ML092450683

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License Amendment Request for Spent Fuel Pool Region I Criticality
ML092450683
Person / Time
Site: Palisades Entergy icon.png
Issue date: 09/01/2009
From: Schwartz C
Entergy Nuclear Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML092450683 (48)
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Text

Entergy Nuclear Operations, Inc.

_____ t ,Palisades Nuclear Plant 27780 Blue Star Memorial Highway Covert, Ml 49043 Tel 269 764 2000 Christopher J. Schwarz Site Vice President September 1, 2009 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 storage requirements in TS section 3.7.16 and criticality requirements for Region I spent fuel pool (SFP) and north tilt pit fuel storage racks, in TS section 4.3.

The criticality analysis supporting the proposed TS change for the Region I fuel storage racks reflects credit for fuel assembly burnup and soluble boron. 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 based on analyses to maintain Keff 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. The proposed change was evaluated for both normal operation and 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 Attachment 1. Attachment 1 also provides a detailed description of the proposed change, a background discussion, a technical analysis, and an environmental review consideration. Attachment 2 provides the revised TS pages reflecting the proposed changes. Attachment 3 provides the annotated TS pages showing the proposed

Document Control Desk Page 2 changes. Attachment 4 contains AREVA NP Inc. report Document No: ANP-2858-001, "Palisades SFP Region 1 Criticality Evaluation with Burnup Credit."

N To support fuel pool operations necessary to accommodate a full reactor core off load during the 2010 refueling outage, ENO requests approval of the proposed license amendment request by September 17, 2010, with the amendment being implemented within 60 days.

A copy of this request has been provided to the designated representative of the State of Michigan.

This letter contains no new commitments and no revision to existing commitments.

I declare under penalty of perjury that the foregoing is true and correct. Executed on September 1, 2009.

Sincerely, cjs/jse Attachment(s): 1. Description of Requested Changes

2. Revised Technical Specification Pages and Renewed Operating License Page Change Instructions
3. Mark-up of Technical Specification Pages
4. Palisades SFP Region 1 Criticality Evaluation with Burnup, Credit cc: Administrator, Region III, USNRC Project Manager, Palisades, USNRC Resident Inspector, Palisades, USNRC

ATTACHMENT 1 DESCRIPTION OF REQUESTED CHANGES

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 spent fuel pool and north tilt pit. 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 changes takes credit for fuel assembly burnup and soluble boron. In accordance with 10 CFR 50.68, Criticality accident requirements, the effective neutron multiplication factor (Keff) limits for Region I storage racks remain the same based on analyses to maintain Keff 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. The proposed change is evaluated for both normal operation and accident conditions.

The arrangement of the Region I and Region II storage racks in the spent fuel pool 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 spent fuel pool.

2.0 PROPOSED CHANGE

ENO proposes to modify (1) the spent fuel pool 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, and 1 E and (2) the Region I fuel storage criticality requirements in the TS design features in section 4.3 by describing revised requirements for Regions 1 B and 1E and adding requirements for new Region 1C. Requirements in 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, and 1 E.

The supporting analysis for Region 1B, 1C, and 1 E requirements in Attachment 4 has resulted in proposing restrictions on fuel assemblies that are unique and do not allow the use of exact verbiage from NUREG-1432, "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 19

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, and 1E 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, or 3.7.16-4; 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.

Page 2 of 19

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

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

Enrichment (Batches L (Batches A (Wt%) and later) through K) 2.50 0 1.0 2.60 0.81 1.81 2.80 2.43 3.43 3.00 4.05 5.05 3.20 5.67 6.67 3.40 7.28 8.28 3.60 8.90 9.90 3.80 10.52 11.52 4.00 12.13 13.13 4.20 13.75 14.75 4.40 15.37 16.37 4.54 16.50 17.50 (a) Linear interpolation between two consecutive points for nominal planar average U-235 enrichments between 2.50 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 19

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

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

Enrichment (Batches L (Batches A (Wt%) and later) through K) 1.80 0 1.0 2.40 7.07 8.07 2.60 9.43 10.43 2.80 11.78 12.78 3.00 14.15 15.15 3.20 16.50 17.50 3.40 18.98 19.98 3.60 21.45 22.45 3.80 23.93 24.93 4.00 26.40 27.40 4.20 28.84 29.84 4.40- 31.28 32.28 4.54 33.00 34.00 (a) Linear interpolation between two consecutive points for nominal planar average U-235 enrichments between 1.80 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 19

