ML11362A468

License Amendment, Issuance of Amendment No. 246 to Revise TS Sections 3.7.16 and 4.3 Regarding Fuel Storage Criticality

ML11362A468
Person / Time
Site:	Palisades
Issue date:	01/27/2012
From:	Mahesh Chawla Plant Licensing Branch III
To:	Entergy Operations
Chawla M
References
TAC ME5419
Download: ML11362A468 (23)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 January 27, 2012 Vice President, Operations Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043-9530

SUBJECT:

PALISADES NUCLEAR PLANT - ISSUANCE OF AMENDMENT RE: SPENT FUEL POOL REGION I CRITICALITY (TAC NO. ME5419)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 246 to Renewed Facility Operating License No. DPR-20 for the Palisades Nuclear Plant (PNP). The amendment consists of changes to the Technical Specifications in response to your application dated January 31, 2011, supplemented by your letter dated October 11,2011.

The amendment modifies the Spent Fuel Pool storage requirements in PNP Technical Specifications (TS) Section 3.7.16 by revising a limiting condition for operation for Region I fuel and non-fissile bearing component storage and by inserting tables containing spent fuel minimum burn-up for Regions 1B, 1C, 10, and 1E; and also modifies the Region I fuel storage criticality requirements, and design features in TS section 4.3, by describing revised requirements for Regions 1Band 1E and adding requirements for new Regions 1C and 1D.

A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, Mahesh L. Chawla, Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-255

Enclosures:

1. Amendment No. 246 to DPR-20

2. Safety Evaluation cc w/encls: Distribution via ListServ

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 ENTERGY NUCLEAR OPERATIONS, INC.

DOCKET NO. 50-255 PALISADES NUCLEAR PLANT AMENDMENT TO RENEWED FACILITY OPERATING LICENSE Amendment No. 246 License No. DPR-20

1. The U.S. Nuclear Regulatory Commission (the Commission) has found that:

A The application for amendment by Entergy Nuclear Operations, Inc. (the licensee),

dated January 31, 2011, supplemented by letter dated October 11, 2011, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B. The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public: and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; E. The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

- 2

2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to the license amendment and Paragraph 2.C.(2) of Renewed Facility Operating License No. DPR-20 is hereby amended to read as follows:

The Technical Specifications contained in Appendix A, as revised through Amendment No. 246, and the Environmental Protection Plan contained in Appendix B are hereby incorporated in the license. ENO shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

3. This license amendment is effective as of the date of issuance and shall be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION Shawn Williams, Acting Chief Plant licenSing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

Attachment:

Changes to the Facility Operating License and Technical Specifications Date of Issuance: January 27, 2012

ATTACHMENT TO LICENSE AMENDMENT NO. 246 RENEWED FACILITY OPERATING LICENSE NO. DPR-20 DOCKET NO. 50-255 Replace the following page of the Renewed Facility Operating License No. DPR-20 with the attached revised page. The changed area is identified by a marginal line.

REMOVE INSERT Page 3 Page 3 Replace the following pages of Appendix A, Technical Specifications, with the attached revised pages. The revised pages are identified by amendment number and contain marginal lines 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 Page 4.0-1 through 4.0-9 Page 4.0-1 through 4.0-6

-3 (1) Pursuant to Section 104b of the Act, as amended, and 10 CFR Part 50, "Licensing of Production and Utilization Facilities," (a) ENP to possess and use, and (b) ENO to possess, use and operate, the facility as a utilization facility at the designated location in Van Buren County, Michigan, in accordance with the procedures and limitation set forth in this license; (2) ENO, pursuant to the Act and 10 CFR Parts 40 and 70, to receive, possess, and use source and special nuclear material as reactor fuel, in accordance with the limitations for storage and amounts required for reactor operation, as described in the Updated Final Safety Analysis Report, as supplemented and amended; (3) ENO, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use byproduct, source, and special nuclear material as sealed sources for reactor startup, reactor instrumentation, radiation monitoring equipment calibration, and fission detectors in amounts as required; (4) ENO, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to receive, possess, and use in amounts as required any byproduct, source, or special nuclear material for sample analysis or instrument calibration, or associated with radioactive apparatus or components; and (5) ENO, pursuant to the Act and 10 CFR Parts 30, 40, and 70, to possess, but not separate, such byproduct and special nuclear materials as may be produced by the operations of the facility.

