L-2002-221, Enclosure 3, St. Lucie Unit 1, Request for Amendment to License DPR-67, by Incorporating Attached Technical Specification (TS) Revisions

Enclosure 3, St. Lucie Unit 1, Request for Amendment to License DPR-67, by Incorporating Attached Technical Specification (TS) Revisions

ML023450403
Person / Time
Site:	Saint Lucie
Issue date:	11/25/2002
From:	Jernigan D Florida Power & Light Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-2002-221 HI-2022940
Download: ML023450403 (97)

Text

L-2002-221 St. Lucie Unit 1 Docket No. 50-335 Proposed License Amendment Spent Fuel Pool Soluble Boron Credit NON-PROPRIETARY HOLTEC LICENSE AMENDMENT REPORT (Bound Report)

.mmm.

HO LT E C INTER-NATIO NAL Holtec Center, 555 Lincoln Drive West, Marlton, NJ 08053 Telephone (856) 797-0900 Fax (856) 797-0909 ST. LUCIE UNIT 1 SPENT FUEL POOL STORAGE RACK BORAFLEX DEGRADATION REMEDY FOR FPL Holtec Report No: HI-2022940 Holtec Project No: 1237 Report Class: Safety Related

S esa..

HOLTEC INTERNA I IUNAL DOCUMENT ISSUANCE AND REVISION STATUSN ThflCiI MENT NAME: St. Lucie Unit 1 SFP Storage Rack Boraflex Degradation Remedy DOCUMENT NO.:

HI-2022940 CATEGORY: D-GENERIC PROJECT NO.:

1237

[

PROJECT SPECIFIC Rev.

Date Author's Rev.

Date Author's No. 2 Approved Initials VIR #

No.

Approved Initials VIR #

0 11/13/02 SP 118818 DOCUMENT CATEGORIZATION In accordance with the Holtec Quality Assurance Manual and associated Holtec Quality Procedures (HQPs), this document is categorized as a:

Calculation Package 3 (Per HQP 3.2)

F--

Design Criterion Document (Per HQP 3.4)

Technical Report (Per HQP 3.2)

(Such as a Licensing Report)

Design Specification (Per HQP 3.4)

E]

Other (Specify):

DOCUMENT FORMATTING The formatting of the contents of this document is in accordance with the instructions of HQP 3.2.

DECLARATION OF PROPRIETARY STATUS This document is labeled:

E]

Nonproprietary FD Holtec Proprietary 0

TOP SECRET Documents labeled TOP SECRET contain extremely valuable intellectual/commercial property of Holtec International.

They cannot be released to external organizations or entities without explicit approval of a company corporate officer.

The recipient of Holtec's proprietary or Top Secret document bears full and undivided responsibility to safeguard it against loss or duplication.

Notes

1.

This document has been subjected to review, verification and approval process set forth in the Holtec Quality Assurance Procedures Manual. Password controlled signatures of Holtec personnel who participated in the preparation, review, and QA validation of this document are saved in the N-drive of the company's network. The Validation Identifier Record (VIR) number is a unique six-digit random number that is generated by the computer after the specific revision of this document has undergone the required review and approval process, and the appropriate Holtec personnel have recorded their password-controlled electronic concurrence to the document.

2.

A revision to this document will be ordered by the Project Manager and carried out if any of its contents is materially affected during evolution of this project. The determination as to the need for revision will be made by the Project Manager with input from others, as deemed necessary by him.

3.

Revisions to Calculation Packages may be made by adding supplements to the document and replacing the "Table of Contents", the "Review and Certification" page and the "Revision Log".

TABLE OF CONTENTS

SUMMARY

OF REVISIONS Revision 0 contains the following pages COVER PAGE 1 ae DOCUMENT ISSUANCE AND REVISION STATUS 1 page

SUMMARY

OF REVISIONS 1 page TABLE OF CONTENTS 4 pages

1.0 INTRODUCTION

3 pages 2.0 GENERAL ARRANGEMENT 7 pages 3.0 SOLUBLE BORON DILUTION ACCIDENT I pages 4.0 CRITICALITY SAFETY ANALYSES 46 pages APPENDIX 4A 26 pages 5.0 THERMAL-HYDRAULIC CONSIDERATIONS 1 age 6.0 RACK SEISMIC/STRUCTURAL CONSIDERATIONS 1 page 7.0 MECHANICAL ACCIDENT ASSESSMENT 1 page 8.0 POOL STRUCTURE ASSESSMENT 1 page 9.0 RADIOLOGICAL CONSIDERATIONS 1 page 10.0 ENVIRONMENTAL COST/BENEFIT ASSESSMENT 1 page ITOTAL 96 pages Holtec Report HI-2022940 R-1 1253

TABLE OF CONTENTS

1.0 INTRODUCTION

1-1 1.1 References............................................................................................................

1-3 2.0 GENERAL ARRANGEM ENT...........................................................................

2-1 3.0 SOLUBLE BORON DILUTION ACCIDENT...................................................

3-1 4.0 CRITICALITY SAFETY ANALYSES..............................................................

4-1 4.1 Introduction and Summary..................................................................................

4-1 4.2 Acceptance Criteria..............................................................................................

4-5 4.3 Assumptions.........................................................................................................

4-5 4.4 Design and Input Data.........................................................................................

4-6 4.4.1 Fuel Assembly and Fuel Insert Specification.................................................

4-6 4.4.2 Holtec Storage Rack Specification......................................................................

4-6 4.5 M ethodology.....

..................................... 4-6 4.6 Analysis................................................................................................................

4-8 4.6.1 Bounding Fuel Assemblies..................................................................................

4-8 4.6.2 Pool W ater Temperature Effects.....................................................................

4-9 4.6.3 Uncertainties Due to M anufacturing Tolerances...............................................

4-10 4.6.4 Uncertainty in Depletion Calculations and Assembly Burnup.......................... 4-11 4.6.5 Isotopic Compositions................................................................................

4-11 4.6.6 Effects of Gadolinium........................................................................................

4-12 4.6.7 Effect of Distributed Enrichments...................................................................

4-12 4.6.8 Eccentric Fuel Assembly Positioning................................................................

4-13 4.6.9 Reactivity Effect of Axial Burnup and Enrichment Distribution...................... 4-13 4.6.10 B-10 Depletion in CEAs....................................................................................

4-14 4.6.11 Calculation of Burnup versus Enrichment Curves.............................................

4-15 4.6.12 Interfaces............................................................................................................

4-17 4.6.12.1 Region 1 to Region 1 Interfaces.........................................................................