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

Spent Fuel Minimum Burnup Requirements for Storage in Region IE 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) 2.50 0 1.0 2.60 0.81 1.81 2.80 2.43 3.43 3.00 4.05' 5.05 3.20 5.67 6.67 3.40 7.28 8.28 3.60 8.90 9.90 3.80 10.52 11.52 4.00 12.13 13.13 4.20 13.75 14.75 4.40 15.37 16.37 4.54 16.50 17.50 (a) Linear interpolation between two consecutive points for nominal planar average U-235 enrichments between 2.50 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%.

TS 4.3.1.1 would be revised and read as follows:

"The Region I fuel storage racks (See Figure B 3.7.16-1) incorporating Regions 1A, 1B, 1C 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 subsection Region 1C within Region I. The requirements in Region 1A, 1B, 1C and 1 E are described below. The proposed changes within TS 4.3.1.1, Page 5 of 19

including those below, would result in revision of the specification that differs from the wording in NUREG-1432.

TS 4.3.1 .1a. would be revised and read as follows:

"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 1B, 1C, and 1E as determined in the analysis in Attachment 4. 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 as follows:

Changed the font for "Keff."

TS 4.3.1 .ld. would be revised as follows:

"Regions 1A, 1 B 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 will be inserted in the revised TS 4.3.1.le below.

TS 4.3.1.le. would be revised as follows:

"Region 1E has 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 the Region 1 E single Type E rack from TS 4.3.1.ld.

TS 4.3.1.lf. would be revised as follows:

"Region 1A is defined as a subsection 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 k.2. below;"

Page 6 of 19

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.

New TS 4.3.1.1.q would be revised as follows:

"Region 1B is defined as a subsection 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 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 4.3.1.1 k.2. below;"

The proposed change reflects the analysis in Attachment 4.

TS 4.3.1.1 h would be revised as follows:

"Region 1C is defined as a subsection 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 4. The existing interface requirements for the main spent fuel pool would be relocated to TS 4.3.1 .li.

TS 4.3.1.1 i would be revised 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 any manner provided that any 2-by-2 grouping of storage cells and the assemblies in them correspond to the requirements of 4.3.1.lf, 4.3.1.1 g or 4.3.1.1 h above;"

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 4 as related to Regions 1A, 1B, and 1C in the main pool. The existing Region 1 E storage requirements would be relocated to TS 4.3.1 .lj.

Page 7 of 19

TS 4.3.1.1 i. would be revised as follows:

"Region 1E is defined as 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 3.7.16-4;"

The proposed change would reflect the analysis in Attachment 4. Non-fissile bearing component requirements would be relocated to TS 4.3.1.1 k.

TS 4.3.1 .1 k. would be revised as follows:

"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 or 1E without restriction.
2. The following non-fuel bearing components (NFBC) may be stored in designated empty cells in Region 1A or lB.
a. The gauge dummy assembly and the lead dummy assembly may be stored anywhere in Region 1A or 1B.
b. A component comprised primarily of stainless steel that displaces less than 30 square inches of water in any horizontal plane within the active fuel region may be stored in a 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 k.2.a. above."

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

The proposed change would remove from TS 4.3.1 .1j.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 lB. This exception would be removed due to the analysis in Attachment 4.

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 TS 4.3.1.lj.3., which restricted non-fissile bearing components from being stored in designated empty cells in Region 1E. This Page 8 of 19

restriction would be removed because Region 1 E would no longer have designated empty cells.

TS 4.3.1.2a. would be revised as follows:

"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 4.3.1.2a. would insert "nominal" into the specification.

TS 4.3.1.2c. would be revised as follows:

The term "Keff" is reformatted for consistency.

TS 4.3.1.2e. would be revised as follows:

"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 as follows:

"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-10 Areal Density Gage for Evaluating Racks (BADGER) testing, of the Region I spent fuel pool (SFP) storage Page 9 of 19

racks, indicated that the neutron absorber material contained less boron-1 0 than assumed in the then spent fuel pool 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 4.3.1.1 b criticality requirements. TS 4.3.1.1 b required that Keff for Region I fuel racks be less than or equal to 0.95 if fully flooded with unborated water.