C. This renewed operating license shall be deemed to contain and is subject to the conditions specified in the Commission's regulations in 10 CFR Chapter I and is subject to all applicable provisions of the Act; to the rules, regulations, and orders of the Commission now or hereafter in effect; and is subject to the additional conditions specified or incorporated below:

(1) ENO is authorized to operate the facility at steady-state reactor core power levels not in excess of 2565.4 Megawatts thermal (100 percent rated power) in accordance with the conditions specified herein.

(2) The Technical Specifications contained in Appendix A, as revised through Amendment No. 246, and the Environmental Protection Plan contained in Appendix B are hereby incorporated in the license. ENO shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.

(3) ENO shall implement and maintain in effect all provisions of the approved fire protection program as described in the Final Safety Analysis Report for the facility and as approved in the SERs dated 09/01/78,03/19/80,02/10/81, 05/26/83,07/12/85,01/29/86, 12/03/87, and 05/19/89 and subject to the following provisions:

Renewed License No. DPR-20 Amendment No. 246

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 burn up of Tables 3.7.16-2, 3.7.16-3,3.7.16-4 or3.7.16-5; 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

---

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.

I 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. 489,~,~, 246

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 s 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 I 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 I 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 s 10%.

Palisades Nuclear Plant 3.7.16-2 Amendment No. ~,.:w+,~. 246

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 1B (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. 246

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 Bumup Burnup Average U-235 (GWD/MTU) (GWD/MTU)

Enrichment (Batches L (Batches A (Wt%) and later) through K) s1.35 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. 246

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 1D (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) s2.35 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. 246

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 1E (four-of-four loading configuration) of the North Tilt Pit Nominal Planar Burnup Burnup Average U-235 (GWDfMTU) (GWDfMTU)

Enrichment (Batches L (Batches A (Wt%) and later) through K) s1.48 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 I 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:s; 10%.

Palisades Nuclear Plant 3.7.16-6 Amendment No. 246

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. 89,29+, 6M, ~,

2J£,246

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 :s 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, 18, 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 18 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 18 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 18 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;

i. Interface requirements for the main spent fuel pool between Region 1A, 18, and 1C are as follows. Region 1A, 18, 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., 4.3.1.1 g., or 4.3.1.1 h. above; Palisades Nuclear Plant 4.0-2 Amendment No. 489,~,~, 246

Design Features 4.0 4.3 Fuel Storage 4.3.1 Criticality (continued)

j. Region 10 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 10 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 10 shall meet the enrichment dependent burnup restrictions listed in Table 3.7.16-4;

k. Region 1E 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; I. Interface requirements for the north tilt pit between Region 10 and 1 E are as follows. Region 10 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;

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, 10, 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 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 designated empty cell as long as the N FBC 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. 2*, 246

Desjgn 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 1E, with no limitation on the number.

4.3.1.2 The Region" 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, burn up, 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 4.04 Amendment No. -+8-9, 2G+-, 2J6, 246

Design Features 4.0 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. 2M. 236

Design Features 4.0 D0 D0 D0 D0 D0 D0 D D D D D D D0 D 0 D0 D0 D0 D0 CENTERLINE* LEGEND PATTERN REPEATS D 8 X 8 STEEL BOX BEAM o 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. 2W, 236

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT NO. 246 TO RENEWED FACILITY OPERATING LICENSE NO. DPR-20 ENTERGY NUCLEAR OPERATIONS, INC.