4-18 4.6.12.2 Region 2 to Region 2 Interfaces.........................................................................

4-18 4.6.12.3 Region I to Region 2 Interface..........................................................................

4-18 4.6.12.4 Cells Facing the Pool W alls in Region 2 Racks................................................

4-19 4.6.12.5 Fresh Fuel in Region 2 Racks............................................................................

4-19 4.6.13 Soluble Boron Concentration for M aximum Keff of 0.95..................................

4-20 4.6.14 Abnormal and Accident Conditions...................................................................

4-20 4.6.14.1 Temperature and W ater Density Effects..........................................................

4-21 4.6.14.2 Dropped Assembly - Horizontal......................................................................

4-21 4.6.14.3 Dropped Assembly - Vertical............................................................................

4-21 4.6.14.4 Abnormal Location of a Fuel Assembly............................................................

4-22 4.6.14.4.1 M isloaded Fresh Fuel Assembly........................................................................

4-22 4.6.14.4.2 M islocated Fresh Fuel assembly.......................................................................

4-22 4.7 References.................................

............... 4-23 Appendix 4A...........................................................................................

"Benchmark Calculations" Holtec Report HI-2022940 i

1253

TABLE OF CONTENTS Total of 26 Pages including 6 figures 41 4'

4' 4'

4'

4.

4.

4.

4.

5 6

7 8

9 1

.1I Introduction and Summary

............................................... 4A-1 Effect.ofEnrichmen.....................................

4A-3 A.2 Effect of Enrichm ent.........................................................................................

4 A.3 Effect of l0B Loading........................................................................................

4A -4 A.4 Miscellaneous and Minor Parameters...............................................................

4A-5 A.4.1 Reflector M aterial and Spacings.......................................................................

4A-5 A.4.2 Fuel Pellet Diameter and Lattice Pitch.............................................................

4A-5 A..4.3 Soluble Boron Concentration Effects...............................................................

4A-5 A.5 M O X Fuel.........................................................................................................

4A -6 A.6 R eferences....................................................................................................

4A -7

.0 THERMAL-HYDRAULIC CONSIDERATIONS..............................................

5-1

.0 RACK SEISMIC/STRUCTURAL CONSIDERATIONS...................................

6-1

.0 MECHANICAL ACCIDENT ASSESSMENT...................................................

7-1

.0 POOL STRUCTURE ASSESSMENT................................................................

8-1

.0 RADIOLOGICAL CONSIDERATIONS............................................................

9-1

.0.0 ENVIRONMENTAL COST/BENEFIT ASSESSMENT..................................

10-1 Holtec Report HI-2022940 ii 1253

TABLE OF CONTENTS Tables 2.1 R ack D esign D ata............................................................................................................

2-2 2.2 Table of M odule D ata................................................................................................

2-3 2.3 Module Dimensions and Weights......

2-4 4.1.1 Minimum Burnup as a Function of Enrichment for Non-Blanketed Assemblies.

4-24 4.1.2 Minimum Burnup as a Function of Enrichment for Blanketed Assemblies

.......... 4-25 4.4.1 St. Lucie Unit 1 Fuel Assembly Specifications.............................................................

4-26 4.4.2 Control Element Assembly (CEA) Specifications.........................................................

4-27 4.4.3 Core Operating Parameters for Depletion Analyses......................................................

4-28 4.4.4 St. Lucie Unit 1 Fuel Rack Dimensions.........................................................................

4-29 4.6.1 Comparison of Kinf for Various Fuel assembly Types at Representative Fuel Conditions.................................................................................

4-30 4.6.2 Effect of Pool Water Temperature on Kinf for Fuel of 4.5wt% Enrichment and 0 Years Cooling Time at 0 ppm Soluble Boron..........................................................................

4-31 4.6.3 Reactivity Effect of Rack and Fuel Tolerances..........................

4-32 4.6.4 Enrichment and Cooling Time Combinations for Burnup versus Enrichment Curves. 4-33 4.6.5 Representative Calculation for each Case......................................................................

4-34 4.6.6 Results of Additional Selected Calculations for each Case...........................................

4-35 4.6.7 Soluble Boron Concentration for a Maximum keff Value of 0.95 under N orm al C onditions.........................................................................................................

4-36 4.6.8 Soluble Boron Concentration for a Maximum keff Value of 0.95 under A ccident C onditions.......................................................................................................

4-37 4A. 1 Summary of Criticality Benchmark Calculations.......................................

4A-9 thru 4A-13 4A.2 Comparison of MCNP4a and Keno5a Calculated Reactivities for V arious Enrichm ents.......................................................................................

4A -14 4A.3 MCNP4a Calculated Reactivities for Critical Experiments w ith N eutron A bsorbers..............................................................................................

4A -15 4A.4 Comparison of MCNP4a and KENO5a Calculated Reactivities for Various

" B 4 A -16 4A.5 Calculations for Critical Experiments with Thick Lead and Steel Reflectors............ 4A-17 4A.6 Calculations for Critical Experiments with Various Soluble Boron C oncentrations.......................................................................................................

4A -18 4A.7 Calculations for Critical Experiments with MOX Fuel..............................................

4A-19 Holtec Report HI-2022940 ii 1253

TABLE OF CONTENTS Figures 2.1 St. Lucie Unit 1 Fuel Pool Layout 2.2 Typical Rack Elevation - Region 1 2.3 Typical Rack Elevation - Region 2 4.4.1 Schematic View of Region 1 Cell (not to scale) 4.4.2 Schematic View of Region 2 Cell (not to scale) 4.6.1 Minimum Burnup as a Function of Initial Enrichment for Case 1, Low Reactivity Assemblies 4.6.2 Minimum Burnup as a Function of Initial Enrichment for Case 1, High Reactivity Assemblies 4.6.3 Minimum Burnup as a Function of Initial Enrichment for Case 2 4.6.4 Minimum Burnup as a Function of Initial Enrichment for Case 3 4.6.5 Minimum Burnup as a Function of Initial Enrichment for Case 4 4.6.6 Minimum Burnup as a Function of Initial Enrichment for Case 5 4.6.7 Schematic Configuration of the Calculational Model for Fresh Fuel Assemblies in Region 2 Racks for Inspection and Reconstitution 4A.1 MCNP Calculated k-eff Values for Various Values of the Spectral Index 4A.2 KENOSa Calculated k-eff Values for Various Values of the Spectral Index 4A.3 MCNP Calculated k-eff Values at Various U-235 Enrichments 4A.4 KENO5a Calculated k-eff Values at Various U-235 Enrichments 4A.5 Comparison of MCNP and KENO5a Calculations for Various Fuel Enrichments 4A.6 Comparison of MCNP and KENO5a Calculations for Various Boron-10 Areal Densities Holtec Report HI-2022940 iv 1253

1.0 INTRODUCTION

The two-unit St Lucie Plant (PSL) is located on Hutchinson Island in St. Lucie County, Florida, south of the city of Fort Pierce. The plant consists of two Combustion Engineering Pressurized Water Reactor (PWR) nuclear units. Unit 1 has been in commercial operation since 1976 and Unit 2 since 1983.