With soluble boron required to maintain a Keff less than or equal to 0.95 in the Region I fuel racks, PNP was no longer in compliance with the TS requirement or 10 CFR 50.68.

In accordance with NRC Administrative Letter 98-10, "Dispositioning of Technical Specifications that are Insufficient to Assure Plant Safety," compensatory measures were implemented.

The SFP contains storage racks that are designated as Region I and Region I1. The Region I storage racks contain Carborundum neutron absorber plates. 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 No. ML020590151 and ML020440048). Soluble boron, at 850 ppm, is required by TS 4.3.1.2c to maintain Keff less than or equal to 0.95 in the Region II storage racks when fully flooded with water.

ENO submitted a letter dated August 27, 2008 (ML082410132), describing four commitments and plans for a 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.lb. 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. 1.1 b and 10 CFR 50.68. This LAR proposed to revise spent fuel pool 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).

4.0 TECHNICAL ANALYSIS

Region I Criticality Evaluation AREVA NP Inc. report, Document No. ANP-2858-001, "Palisades SFP Region 1 Criticality Evaluation with Burnup Credit," (Attachment 4) provides the technical analysis for the proposed change to store fuel up to a maximum nominal planar average 4.54 weight percent U-235 in Regions 1B, 1C, and 1 E. ENO has reviewed and accepted this report.

Key elements of the report are as follows:

Page 10 of 19

Analysis Conservatisms

  • No credit is taken for intermediatespacer grids or end fittings.
  • No credit is taken for any boron in the Carborundum plates.
  • The maximum fuel enrichment tolerance of 0.05 weight percent is considered in the tolerance evaluation.
  • All fuel box outer steel walls are assumed bowed outward and filled with water (no voiding). For the tolerance calculations, the four-of-four loading configuration bounds the three-of-four loading configuration.

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

Results The results of the analysis determined that the Region I subregion 1B, 1C, 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.

Summary of Boron Dilution Evaluation Consumers Energy (the former owner and license holder) submitted on March 2, 2001, an amendment request for the spent fuel pool boron concentration. The amendment request and supplements provided the basis for NRC issuance of Amendment 207 to the Palisades Operating License, allowing changes to enrichment limits in the spent fuel pool. The amendment request provided a spent fuel pool boron dilution evaluation.

The evaluation has been reviewed and remains valid. In summary, available dilution sources were compiled and analyzed against the calculated dilution volumes to determine the potential of a spent fuel pool boron dilution event. For each dilution scenario, calculations were performed to define the dilution time for the spent fuel pool to reach 850 ppm.

The evaluation shows that a large volume of water (123,007 gallons) is necessary to dilute the spent fuel pool from the present TS limit of 1720 ppm to a soluble boron Page 11 of 19

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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. 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 be over boarded onto the pool deck and down the equipment hatch, elevator shaft, or the stair well, all of which are located within 4 to 10 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 amounts of water on the floor 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 aux 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 spent fuel pool drops below 850 ppm at 9.8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 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 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." The 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 move plans are developed by experienced and qualified reactor engineering personnel. Engineering Manual procedure EM-04-29, "Guidelines for Preparing Fuel Page 12 of 19

Movement Plans," is the governing document for preparation of fuel move plans. The procedure requires an independent review by another qualified reactor engineer and ensures that both the preparer and reviewer verify that the fuel move plan will 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 reactor engineer FHC. The Palisades 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 FHC. For Region 1E, which does not have cell coordinates in the SFHM computer, SOP-28 requires the 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 move plans.

The following human performance tools are used during fuel handling;

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.

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

Page 13 of 19

5.0 REGULATORY SAFETY ANALYSIS 5.1 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 racks when considering the presence of soluble boron in the pool water for criticality control. 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 spent fuel pool racks because the criticality analyses 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 spent fuel pool boron concentration and plant procedures 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 spent fuel pool when considering the presence of soluble boron in the spent fuel pool water for criticality control. The handling of fuel assemblies in the spent fuel is performed in borated water. The criticality analysis has shown the reactivity increase with a fuel assembly drop accident in both a vertical and horizontal orientation is bounded by the misloading accident. Therefore, the consequences of a fuel assembly drop accident in the spent fuel pool would not increase significantly due to the proposed change.