PALISADES NUCLEAR PLANT DOCKET NO. 50-255

1.0 INTRODUCTION

By letter dated January 31, 2011 (ADAMS Accession No. ML110380093), and supplemented by your letter dated October 11,2011 (ADAMS Accession No. ML112850216), Entergy Nuclear Operations (the licensee) requested revisions to Appendix A of the 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 amendment modifies the SFP storage requirements in Palisades Nuclear Plant TS Section 3.7.16 by revising a limiting condition for operation for Region I fuel and non-fissile bearing component storage and by inserting tables containing spent fuel minimum burn-up for Regions 1B, 1C, 1D, and 1E; and also modifies the Region I fuel storage criticality requirements, and design features in TS section 4.3, by describing revised requirements for Regions 1Band 1E and adding requirements for new Regions 1C and 1D.

The licensee submitted Areva Technical Report, ANP-2858P-003 documenting Palisades' criticality analysis.

The supplement dated October 11, 2011, provided additional information that clarified the application, did not expand the scope of the application as originally noticed, and did not change the U.S. Nuclear Regulatory Commission (NRC) staff's original proposed no significant hazards consideration determination as published in the Federal Register on May 10, 2011 (76 FR 27096).

2.0 REGULATORY EVALUATION

Title 10 of the Code of Federal Regulations (10 CFR) Part 50 Appendix A Criterion 62 requires that:

Criticality in the fuel storage and handling system shall be prevented by physical systems or processes, preferably by use of geometrically safe configurations.

Enclosure

-2 10 CFR 50.36(c)(4) requires that:

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.

10 CFR 50.68(b)(4) requires, in part, that:

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 NRC staff issued an internal memorandum on August 19, 1998, containing guidance for performing the review of SFP criticality analysis. This memorandum is known as the 'Kopp Letter,' after the author. The Kopp Letter provides guidance on salient aspects of a criticality analysis. The guidance is germane to boiling-water reactors and pressurized-water reactors, and to borated and unborated conditions. The staff used the Kopp Letter and the guidelines set forth in DSS-ISG-2010-001 and the appropriate NUREGs as guidance for the review of the current analysis.

3.0 TECHNICAL EVALUATION

Selection of Bounding Fuel Assembly Region I of the licensee's SFP is comprised of two rack designs, rack type 'c' and rack type 'E'.

The main region contains six rack type 'C' racks that can hold 372 fuel assemblies. Rack type

'E' can hold 50 fuel assembles and is located in the 'North Tilt Pit'. Region 1A, 1B, and 1C are rack type 'C' racks. Regions 1D and 1E are rack type 'E' racks.

The criticality analysis uses Batch X1, which contains an initial nominal planar average enrichment of 4.54 weight percent 235U, to bound the possible enrichments and different fuel types in the storage racks. Table 3-1 of ANP-2858P-003 shows the dimensions of the bounding model. In addition, Section 8.2 provides information that supports the bounding model to assemblies that contain legacy fuel from earlier fuel cycles.

The method used to determine the limiting assembly follows the guidance set forth in DSS-ISG-2010-01.

Depletion Analysis Section 4.0 of ANP-2858P-003 provides information on depletion calculations and burnup credit.

The methodology for depleting fuel assemblies in-reactor to support burnup credit in SFP criticality safety calculations includes the depletion of two dimensional unit assemblies as an infinite array in reactor core geometry with the use of the CASMO-3 code at the bounding core

-3 conditions. These reactor parameters include moderator temperature, fuel temperature, soluble boron concentration, and specific power and operating history. The licensee determines the axial burnup profiles to fit a 10 axial node model in KENO-V.a. The fuel assembly is then depleted by using CASMO-3 to the desired burnup value at each axial segment of the fuel.

From this depleted fuel, isotopic concentrations and cross sections, including those for the lumped fission products at cold rack geometry, are obtained. The CASMO-31umped fission products are then converted to an absorption-equivalent amount of Nd-145. Actinides and fission products are also included in the model.

