREVA NP INC. AFFIDAVIT Three pages follow

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) ss.

COUNTY OF CAMPBELL

1. My name is George L. Pannell. I am Manager, Product Licensing, for AREVA NP Inc. (AREVA NP) and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to'ensure the proper application of these criteria.

3. I am familiar with the information contained inAREVA NP Document Number ANP-2858P, Revision 003, entitled "Palisades SFP Region 1 Criticality Evaluation with Burnup Credit," dated January 2011 and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is

requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA NP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b) and 6(c) above.

7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

p

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this day of 2011.

67"

Kathleen Ann Bennett NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 8/31/11 Reg. # 110864

v t e Letter Sequence Request
TAC:ME5419, (Approved, Closed)
Initiation Request, Request Acceptance Administration Withholding Request Acceptance Questions 25 August 2011 Answers 11 October 2011 Results Approval Other:
MONTHYEAR2011-03-07 [Table View]

PNP 2011-002, License Amendment Request for Spent Fuel Pool Region I Criticality

Text

SUBJECT:

Dear Sir or Madam:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

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PNP 2011-002, License Amendment Request for Spent Fuel Pool Region I Criticality

Text

SUBJECT:

Dear Sir or Madam:

1.0 DESCRIPTION

2.0 PROPOSED CHANGE

3.0 BACKGROUND

4.0 TECHNICAL ANALYSIS

6.0 ENVIRONMENTAL CONSIDERATION

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