The spent fuel pool TS boron concentration requirement In TS 3.7.15 requires a minimum of 1720 ppm which bounds the analysis. Soluble boron has been maintained in the spent fuel pool water as required by TS and controlled by procedures. The criticality safety analyses for Region I Region 1A and Region II of the spent fuel pool 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 1B, 1C, and 1 E.

Page 14 of 19

Crediting soluble boron and burnup in the Region I Region 1 B, 1C, and 1E spent fuel pool criticality analysis would have no effect on normal pool operation and maintenance. Thus, there is no change to the probability or the consequences of the boron dilution event in the spent fuel pool.

Since soluble boron is maintained in the spent fuel pool water, implementation of the proposed changes would have no effect on the normal pool operation and maintenance. Also, since soluble boron is present in the spent fuel pool, a dilution event has always been a possibility. The loss of substantial amounts of soluble boron from the spent fuel pool 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 Region 1A and Region II spent fuel storage racks. In the unlikely event that soluble boron in the spent fuel pool is completely diluted, the fuel in Region I Region 1B, 1C, and 1 E of the spent fuel pool would remain subcritical by a design margin of at least 0.017 delta K, so the Keff of the fuel in these regions will remain below 1.0. 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 spent fuel pool water, when considering the presence of soluble boron in the pool water for subcriticality control, since a high concentration of soluble boron is always maintained in the spent fuel pool.

The criticality analyses documented in AREVA NP report ANP-2858-001 , "Palisades SFP Region 1 Criticality Evaluation with Burnup Credit," show, at a 95% probability and a 95% 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.

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 spent fuel pool have been analyzed in previous criticality evaluations, which are the bases for the present 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 Page 15 of 19

in the Region 1B storage rack, and 3.05 weight percent in the 1 E Region storage rack with the exception of one assembly in Region 1E 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 Region 1C would allow fuel enrichment to 4.54 weight percent 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 fuel manufacturer's procedures and plant handling procedures. 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 include procedures to develop the plans for fuel movement and operate the fuel handling equipment. These procedures include appropriate reviews and verifications to ensure TS requirements are maintained.

Furthermore, the existing TS contain limitations on the spent fuel pool 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 spent fuel pool water, implementation of the proposed changes would have no effect on the normal pool operation and maintenance. Since soluble boron is present in the spent fuel pool, a dilution event has always been a possibility. The loss of substantial amounts of soluble boron from the spent fuel pool was evaluated as part of the analysis in support of Amendment 207. That 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 as were used in previous analyses for the Region I Region 1A and Region II spent fuel storage racks. In the unlikely event that soluble boron in the spent fuel pool is completely diluted, the fuel in Region I Regions 1B, 1C, and 1E of the spent fuel pool would remain subcritical by a design margin of at least 0.017 delta K, so the Keff of the fuel in Region 1 would remain below 1.0 with burnup credit.

The combination of controls to prevent a mispositioned fuel assembly, ability to readily identify and correct a dilution event, and 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.

Page 16 of 19

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%

probability at a 95% confidence level, that the Keff of the Region I Region 1B, 1C, and 1E fuel storage racks in the spent fuel pool, 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. 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 spent fuel pool boron concentration is 1720 ppm, which provides assurance that the spent fuel pool would remain subcritical under normal, abnormal, or accident conditions.

The current analysis basis for the Region I Region 1A 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.

5.2 Applicable Regulatory Requirements/Criteria The SFP storage racks maintain fresh and irradiated assemblies in a safe storage condition. The federal code requirements in the Code of Federal Regulations, Title10, Part 50, Section 50.68 (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: "If credit is taken for soluble boron, the Keff of the spent fuel storage racks loaded with fuel of the maximum fuel assembly reactivity must not exceed 0.95, at a 95% probability, 95% confidence level, if flooded with borated water, and the Keff must remain below 1.0 (subcritical), at a 95%

probability, 95% confidence level, if flooded with unborated water."

Page 17 of 19

The evaluation in Attachment 4 provides results of analyses for Region I proposed Regions 1B, 1C, and 1 E that, with burnup credit, 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, 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 is shown to be less than 0.95 with a 95% probability at a 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 reactivity 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/N16.1-1975 and NRC letter of April 14, 1978, specifies that it shall require at least two unlikely, independent and concurrent events before a criticality accident is possible. This principle precludes the necessity of considering the simultaneous occurrence of multiple accident conditions.