Palisades Region I storage rack uses fixed spacing, burnup credit, and soluble boron credit to provide safe storage of discharged fuel assemblies. For the burn up credit calculations, the licensee considered operating history of the fuel, axial burnup distribution as a function of assembly-average burn up, measured burnup uncertainty, and a five percent uncertainty of reactivity decrement due to burnup.

The methodology used to determine the limiting assembly follows the guidance set forth in DSS-ISG-2010-01 and was found acceptable.

Criticality Analysis Sections 4.1 through 4.5 of ANP-2858P-003 provides analysis for each of the regions for the unborated and borated condition. The acceptance criteria for all regions with credit for soluble boron was K95/95 < 1.0 without soluble boron and ~ 0.95 with 850 parts per million (ppm) of soluble boron. For each region, the assembly misload condition was also calculated for the borated condition of 1350 ppm boron.

The interface requirements for ANP-2858P-003 are that each 2X2 array in any two regions must meet a keff value of less than 0.95, including all biases and uncertainities. The assemblies must also have at least the required burnup for the appropriate subregion. The interactions between Regions 1A, 1B, and 1C, the interactions between 1D and 1E, as well as the interactions between Region I and Region II are analyzed and were found to be acceptable.

The interactions between the C-rack and the E-rack loading zones did not require additional restrictions on loading patterns, due to the rack with the highest value of keff dominating during analysis. The same was shown to be true for Region I to Region II interfaces. Therefore, there are also no interface loading constraints on the Region I or II interfaces.

Section 4.0 of ANP-2858P-003 provides information on normal and abnormal conditions. The licensee has listed 8 abnormal conditions that were taken into consideration in the analysis.

These conditions include deboration of the pool, misplacement of a fresh fuel assembly within a cell that should be empty, replacement of a burnup credit assembly, drop of a fuel assembly outside the rack but adjacent to the rack, off-center fuel assembly, 'straight drop' accident, t-bone drop accident, and rack interactions. The deboration of the SFP is calculated at both o ppm as well as 850 ppm. The other abnormal conditions are evaluated at 1350 ppm. In all cases, normal and accident conditions provided a keff that is within regulations.

-4 Criticality Code Validation The analysis methodology uses the two-dimensional transport theory program CASMO-3 to calculate isotopic compositions at given burnups for the assemblies. KENO-V.a, a part of the SCALE 4.4a package, was used to calculate the keff of 100 critical systems. SCALE 4.4a is validated by using Haut Taux de Combustion critical experiment data.

The validation of SCALE Version 4.4a to perform criticality safety calculations was performed in accordance with NUREG/CR-6698, "Guide for Validation of Nuclear Criticality Safety Calculational Methodology," and follows the guidance set forth in DSS-ISG-2010-01.

4.0

SUMMARY

Currently, there is not a generically approved methodology for performing SFP criticality analysis. Therefore, the licensee must submit a plant-specific SFP criticality analysis that includes technically supported margins. The NRC staff reviewed the analysis to ensure that the assumptions made, both stated and unstated, are technically SUbstantiated. The NRC staff reviewed the application and supplemental information to determine whether the submittal provides reasonable assurance that the regulatory requirements will be met. As discussed earlier, the NRC staff finds that the licensee has provided the technical information needed for the NRC staff to complete its review of the license amendment request.

The licensee has demonstrated though its submittal that the methodologies used in its criticality analysis follows the guidelines set forth in DSS-ISG-2010-001 and the appropriate NUREGs.

After reviewing the licensee's original submittal and subsequent supplemental information, the NRC staff could determine reasonable assurance that the proposed TS changes would comply with the regulatory requirements.

5.0 STATE CONSULTATION

In accordance with the Commission's regulations, the Michigan State official was notified of the proposed issuance of the amendment. The Michigan State official had no comments.