The existing Unit 1 spent fuel storage racks credit BoraflexrM as a neutron absorber to ensure subcriticality of the stored fuel. It is known [1] that Boraflex degrades during service conditions within the Spent Fuel Pool (SFP). The existing PSL Unit I Technical Specifications provide a description of the racks, including Boraflex, and include storage limitations based on reliance on the Boraflex. FPL seeks to re-license the storage racks in Unit 1 by crediting soluble boron in the pool water coupled with specific rules on fuel positioning to ensure subcriticality in lieu of crediting Boraflex as a neutron absorber. This report provides the design basis, analysis methodology, and results for the re-evaluation of the fuel storage racks in the St. Lucie Unit 1 Spent Fuel Pool without consideration of the Boraflex neutron absorber.

Neglecting Boraflex in the fuel storage criticality evaluations does not require any physical changes to the pool storage or rack configuration or require replacement of any storage racks. Safe storage will continue to be assured through rules on positioning fuel and by the neutron absorption provided by soluble boron in the SFP coolant. The soluble boron concentration is controlled by Technical Specifications.

St. Lucie Unit I has a current licensed storage capacity of 1,706 fuel assemblies. The existing high density racks were installed subsequent to a reracking analysis effort performed by Holtec in 1987.

Holtec licensing report HI-87105 [2] provides a detailed summary of the evaluations performed to support the re-licensing effort. Since there will not be any physical changes required to the pool, storage racks, or fuel contained within the racks, the analyzed configurations and results documented in the previous Holtec report remain valid with respect to structural, thermal-hydraulic, radiological and accident conditions. However, the racks have been re-evaluated for the criticality considerations discussed in detail herein.

Holtec Report HI-2022940 1-1 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

The methodologies employed to perform the rack criticality evaluations are a direct evolution of previous license applications. This report documents the analyses performed to demonstrate that the racks meet all governing requirements of the applicable codes and standards, in particular, I OCFR50.68(b)(4).

Section 2 of this report provides an abstract of the design and material information for the existing SFP storage racks. Section 3 provides an overview of the methodology used in an evaluation of postulated spent fuel pool boron dilution events and a summary of the results.

Section 4 provides a summary of the methods and results of the criticality evaluations performed for the spent fuel pool storage racks. The criticality safety analysis requires that the effective neutron multiplication factor (klff) be less than or equal to 0.95 with the storage racks fully loaded with fuel of the highest permissible reactivity and with the pool flooded with borated water at a temperature corresponding to the highest reactivity. In addition, the analysis requires that keff remains less than 1.0 following the assumed loss of soluble boron in the pool water, i.e. assuming unborated water in the spent fuel pool. The maximum calculated reactivities include a margin for uncertainty in reactivity calculations, including manufacturing tolerances, and are calculated with a 95% probability at a 95%

confidence level [5].

Thermal-hydraulic considerations are discussed in Section 5. Rack module structural analysis considerations are presented in Section 6. The structural qualification also requires that subcriticality of the stored fuel array be maintained under all postulated accident scenarios. The structural consequences of these postulated accidents are addressed in Section 7 of this report.

Section 8 establishes the continued adequacy of the SFP structure. The radiological considerations are documented in Section 9. Section 10 summarizes a cost/benefit and environmental assessment prepared by FPL to address the Boraflex degradation remediation proposal.

Holtec Report HI-2022940 1-2 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

All computer programs utilized to perform the criticality analyses documented in this report are benchmarked and verified. Holtec International has utilized these programs in numerous license applications over the past decade.

The analyses presented herein demonstrate that the Unit 1 SFP rack module arrays remain subcritical when soluble boron and specific rules on fuel assembly positioning are credited for reactivity control in lieu of Boraflex.

1.1 References

[1]

NRC Information Notice 95-38, Degradation of Boraflex Neutron Absorber in Spent Fuel Storage Tacks," September 1995.

[2]

Holtec International Report HI-87105, "Licensing Report for Reracking St. Lucie Unit 1 Fuel Pool," Revision 3, dated April 1987.

[31 Not Used.

[4]

American Society of Mechanical Engineers (ASME), Boiler & Pressure Vessel Code,Section III, 1989 Edition, Subsection NF, and Appendices.

[5]

M.G. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91, August 1963.

Holtec Report HI-2022940 1-3 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

2.0 GENERAL ARRANGEMENT The existing PSL Unit 1 high-density fuel racks consist of individual cells with 8.65 inch (nominal) square cross-section, each of which accommodates a single fuel assembly. A total of 1706 cells are arranged in 17 distinct modules of varying sizes of which four are Region I design with water gaps between cells, and thirteen are Region 2 design with no water gaps (see Table 2.1). Figure 2.1 shows the arrangement of the rack modules in the spent fuel pool.

The high density racks are engineered to achieve maximum protection against structural loadings (arising from ground motion, thermal stresses, etc.), the maximum number of available storage locations, and to maintain fuel assemblies in a subcritical array. Each rack module is equipped (see Figures 2.2 and 2.3) with girdle bars measuring 3/4 inches thick by 3-1 inches high. The girdle bar thickness on each rack ensures that a minimum gap of 1 -1/2 inches is maintained between modules. Table 2.1 gives the relevant design data for each region. The modules of the two regions are of eight different types. Tables 2.2 and 2.3 summarize the physical data for each module type.