The following applicable codes, standards, regulations and guidance, or pertinent sections thereof, were used in the analyses described in Attachment 4:

  • NUREG-0800 "Standard Review Plan," Section 9.1.1, "Criticality Safety of Fresh and Spent Fuel Storage and Handling," Revision 3, March 2007 0 USNRC 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)
  • 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 0 ANSI ANS-8.17-1984, "Criticality Safety Criteria for the Handling, Storage and Transportation of LWR Fuel Outside Reactors" 0 NUREG/CR-6698 "Guide for Validation of Nuclear Criticality Safety Methodology" Page 18 of 19

6.0 ENVIRONMENTAL CONSIDERATION

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.

7.0 REFERENCES

1. NRC letter issuing Amendment No. 207 to Palisades Facility Operating License dated February 26, 2002 (Accession Nos. ML020440048 & ML020590151)
2. Entergy Nuclear Operations, Inc. letter to the NRC, dated November 25, 2008, "License Amendment Request for Spent Fuel Pool Region I Criticality" (Accession Nos. ML083360619 and ML083360624)
3. NRC letter issuing Amendment 236 to Palisades Facility Operating License dated February 6, 2009 (Accession No. ML090160238)

Page 19 of 19

ATTACHMENT 2 REVISED TECHNICAL SPECIFICATION PAGES 3.7.16-1 through 3.7.16-5 and 4.0-1 through 4.0-5 AND RENEWED OPERATING LICENSE PAGE CHANGE INSTRUCTIONS 11 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 3.7.16-5 Pages 4.0-1 through 4.0-9 Pages 4.0-1 through 4.0-5

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 or 3.7.16-4; 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.

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. 41-8, 2-7, 2-6

Spent Fuel Pool Storage 3.7.16 TABLE 3.7.16-1 (D)aae 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. 4-89, 207, 2-36

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 of the Spent Fuel Pool Nominal Planar Burnup Burnup Average U-235 (GWD/MTU) (GWD/MTU)

Enrichment (Batches L (Batches A (Wt%) and later) through K) 2.50 0 1.0 2.60 0.81 1.81 2.80 2.43 3.43 3.00 4.05 5.05 3.20 5.67 6.67 3.40 7.28 8.28 3.60 8.90 9.90 3.80 10.52 11.52 4.00 12.13 13.13 4.20 13.75 14.75 4.40 15.37 16.37 4.54 16.50 17.50 (a) Linear interpolation between two consecutive points for nominal planar average U-235 enrichments between 2.50 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.

Spent Fuel Pool Storage 3.7.16 TABLE 3.7.16-3 (paqe 1 of 1)

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

Enrichment (Batches L (Batches A (Wt%) and later) through K) 1.80 0 1.0 2.40 7.07 8.07 2.60 9.43 10.43 2.80 11.78 12.78 3.00 14.15 15.15 3.20 16.50 17.50 3.40 18.98 19.98 3.60 21.45 22.45 3.80 23.93 24.93 4.00 26.40 27.40 4.20 28.84 29.84 4.40 31.28 32.28 4.54 33.00 34.00 (a) Linear interpolation between two consecutive points for nominal planar average U-235 enrichments between 1.80 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.

Spent Fuel Pool Storage 3.7.16 TABLE 3.7.16-4 (rnaae 1 of 1)

Spent Fuel Minimum Burnup Requirements for Storage in Region 1 E 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) 2.50 0 1.0 2.60 0.81 1.81 2.80 2.43 3.43 3.00 4.05 5.05 3.20 5.67 6.67 3.40 7.28 8.28 3.60 8.90 9.90 3.80 10.52 11.52 4.00 12.13 13.13 4.20 13.75 14.75 4.40 15.37 16.37 4.54 16.50 17.50 (a) Linear interpolation between two consecutive points for nominal planar average U-235 enrichments between 2.50 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.