6.0 ENVIRONMENTAL CONSIDERATION

The amendment changes a requirement with respect to the installation or use of a facility component located within the restricted area as defined in 10 CFR Part 20 or changes the surveillance requirements. The NRC staff has determined that the amendment involves no Significant increase in the amounts, and no significant change in the types, of any effluents that may be released offsite, and that there is no significant increase in individual or cumulative occupational radiation exposure. The Commission has previously issued a proposed finding that the amendment involves no significant hazards consideration and there has been no public comment on such finding (76 FR 27096). Accordingly, the amendment meets the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment need be prepared in connection with the issuance of the amendment.

- 5

7.0 REFERENCES

1. Code of Federal Regulations, Title 10, Part 50, Section 68, "Criticality Accident Requirements" .

2. L. I. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants." ADAMS Accession Number ML003728001, August 1998.

3. K. Wood, DSS-ISG-2010-1, "Staff Guidance Regarding the Nuclear Criticality Safety Analysis for Spent Fuel Pools". ADAMS Accession Number ML102220567, August 2010.

4. T. P. Kirwin, "License Amendment Request for Spent Fuel Pool Region I Criticality Palisades Nuclear Plant," Entergy Nuclear Operations, Inc., ADAMS Accession Number ML110380083, January 2011.

8.0 CONCLUSION

The Commission has concluded, based on the considerations discussed above, that: (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

Principal Contributor: Davida Cunanan, NRR Date: January 27, 2012

Vice President, Operations January 27,2012 Entergy Nuclear Operations, Inc.

Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, MI 49043-9530

SUBJECT:

PALISADES NUCLEAR PLANT - ISSUANCE OF AMENDMENT RE: SPENT FUEL POOL REGION I CRITICALITY (TAC NO. ME5419)

Dear Sir or Madam:

The U.S. Nuclear Regulatory Commission has issued the enclosed Amendment No. 246 to Renewed Facility Operating License No. DPR*20 for the Palisades Nuclear Plant (PNP). The amendment consists of changes to the Technical Specifications in response to your application dated January 31, 2011, supplemented by your letter dated October 11, 2011.

The amendment modifies the Spent Fuel Pool storage requirements in PNP Technical Specifications (TS) Section 3.7.16 by revising a limiting condition for operation for Region I fuel and non-fissile bearing component storage and by inserting tables containing spent fuel minimum burn-up for Regions 1B, 1C, 1D, and 1E; and also modifies the Region I fuel storage criticality requirements, and design features in TS section 4.3, by describing revised requirements for Regions 1Band 1E and adding requirements for new Regions 1C and 1D.

A copy of our related safety evaluation is also enclosed. The Notice of Issuance will be included in the Commission's biweekly Federal Register notice.

Sincerely, IRAJ Mahesh L. Chawla, Project Manager Plant Licensing Branch 111-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket No. 50-255

Enclosures:

1. Amendment No. 246 to DPR-20

2. Safety Evaluation cc w/encls: Distribution via ListServ DISTRI BUTION:

PUBLIC (ML11362A468 only) RidsNrrPMPalisades Resource RidsNrrDssStsb Resource RidsAcrsAcnw_MailCTR Resource RidsNrrDorlLpl3-1 Resource RidsOgcRp Resource RidsNrrDorlDpr Resource RidsNrrLABTully Resource RidsRgn3MailCenter Resource RidsNrrDssSrxb Resource RidsNrrDeEvib Resource DCunanan, NRR Package Accession No.: ML11362A471 Accesslon No. for Itt e er, amendmen, . t ary SE ML11362A468 t an d non-proprle E LPL3-1/PM LPL3-1/LA SRXB/BC DSS/STSB/BC OGCINLO LPL3-1/BC(A} LPL3-1/PM NAME MChawla BTuily AUlses RElliott LSUBIN SWilliams MChawla DATE 01/18/12 01/18/12 01/18/12 01/26/12 01/26/12 01127112 01127/12 OFFICIAL RECORD COPY