Holtec Report HI-2022940 2-1 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Holtec Report HI-2022940 2-2 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 2.2 Table of Module Data Module Number of Number of Cells Number of Cells Total Number of Identification Modules in N-S Direction in E-W Direction Cells per Module Region 1 2

9 9

81 Al to A2 Region 1 9

10 90 BI to B2 Region 2 4

13 9

117 CI to C4 Region 2 3

13 8

104 Dl to D3 Region 2 2

11 8

88 El to E2 Region 2 1

12 8

96 F1 Region2 2

12 9

108 G1 to G2 Region 2 1

13 8

96 H1 Holtec Report HI-2022940 2-3 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 2.3 Module Dimensions and Weight Nominal Cross Section Dimensions (inches)

Estimated Dry Weight Module Identification N-S E-W per Module (lbs)

Region 1 90-1/4 90-1/4 26,700 Al to A2 Region 1 90-1/4 100-7/16 29,800 BD toB2 Region 2 115-11/16 80-1/16 24,100 Cl to C4 Region 2 115-11/16 71-3/16 21,500 DI to D3 Region 2 97-7/8 71-3/16 18,200 El to E2 Region 2 106-3/4 71-3/16 19,800 F1 Region 2 106-3/4 80-1/16 22,300 GI to G2 Region 2 115-11/16 71-3/16 19,800 H1I Holtec Report HI-2022940 2-4 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

F-igure 2.1; St. Lucie Unit I Fuel Pool Layout HI-2022940

Figure 2.2 TYPICAL RACK ELEVATIOQ-REGION 1 HI-2022940

-. 0) ý ol I

04-GIRDLE BAR Figure 2.3 TYPICAL RACK ELEVATION-REGION 2

HI-2022940

SOLUBLE BORON DILUTION ACCIDENT Florida Power and Light has prepared an evaluation that examines the potential for an inadvertent dilution of the St. Lucie Unit 1 spent fuel pool. The dilution scenarios presented in this report were developed after identifying the plant systems and components that interface with the Unit I fuel pool.

Periodic activities performed by plant operators that involve the spent fuel pool or systems interfacing with the spent fuel pool were also considered. Time periods required for a loss of reactivity margin to an effective neutron multiplication factor (keff) of 0.95 have been quantified.

Acceptance criteria are met if the evaluation concludes that sufficient time is available to detect and mitigate any credible dilution event before the kef design basis value is exceeded.

Typically, this analysis postulates the occurrence of multiple failures, as in the failure to correctly position a valve at the completion of an evolution coincident with a failure of an annunciator in the control room to alarm, or the failure of personnel to appropriately respond to an alarm. The evaluation did not consider the simultaneous occurrence of an inadvertent fuel pool dilution and a mis-positioned fuel assembly to be a credible scenario.

This analysis concludes that there are no credible spent fuel pool dilution events that could cause the soluble boron concentration to decrease from the assumed initial condition of 1720 ppm to a value such that keff equals 0.95.

The boron dilution analysis is provided as an enclosure to the license amendment request for soluble boron credit.

Holtec Report HI-2022940 3-1 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION 3.0

4.

Criticality Safety Analyses 4.1 Introduction and Summary Overview This section documents a new criticality safety analysis for the storage of PWR nuclear fuel in existing Region 1 & 2 style fuel storage racks installed in the spent fuel pool (SFP) at the St.

Lucie Unit 1 nuclear power plant. The spent fuel pool currently contains about 1350 fuel assemblies and is licensed to store up to 1706 assemblies. The analysis has been performed to qualify the existing racks from a criticality perspective under the assumption of a complete loss of the BoraflexTM neutron poison.

The existing spent fuel pool Region 1 & 2 style racks analyzed herein are used for the storage of irradiated fuel, and for fuel inspection, testing, and fuel reconstitution. This analysis excludes the new Region 1 cask pit rack, which is designed to accommodate fresh fuel and a portion of recently irradiated offload fuel.

The objective of the analysis is to qualify the existing SFP racks for the current spent fuel inventory and for future fuel discharges from Unit 1, without the need for additional neutron absorber inserts in the storage racks to offset an assumed loss of the Boraflex. This analysis credits the presence of soluble boron in the spent fuel pool, and the presence of control element assemblies (CEAs) placed in selected fuel assemblies. In order to achieve this analysis objective, it is necessary to group together fuel assemblies having similar reactivity characteristics and to establish different localized storage arrangements (i.e., checkerboard patterns) within the racks for assemblies with unique reactivity groupings.

Report No. HI-2022940 4-1 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Fuel Assembly Types Analyzed A total of seven fuel assembly types were developed to reflect'different reactivity groupings.

The following table lists each type by number, the type description used in this report, and its minimum burnup requirement based on an initial enrichment of 4.5 weight percent:

Fuel Storage Configurations Analyzed A total of five fuel storage configurations (cases) with different fuel assembly types were analyzed, as follows:

Case 1:

Case 2:

Case 3:

Case 4:

Case 5:

Region 2, Checkerboard of high and low reactivity fuel assemblies Region 1, Checkerboard of once burned and low reactivity fuel Region I, Checkerboard of twice burned and lower reactivity fuel Region 2, Checkerboard of high reactivity fuel assemblies with and without CEAs Region 2, Medium reactivity fuel assemblies only Burnup vs Enrichment Curves For each storage configuration above, and for each assembly type in a checkerboard array, the minimum required burnup has been determined as a function of the initial enrichment of the fuel.

Report No. HI-2022940 4-2 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

These functions, also termed burnup versus enrichment curves, are established as polynomial functions in the form of:

BU = A

• E+ B

• E + C with:

BU Burnup in GWD/MTU E

Initial Enrichment (wt %)

A,B,C Coefficients The current inventory of irradiated fuel at St. Lucie Unit 1 contains fuel assemblies with axial blankets, as well as fuel assemblies without axial blankets. Coefficients for all cases, for non blanketed and blanketed assemblies, and for all relevant post-irradiation cooling times are listed in Table 4.1.1 and 4.1.2, respectively.

Special Fuel Loading Rules A portion of the periphery of Region 2 storage racks faces the fuel pool wall. This part of the rack is analyzed for higher reactivity fuel, crediting the increased neutron leakage in this area.

Also, a designated area is established in Region 2 racks for fuel inspection and reconstitution, allowing a limited number of fresh fuel assemblies to be placed in a predefined pattern surrounded by empty cells. Reactivity effects of interfaces between the adjacent, potentially dissimilar storage arrangements have also been evaluated to assure that under all credible conditions, the fuel pool reactivity will not exceed the regulatory limit of 0.95. These conditions lead to following requirements:

1. Normally, each rack module will contain only one of the above listed configurations, i.e.,

Cases 1. 4, or 5 for a Region 2 rack, and Case 2 or 3 for a Region 1 rack. However, a rack module may contain more than one permissible configuration if an empty row is used to separate fuel stored in one configuration from fuel stored in a different configuration.

Report No. HI-2022940 4-3 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

2. Checkerboard patterns must be aligned across the gap between Region 1 rack modules, i.e., a high reactivity fuel assembly on one side of the gap must face a low reactivity assembly on the opposite side of the gap (i.e., "face-adjacent").

3. Checkerboard patterns need not be aligned across the gap between Region 2 rack modules, i.e., a high reactivity assembly on one side of the gap can face a high reactivity assembly on the opposite side of the gap.