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 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. 1-8, 207, 2-36

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 iffully flooded with water borated to 850 ppm, which includes allowances for uncertainties as described in Section 9.11 of the FSAR;
d. Region 1A, 1B and 1C has a nominal 10.25 inch center to center distance between fuel assemblies;
e. Region 1E has a nominal 11.25 inch by 10.69 inch center to center distance between fuel assemblies;
f. Region 1A is defined as a subsection 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 k.2. below;

g. Region 1 B is defined as a subsection 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 4.3.1.1 k.2. below;
h. Region 1C is defined as a subsection 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 main spent fuel pool between Region 1 A, 1 B and 1C are as follows. Region 1 A, 1 B and 1C can be distributed in Region I in any manner provided that any 2-by-2 grouping of storage cells and the assemblies in them correspond to the requirements of 4.3.1.1 f, 4.3.1 .1g or 4.3.1.1 h above; Palisades Nuclear Plant 4.0-2 Amendment No. 489, 2-7, 2-36

4.3 Fuel Storage 4.3.1 Criticality (continued)

j. Region 1 E is defined as 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 1 E 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 3.7.16-4;
k. Non-fissile bearing component restrictions are as follows:
1. Non-fissile material components may be stored in any designated fuel location in Region 1A, 1 B, 1C or 1E without restriction.
2. The following non-fuel bearing components (NFBC) may be stored in designated empty cells in Region 1 A or 1B.
a. The gauge dummy assembly and the lead dummy assembly may be stored anywhere in Region 1A or lB.
b. A component comprised primarily of stainless steel that displaces less than 30 square inches of water in any horizontal plane within the active fuel region may be stored in a 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 k.2.a. above.

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 Palisades Nuclear Plant, 4.0-3 Amendment No. 1-89, 2-W, 2-6

4.3 Fuel Storage 4.3.1 Criticality (continued)

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.

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 4.0-4 Amendment No. 4-89, 207, 246

DOOI-O 10r_ DO DODODODODODO CENTERLINE- LEGEND PATFERN REPEATS D 8X8STEEL 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 4.0-5 Amendment No. 4-89, 2-(, 2-36

ATTACHMENT 3 MARK-UP OF TECHNICAL SPECIFICATIONS PAGES (showing proposed changes; additions are highlighted and deletions are strikethrough) 14 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 4.3.1.1 ýid, as applicable, within the requirements of the maximum nominal planar average U-235 enrichment andburnupof Tables 3.7.16-2
  • 3.7.1-3 or 3.7,. 16-4; and
b. The combination of maximum nominal planar average U-235i 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

--- -------------- ------ --N OT E ---------------------------------------------------------

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 non-storage requirements. fissile bearing component in the spent fuel pool Palisades Nuclear Plant 3.7.16-1 Amendment No. 4-8@, 2-7, 236

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 Plana4 Average 523 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. 1-89, 207, 236

Spent Fuel Pool Storage 3.7.16 (TABLEE3.7.16-2 (pace 1:of !

&or::torace :in -Re~on 1BoTf 'the Spent,.'Fu-IPool verage 9-urnuOp 7B-r-fpur5u1)

Enichiment (GV&Wt DM O~ WiDMTU)'

t~l.)7 k(,atcheGsLi LBatchesA

,3.20'7j_

105m15 E603E9:7 EN~

W.40 ~ 53 ____

E4.563 16.=&5 0 .

(a) enicHmetd Etwepafi66 na n'o ,be ,t ~en 2-50ý)an yield aept~bo w~~it sfor noretds

'minfI f& n~Uh235 Cc-mp~'io ofhrfi Inlsen

~b) t~'b~i cetbe f~~vrg~u

~me~srem~tunp cqrtinjy'is <, O

Spent Fuel Pool Storage 3.7.16

[FABLE 3.7.16-3 (naae 1 of U

  • entflVNiiif,,imBurnup ,Requirem*ini*

ema9 ,a r1eme Sbrle ing~~6iIo h MO1a pn uIPo AyVerage U-235 Burniu, -fdfu Enri*bhW)enC aGnDWD/MTU' TU

__k lather),tLA (Bartchgs ater and th oul) fL1879 1i9.9 214 2.45 3.8 4.93D EN_____ 7.40

_ _ _ 9.8

(~)Lin~r4ntrpoýiM!ioR 15tween wogcnsecutive,ý poiristornomldffihi pIhaN M'iira, e,0-23,5 inrichtd

~etwqn V80and 4.54 will yedaoceptZ1er~jll ffp)i Qot~ipr~ison of norninal assembly veyr~aige utipý ubr bgthesin thetbeiaceabe mesuemn iun C'edai~isnk /o

Spent Fuel Pool Storage 3.7.16

~TABLE 3.7.16-4 (oaael of 1)!

~Sent ~el Minimunm Burnu Reqieet

~or~S~org~;inReIn lE of &ieNorthf~P nidhe nF" (GWJ/M#U~ ,(GW.D/MTU)'

____ (Ba~tchesLatheUA INER B LE 11 U

260 =0-81 E8:1]

YM6 5=7_ -

noJ 4.75136-4=40 15.3,31 W5i6.50 =.;V.50 richejnsbbewe 25an4.54,will yij~e ga e st_

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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 1A, 1 B, 1J and 1 E are designed and shall be maintained with:

a. [Ngew or irraldiae fuel assemblies having a maximum nominal planar average U-235 enrichment of 4.54 weight percentp iin Region !A, 4.31 weight percent in rogion i B, and 3.05 weight pcrcent in Rcegion 1E With tho oxGeption of one assembyi egion E.3, in .*.li bclow, hain aAmaimum noRminal planer average U 235 enrichment Pf 3.26 weight PeFGeRt.

Palisades Nuclear Plant 4.0-1 Amendment No. 1-89, 249, 224, 2-4, 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. Re~gion1 A, lB and 10 has a nominal 10.25 inch center to center distance between fuel assemblies with the exceptfion of the intgle Type erack iscehen has A nmial 11f.25 inch by 10.69 inch cen~ter to cepnter dista;nco betweenP fuelI ARsscMblIe.:
e. N. . .or fuel assemblies ~egion 1 E has a nommina Irradiated q 1by OW

',ln cehterd6 center stance between fuel assem blies-Region lIA is defined as the Region 1storage racks located in rN Ef. the- main spent fuel pool and are subject to the following restrictionR. All fuel located in Region !A shall be in a two of-four checkerboard loading pattFRn with emnpty cells as shown inth figure below. Rtegion lIA fuel is limited to those assemblies haVing a nominal planar average U-235 enrichmnent Of less thcan or equal to 4.54 weight percent. Region IlA shall not contain v

beaFrGIn Gomignents are decibdi section 4.3. 1.1m below:

J i w-uel eaueang Patewrn for ii Illon i Empty Gell

-. . 4.54 wt% U-235 Assembly -

Palisades Nuclear Plant 4.0-2 Amendment No. 1-88, 2-W, 236

4.3 Fuel Storage 4.3.1 Criticality (continued)

g. Rcegion 1B is defined as the Region I storage racks located i the main spent fuel pool with face adjacent fuel that is su-rrounded by empty face adjacent cells. Region 1B fuel is limited to thoýse assemblies haVing a nomninal planar average U-235 enrichment of less than or equal to 4.34 Weight percnt Region IAcells that are diagonally adjacent to Region 1B may contain fuel assemblies provided condlitions of Section 4.3..4lf-,

4.3..1.g.1 anýd 4.3..1.g.2 are mnet. Restrictions6 for non fissile bearing comAponents are descGribed in section 4.3.1.lj below.

Addtioalgeometric GdtG6G Region a;-

1. Up to four face adjacent fuel assemblies in a single contiguous row are allowed as shown in the figures below.

All other face adjacent cells shall be emnpty Or contain non fissile bearing com~ponents as described in sectionR 4.3.4.j beIGw-.

Reci in 1B Patterns for Four-or Fewer Face Adjacent Assemblies in a Rowr Emply cell

< 4.54 wt% U-235 Assembly

-4.340wt% U-235 Assembly Palisades Nuclear Plant 4.0-3 Amendment No. 236

4.3 Fuel Storage 4.3.1 Criticality (continued)

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i n~ II-Iiiirrmn I'* I'*13 nilr AI~4-~ chssri +hr fir. ,rnr h-irs All f~n A,-,-n

  • . *r...

t-*llc . .. . . . . . . . . . .l..*.*_*.._'-....; l -- -.

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cells r urroulding the two by-two blocGk contaiRi* g thc I. nattFRn r,.". ... ..

s~hall b eo oMt" or Gontan non fkisile born

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....... j ......

RegioIn 1 PatteFRn for Three inR anL I

i I WO Goll. 1 4 L

< 4.54 wt% U-235 Assemblv

!54.84 wt% Y 235, Ay 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 II-INSERT 3 -1o 14. RegiOn 1fuel raeks that haVe--...--.--- . Geni that oeeun'i IGcatiOns F=241--. m .

throiioh I')24 of the. -%! SMIZ~t FL0 Dnr~d 'nrn r' nn+

fupl r-,k-; in Ep*-on II of the MA*in-n,* nt Fr-,ul PonI These

4-1, .,.1,4 each fuel assembly within th is group Of cells.