4. The outer row of cells of Region 2 racks facing the pool wall or the cask pit wall is qualified to accept assemblies meeting the burnup and enrichment requirements for Case 4 (Type 3 fuel assemblies), and need not contain a CEA, regardless of the fuel assembly characteristics in the remainder of the rack.

5. Up to 4 (four) fresh assemblies or fuel rod baskets can be placed in a storage rack module having a Case 1 or Case 5 configuration, as long as each fresh assembly or rod basket directly faces 4 empty cells, and each of the diagonal cells is either empty or contains a Type 4, 6, or 7 assembly. Empty cells may contain non-actinide material, such as an empty fuel assembly skeleton, or other hardware, so long as the material occupies no more than 75% of the cell volume.

Analysis Results Analyses demonstrate that the effective neutron multiplication factor (keff) for all these cases is less than or equal to 0.95 when the storage racks are assumed to be fully loaded with fuel of the highest permissible reactivity and the pool is assumed to be flooded with borated water at a temperature corresponding to the highest reactivity. In addition, these analyses demonstrate that kf is less than 1.0 when the fuel pool is assumed to be flooded with unborated water. The maximum calculated values of the neutron multiplication factor include a margin for uncertainty in reactivity calculations, including manufacturing tolerances, and are calculated with a 95%

probability at a 95% confidence level [4.7.11.

Report No. HI-2022940 4-4 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

A minimum soluble boron concentration of 500 ppm must be maintained in the spent fuel pool to ensure that the effective neutron multiplication factor (keff) is less than or equal to 0.95 under all normal conditions.

Reactivity effects of accident conditions have also been evaluated. The most limiting accident condition involves the placement of a fresh fuel assembly between and directly adjacent to two other fresh fuel assemblies previously placed into a Region 2 rack module for inspection, testing or reconstitution. A minimum soluble boron concentration of 1090 ppm must be maintained in the spent fuel pool to ensure that the effective neutron multiplication factor (klff) is less than or equal to 0.95 under this condition.

St. Lucie Unit 1 Technical Specifications require that the fuel pool soluble boron concentration be maintained > 1720 ppm at all times.

4.2 ACCEPTANCE CRITERIA The objective of this analysis is to ensure that the effective neutron multiplication factor (kiff) is less than or equal to 0.95 with the storage racks fully loaded with fuel of the highest permissible reactivity and with the pool flooded with borated water at a temperature corresponding to the highest reactivity. In addition, the analysis shall ensure that for all storage configurations considered, keff is less than 1.0 when the fuel pool is assumed to be flooded with unborated water. The maximum calculated values of the neutron multiplication factor shall include a margin for uncertainty in reactivity calculations, including manufacturing tolerances, and are calculated with a 95% probability at a 95% confidence level [4.7.1].

4.3 ASSUMPTIONS To assure the true reactivity will always be less than the calculated reactivity, the following conservative design criteria and assumptions were employed:

Report No. HI-2022940 4-5 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

1) Moderator is borated or unborated water at a temperature that results in the highest reactivity, as determined by the analyses.

2) Neutron absorption in minor structural members is neglected, i.e., spacer grids are replaced by water.

3) Absorber rods present in some fuel assemblies are conservatively assumed to be fuel rods.

4) The effective multiplication factor of an infinite radial array of fuel assemblies or assembly patterns was used in the analyses, except for the assessment of peripheral and interface effects, and for certain abnormal/accident conditions where neutron leakage is inherent.

5) For the moderator temperature during fuel depletion, the highest core average value found at any axial location is used. This is conservative, since depletion with a higher moderator temperature results in higher fuel reactivity.

4.4 DESIGN AND INPUT DATA 4.4.1 Fuel Assembly and Fuel Insert Specification The design specifications for the Combustion Engineering (CE) and Framatome (FR) fuel assemblies, which were used for this analysis, are given in Table 4.4.1. Table 4.4.2 shows the specifications of the CEA fuel inserts used in the evaluations. Both tables also contain the applicable tolerances. The operating parameters used in the depletion analysis are given in Table 4.4.3.

4.4.2 Holtec Storage Rack Specification Specifications of the storage racks used in the criticality evaluations are summarized in Table 4.4.4 for the Region I and the Region 2 racks. Figures 4.4.1 and 4.4.2 show sketches of the cells for the Region 1 and Region 2 racks, respectively, indicating all relevant nominal dimensions.

4.5 METHODOLOGY The principal method for the criticality analysis of the storage racks is the three-dimensional Monte Carlo code MCNP4a [4.7.2]. MCNP4a is a continuous energy three-dimensional Monte Report No. HI-2022940 4-6 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Carlo code developed at the Los Alamos National Laboratory. MCNP4a was selected because it has been used previously and verified for criticality analyses and has all of the necessary features for this analysis. MCNP4a calculations used continuous energy cross-section data based on ENDF/B-V and ENDF/B-VI.

Benchmark calculations, presented in Appendix A, indicate a bias of 0.0009 with an uncertainty of+/- 0.0011 for MCNP4a, evaluated with a 95% probability at the 95% confidence level [4.7.1].

The calculations for this analysis utilized the same computer platform and cross-section libraries used for the benchmark calculations discussed in Appendix A.

The convergence of a Monte Carlo criticality problem is sensitive to the following parameters:

(1) number of histories per cycle, (2) the number of cycles skipped before averaging, (3) the total number of cycles and (4) the initial source distribution. The MCNP4a criticality output contains a great deal of useful information that may be used to determine the acceptability of the problem convergence. This information has been used in parametric studies to develop appropriate values for the aforementioned criticality parameters to be used in storage rack criticality calculations.

Based on these studies, the final calculations use a minimum of 10,000 histories per cycle, a minimum of 25 cycles were skipped before averaging, a minimum of 100 cycles were accumulated, and the initial source was specified as uniform over the fueled regions (assemblies).

Further, the output was reviewed to ensure that each calculation achieved acceptable convergence. These parameters represent an acceptable compromise between precision and computation time for design basis calculations.

Analyses of fuel depletion during St. Lucie Unit 1 power operation were performed with CASMO-4 (using the 70-group cross-section library), a two-dimensional multigroup transport theory code based on capture probabilities [4.7.3-5]. CASMO-4 is used to determine the isotopic composition of the spent fuel. In addition, the CASMO-4 calculations are restarted in the storage rack geometry to yield the two-dimensional infinite multiplication factor (kinf) for the storage rack. These restart calculations are used to determine the reactivity effect of fuel and rack Report No. HI-2022940 4-7 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

tolerances, and to perform various studies. For all calculations in the spent fuel pool racks, the Xe-135 concentration in the fuel is conservatively set to zero.