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+

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............................................................ L, ii k i  ; 11 -u..

I 11 J 1tlWven eacn fuel assembly Witl ths goup of ceslls; Region 1E is defined as the Region 1storage rack locatod in th north tilt pit. Region 1 FEshall mainataifni the selective loading

[JNSERT 4iJ pa#e-*r as shown in the figur below. This SeIective - ladi*g patterm allows for one fuel assembly having a nominal planr averane U235 enRichmnent Of less thaR or equali to 3 266 eiAlh iIP[UP~11 I f11-' -P--~f-~[I1[)I 5~fl~1II nfl nI'z~~'r~~ I~ I~I~ ~S~I ~

loc~itin of the raek All 34l ehr+h-. alin e fuel IrGra+iGR-I l*

on lid fh,

  • I I*

iF; 11 i 1;  ;+ A *I*V* I I "]=J*

enrirch ent of les thaR OF 13CU2 tO 3 Or Wndnr'ht rrr~ ThP remaining fifteen cells shall be empty; and Region 1E Allowed Fuel Storage Pattorn N94h o

  • 3.05 wlo:-:235. A-. ombly Location Of A cinglo assombly 9 3.26 wt% U-235 Palisades Nuclear Plant 4.0-5 Amendment No. 236

4.3 Fuel Storage 4.3.1 Criticality (continued)

j. Non Fissile Bearing Com~ponents and restrictionsE are defined as fellews4 I !NSERT 5_
1. Non fissile material com~ponent may be stored in any deignated* fu locat-io.n in Region "A IB, or l E wAidthou', t
2. The follwing non fue l beaFi*ng**m9pnents (NFBC) m.ay*e*

storedd facr-e adjacent to fuel in designated emnpty cells in RegionI A or 1B, except for interfaGe Iocatios described above in R .3.1.1h.

a. The glauge dumnmy assembly and the lead dummy assembly may be stored annhere in Regio;n 1A or
b. An as6embly compr~iedof up to 216 solid staiRless steel (SS) rods may be toFred face adjacent to fuel inR a designated empty cell as long as the- NIFBC i at least ten locations away from anothe-r NIFBCr- that is face adjacent to a fuel assembly. Loc~ations within this NFBC ass.mbly not containing SS rod(&)

shall be left empty, or G. A comnponent comprised primarily of SS that displaces less than 30 square inches of water in any horizontal plane within the ac+"tive fuel regi may be stored face adjacent to fuel, in a designate empty cell, as long as the NIFBG iis at least ten locations away fromn ano-ther NIF-B that is face adja**nt to a fuel assemRy.

3.Non fi-ssfile bearing comnponents shall not be stored in desigRated emrpty cells in Region 1E.

I,:: S ,'N

,l~l' iEf,,,f-q,,,

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 Fonii 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. 44a 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 averagce 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 F6noi--ia- 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

, opjmia 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.

lýN FqlM`ATlONT`03W-AG'E~S40-8 AND

~ ~WLL~E RR~MNATEýP.

4".014j Palisades Nuclear Plant 4.0-7 Amendment No. 1-89, 207, 236

ýINSERT 11 f. Region 1A is defined as a subsection 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 k.2. below; INSERT 2 g. Region 1 B is defined as a subsection 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 4.3.1.1 k.2. below; INSEFRT3 h. Region 1C is defined as a subsection 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; IINSERT 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 any manner provided that any 2-by-2 grouping of storage cells and the assemblies in them correspond to the requirements of 4.3.1 .lf, 4.3.1 .lg or 4.3.1.1 h above;

________ J. Region 1 E is defined as 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 1 E 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 3.7.16-4;

INSERT 6 k. 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 or 1E without restriction.
2. The following non-fuel bearing components (NFBC) may be stored in designated empty cells in Region 1A or 1 B.
a. The gauge dummy assembly and the lead dummy assembly may be stored anywhere in Region 1A or lB.
b. A component comprised primarily of stainless steel that displaces less than 30 square inches of water in any horizontal plane within the active fuel region may be stored in a 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 k.2.a. above.
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