4.6 ANALYSIS This section describes the calculations that were used to determine the acceptable storage criteria for both the Region 1 and Region 2 style racks and it summarizes their results. In addition, this section discusses the postulated abnormal and accident conditions applicable to St. Lucie Unit 1 fuel pool storage.

Unless otherwise stated, all calculations assumed nominal characteristics for the fuel and the fuel storage cells. The effect of manufacturing tolerances is accounted for with a reactivity adjustment as discussed below.

All calculations are made using an explicit model of the fuel and storage cell geometry. The MCNP models contain a 2-by-2 array of cells surrounded by periodic boundary conditions. This represents an infinite checkerboard array. In CASMO, only a single cell is modeled. Since CASMO-4 is a two-dimensional code, the fuel assembly hardware above and below the active fuel length is not represented. The three-dimensional MCNP4a models that included axial leakage assumed 30 cm of water above and below the active fuel length. Additional models with more than four cells and with different boundary conditions were developed for MCNP to investigate the effect of rack module interfaces and to evaluate accident conditions. These models are discussed in the appropriate sections below.

4.6.1 Bounding Fuel Assemblies To determine the bounding assembly, calculations are performed for both assembly types listed in Table 4.4.1, and for both the upper bound and lower bound cladding thickness listed in that table. Further, calculations are performed for various enrichments, cooling times and burnups, Report No. HI-2022940 4-8 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

and for both Region 1 and Region 2 racks. Typical results are shown in Table 4.6.1, and demonstrate that for Region 1, the FR 14x14 assembly with a cladding thickness of 0.028 inches is the bounding assembly, whereas for Region 2, the CE 14x 14 assembly with a cladding thickness of 0.026 inches is the bounding assembly. These assemblies are therefore used in all further calculations for the respective rack types.

4.6.2 Pool Water Temperature Effects Pool water temperature effects on reactivity at 0 ppm soluble boron have been calculated with CASMO-4 and the results are presented in Table 4.6.2. The results in this table show that the spent fuel pool temperature coefficient of reactivity is positive for assemblies without CEAs (Region 1 and Region 2). In these cases, a higher temperature results in a higher reactivity, and the maximum normal pool temperature of 150 OF is therefore the bounding condition. However, for assemblies containing CEAs (only credited in Region 2 calculations), the temperature coefficient is negative, i.e. a lower temperature results in a higher reactivity. Consequently, all CASMO calculations for assemblies without CEAs are evaluated at 150 OF, whereas CASMO calculations for assemblies crediting CEAs are evaluated at 4 'C, which corresponds to the highest water density. For cases containing only assemblies without CEAs (cases 1, 2, 3 and 5),

the tolerances for 150 OF are applied. For Case 4, which uses a checkerboard of assemblies with and without CEAs, conservatively the maximum of the tolerance effect is applied. Pool water temperature effects on reactivity have also been evaluated in the presence of soluble boron; these effects are reported on Tables 4.6.7 and 4.6.8.

In MCNP, the Doppler treatment and cross-sections are valid only at 300K (27 °C). Therefore, a conservative Ak value is determined in CASMO-4 from 20 'C (68 OF) to 150 OF, and is included in the final ken calculation as a bias. Conservatively, the maximum value of this bias for each rack type shown in Table 4.6.2 is used in the final kff calculations. Although Case 4 contains assemblies with CEAs, which have a negative temperature coefficient of reactivity in the storage racks, a bias value derived from assemblies without CEAs is applied. This is conservative, since the reactivity effect of a temperature change between 20 'C and 150 °F for assemblies without Report No. HI-2022940 4-9 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

CEAs is larger than the reactivity effect of the temperature change from 20 'C to 4 'C for assemblies containing CEAs.

Fuel pool water temperatures exceeding 150 'F are considered accident conditions, and are discussed in Section 4.6.14.1.

4.6.3 Uncertainties Due to Manufacturing Tolerances In the calculation of the final k-infinity (kinf), the effect of manufacturing tolerances on reactivity must be included. CASMO-4 was used to perform these calculations. Factors considered include tolerances of the rack dimensions (see Table 4.4.4), tolerances of the fuel dimensions (see Table 4.4.1) and tolerances of the CEA specifications (see Table 4.4.2). In addition to the tolerances specified in these tables, an enrichment tolerance of 0.05 wt% is analyzed. As was done to identify the bounding assembly, calculations are performed for Region 1 and Region 2 racks, and CEAs, at a variety of enrichments, cooling times and burnups. The reference condition is the condition with nominal dimensions and properties. To determine the Ak associated with a specific manufacturing tolerance, the kinf calculated for the reference condition is compared to the kinf from a calculation with the tolerance included. All of the Ak values from the various tolerances represent independent effects and may be statistically combined (square root of the sum of the squares) to determine the final reactivity allowance for manufacturing tolerances. Only the positive Ak values (signifying increasing reactivity) were used in the statistical combination.

Table 4.6.3 shows the individual reactivity effects of tolerances, which when statistically combined, result in the highest total reactivity effect for Region 1, Region 2 and Region 2 containing assemblies with CEAs.

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4.6.4 Uncertainty in Depletion Calculations and Assembly Burnup CASMO-4 was used to perform the depletion calculations. Since critical experiment data with spent fuel is not available for determining the uncertainty in bumup-dependent reactivity calculations, an allowance for uncertainty in reactivity was assigned based upon other considerations. This analysis assumes the uncertainty in depletion calculations is less than or equal to 5% of the total reactivity decrement, and it assigns a burnup dependent uncertainty in reactivity for burnup calculations on this basis [4.7.6]. Additionally, the uncertainty of the assembly burnup value is 2.5 %. The reactivity effect of this uncertainty in burnup is determined and then these uncertainties are statistically combined with the other reactivity allowances to determine the maximum kerr for comparison with the limit of 0.95 for normal and accident conditions.

4.6.5 Isotopic Compositions To perform the criticality evaluation for spent fuel in MCNP, the isotopic composition of the fuel is calculated with the depletion code CASMO and then this isotopic composition is specified as input data to MCNP. Three isotopes or grouped isotopes in CASMO do not have a corresponding cross section in the MCNP cross section library. These are Pm-148M, and the lumped fission products LFP1 and LFP2. To account for these isotopes in the MCNP calculations, an equivalent amount of B-10 is calculated for each, and this B-10 amount is used in the MCNP calculation instead. The B-10 amount is specified through a multiplier on the atom density for each isotope, i.e. the B-10 atom density is calculated to be the Pm-148M / LFP1 /

LFP2 atom density calculated in CASMO multiplied by a constant factor. For each of the isotopes or isotope groups, a bounding factor is determined, and applied for the MCNP calculations.

The CASMO calculations to obtain the isotopic compositions for MCNP were performed generically, with one calculation for each rack type, enrichment and cooling time, using burnup Report No. HI-2022940 4-11 1237 SHADED AREAS DENOTE PROPRIETARY IN"FORMATION

increments of 2.5 GWD/MTU or less. The isotopic composition for any given burnup is then determined by linear interpolation.

4.6.6 Effect of Gadolinium At higher enrichments, assemblies contain up to 20 rods with Gadolinium (Gd) added to the fuel.

These rods are in specific locations around the control rod guide tubes. Rods containing Gadolinium also have a lower U-235 enrichment than do rods without Gd in the same assembly.

For a maximum assembly enrichment of 4.5 wt%, the highest U-235 enrichment in the rods with gadolinium will be approximately 2.6 wt%. A comparison of depletion calculations for fuel assemblies of equivalent enrichment, with and without Gd in these rods, shows that the assembly without Gd has a significantly higher reactivity for most conditions, and the presence of Gadolinium is therefore conservatively neglected in all further calculations.

4.6.7 Effect of Distributed Enrichments As noted in the previous paragraph, some assemblies contain fuel rods with lower enrichments around the guide tubes. As an example, an assembly with a maximum fuel rod enrichment of 4.5 wt% can have up to 20 fuel rods with enrichments as low as 2.6 wt% (see previous section). In addition, an assembly can have up to 40 fuel rods, not containing Gd, at a reduced enrichment to control radial power peaking. In an assembly with a maximum fuel rod enrichment of 4.5 wt%,

these additional 40 rods would typically be at an enrichment of 4.1 wt%. As a result, the planar average enrichment of a fuel assembly can be significantly lower than the maximum pellet enrichment. To show that it is acceptable to use the maximum planar average enrichment when determining the minimum required burnup from burnup vs. enrichment curves, calculations were performed for assemblies with radially distributed enrichments, and compared to calculations where all rods were set to a conservatively calculated planar average enrichment.

The calculations performed using the planar average enrichment result in slightly higher reactivity values than do the calculations performed using the actual assembly enrichment Report No. HI-2022940 4-12 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

distribution. It is therefore acceptable to use the maximum planar average enrichment of an assembly to determine the minimum required burnup.

4.6.8 Eccentric Fuel Assembly Positioning The fuel assembly is assumed to be normally located in the center of the storage rack cell.

Nevertheless, MCNP4a calculations assumed the fuel assemblies were positioned in the corner of the storage rack cell (a four-assembly cluster at closest approach). These calculations indicated that eccentric fuel positioning increases the reactivity of Region 1 by up to 0.0 127 delta-k, and decreases the reactivity in Region 2. For Region 1 calculations, the maximum difference in reactivity of 0.0127 delta-k is included in the uncertainties in the final k1f calculations.

4.6.9 Reactivity Effect of Axial Burnup and Enrichment Distribution Initially, fuel loaded into the reactor will bum with a slightly skewed cosine power distribution.

As power operation progresses, the axial burnup distribution will tend to flatten, becoming more highly burned in the central regions than in the upper and lower ends. At high burnup, the more reactive fuel near the ends of the fuel assembly (having less than average burnup) exists in a region of lower reactivity worth due to the ambient neutron leakage. Consequently, it would be expected that over most of their operating history, distributed burnup fuel assemblies would exhibit a slightly lower reactivity than that calculated for an assembly where all portions of fuel rods have the average burnup. As operation progresses, the distribution, to some extent, tends to be self-regulating as controlled by the axial power distribution, precluding the existence of large regions of significantly reduced burnup.

Generic analytic results of the axial bumup effect for assemblies without axial blankets were presented in [4.7.7]; these results are based upon comparisons of calculated and measured axial burnup distributions. These analyses confirm the minor and generally negative reactivity effect of the axially distributed burnup, which becomes positive at burnups greater than about 30 GWD/MTU. The trends observed [4.7.7] suggest the possibility of a small positive reactivity Report No. HI-2022940 4-13 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

effect above 30 GWD/MTU increasing to slightly over 1% Ak at 40 GWD/MTU. Since the required burnup for some enrichments and cases is greater than 30 GWD/MTU, the reactivity effect of the axially distributed burnup must be considered.

The St. Lucie Unit 1 plant also possesses fuel assemblies with natural (0.71 wt% 235U) and low enriched (2.6 wt% 235U) axial blankets on the ends, which effect the axial burnup distribution.

Calculations have been performed for the various axial burnup and enrichment variations, and the results were compared with a reference case, i.e. a case with an assumed axially constant burnup and enrichment. The results of this comparison indicate that, as expected, there is a positive reactivity effect from considering the axial burnup and enrichment distribution at higher burnup and cooling times for non-blanketed assemblies and for assemblies with enriched blankets. The effect of the axial burnup and enrichment distribution is considered in the calculations that establish burnup vs. enrichment curves, by conservatively performing calculations with both a uniform and non-uniform axial burnup and enrichment distribution, and selecting the higher of the resulting reactivity values as the representative reactivity value.

Enriched blankets are used in all blanketed calculations for conservatism.

In addition, the spent fuel pool contains Vessel Flux Reduction assemblies, which contain depleted uranium at an axially constant initial enrichment of about 0.3 wt%. Although the reactivity of such assemblies initially increases slightly with burnup, the reactivity is still significantly below the reactivity of all other permissible assemblies in the pool. Therefore, these assemblies can be placed in any location in the racks designated for a fuel assembly, and no further evaluation is required with these assemblies for any of the cases.

4.6.10 B-10 Depletion in CEAs CEAs are typically withdrawn from the active fuel region during full power operation of the core.

A significant depletion of the B-10 in the CEAs therefore does not occur, and the initial B-10 Report No. HI-2022940 4-14 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

loading of the CEA is used in the analyses. This lack of significant depletion of the B-10 in the CEAs is verified by measuring the CEA worth during startup physics testing after each reload.

However, to evaluate the effect of a conservative value for potential B-10 depletion, an additional calculation has been performed, wherein the B-10 concentration in the CEA was reduced by 30% in the lower 40 inches of each control rod finger. The results show that even this conservative reduction of the B-10 concentration does not lead to a significant difference in reactivity. It is therefore acceptable to model the CEA with the initial B-10 loading. Note that the dimensional tolerances of the CEA, including initial B-10 loading, were evaluated for their effects on reactivity and included in the total uncertainty calculation (Section 4.6.3).

4.6.11 Calculation of Burnup versus Enrichment Curves This analysis considers the following parameters and parameter combinations:

• Two fuel storage rack styles, with a total of five different fuel loading configurations.

• Fuel enrichments between 1.9 and 4.5 wt% 235U.

• Assemblies with and without axial blankets.

• Cooling times between 0 and 20 years Not all combinations of enrichment and cooling time are of practical relevance. The parameter combinations which are required to ensure that all current and future discharged fuel assemblies can be safely loaded into the racks are summarized in Table 4.6.4, and burnup vs. enrichment curves are determined for these parameter combinations. Prior analysis has indicated that it is necessary to account for the presence of the axial blankets in fuel assemblies in order to demonstrate that all fuel assemblies can be loaded into the racks without credit for Boraflex, since these blankets reduce the reactivity of certain high bumup fuel assemblies. Currently, the minimum enrichment of blanketed assemblies in the pool is about 3.55 wt%. However, it is possible that blanketed assemblies with a lower enrichment could be used in the future as a replacement for a damaged assembly unloaded from the core. The enrichment range for blanketed assemblies has therefore been extended down to 2.5 wt%, as shown in Table 4.6.4, to cover such assemblies. This assembly average enrichment is close to the current target Report No. HI-2022940 4-15 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

enrichment for the blanketed region of 2.6 wt%. Replacement assemblies with average enrichments below 2.5 wt% are bounded by the evaluations for non-blanketed assemblies, which were analyzed down to an assembly average enrichment of 1.9 wt%.

All calculations to establish and validate the burnup versus enrichment curves are performed as full three-dimensional criticality calculations considering the axial bumup distribution of each assembly in the model.

The coefficients of the burnup vs. enrichment curves for all conditions listed in Table 4.6.4 are shown in Table 4.1.1 for non-blanketed assemblies, and in Table 4.1.2 for assemblies containing axial blankets. These tables also provide the required minimum burnup for selected values of initial enrichment. Figures 4.6.1 through 4.6.6 present this information in a graphical form.

Fuel specifications for the checkerboard arrays have been chosen to maximize the calculated reactivity. The results of one representative calculation of the effective neutron multiplication factor (kerr) for each checkerboard storage arrangement is shown in Table 4.6.5 along with a tabulation of all biases and uncertainties applied to the calculated value prior to comparison with the 1.0 kerr limit. This table shows that the total addition for each case, i.e. the sum of all the applicable biases and uncertainties varied between 0.0177 Ak (Cases 1 and 5) and 0.0315 Ak (Case 2). Additional results from selected calculations for each case are listed in Table 4.6.6; these results identify the fuel specifications for each side of the checkerboard array and present the maximum kef (after application of biases and uncertainties) for the array as a whole when analyzed at these conditions. Note that Case 5 is also treated as a checkerboard pattern, but with the same burnup vs. enrichment curves for both assemblies in the pattern. The highest maximum keff of any case with any analyzed combination of fuel parameters is below the regulatory limit of 1.000 applicable when considering no soluble boron to be present in the fuel pool water. It should be noted that the calculations contain a significant amount of additional safety margin as a result of the underlying conservative assumptions, such as:

Report No. HI-2022940 4-16 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

• Maximum normal temperature in the pool

• Upper bound in-core moderator temperature

• Temperature bias and uncertainties calculated as maximum over the entire burnup /

enrichment / cooling time range

• No interpolation of cooling times allowed between loading curves.

The selection of the fuel specifications for the confirmatory calculations, and the embedded conservatisms will ensure that the actual reactivity of the pool, under the assumed accident condition of the loss of the soluble boron in the pool, will always be below 1.0. All burnup vs.

enrichment curves are therefore acceptable and result in reactivity values below the regulatory limit.

4.6.12 Interfaces In general, only one of the five fuel checkerboard arrangements is planned in each storage rack module. Therefore, only interfaces between the five cases across the inter-module gap need to be considered. However, additional special situations are permitted as follows:

"* Cells adjacent to the pool walls in Region 2 racks are qualified for a homogeneous loading of higher reactivity fuel, with the minimum burnup requirement as for Case 4.

"* Fresh fuel assemblies may be placed in certain Region 2 rack module locations for inspection, testing or reconstitution, provided they are placed face adjacent to vacant cells and any diagonally adjacent fuel assemblies meet certain criteria noted below.

"* A rack module may contain more than one permissible fuel storage configuration if an empty row is used to separate fuel in one configuration from fuel stored in another configuration.

This condition is bounded by the evaluations of the interfaces across the inter-module gaps, since an empty row is much wider than any of the inter-module gaps.

The results for all calculations of the interface reactivity effect discussed in the following subsections are statistically equivalent to (i.e. agree within two standard deviations), or lower than the result of the corresponding reference calculation. This agreement demonstrates that these interface configurations are acceptable.

Report No. HI-2022940 4-17 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

4.6.12.1 Region 1 to Region 1 Interfaces The four Region 1 rack modules are separated from each other by a gap of 1.5 inches, which is only slightly larger than the cell to cell gap within each rack module. It is therefore required that the checkerboard storage patterns between Region 1 rack modules are aligned, i.e. that high reactivity assemblies on the rack module boundary face low reactivity assemblies across the inter-module gap. Rack modules facing each other with the same checkerboard pattern are bounded by the calculation for the individual module, since the inter-module gap is slightly larger than the cell-to-cell gap within the racks. However, a calculation has been performed for two adjacent racks with differing checkerboard storage characteristics, i.e. a Case 2 arrangement in one rack module and Case 3 in the other module; results of this calculation show that this configuration is acceptable.

4.6.12.2 Region 2 to Region 2 Interfaces The bounding condition for Region 2 rack interfaces are at the corners of four rack modules, where each corner cell is occupied by a fuel assembly with the highest permissible reactivity for a Region 2 rack. This condition conservatively implies that checkerboard patterns in adjacent rack modules need not be aligned, i.e. it is permitted that higher reactivity assemblies face each other across rack module boundaries. The calculational model used to analyze this condition consists of a corner of a rack module with reflective boundary conditions on all four sides, thus effectively modeling an infinite array of racks with highest reactivity assemblies at all corners.

The results show that this configuration is acceptable.

4.6.12.3 Region 1 to Region 2 Interface Region 1 and Region 2 rack modules are separated by a gap of 1.5 inches. To model the interface with appropriate boundary conditions, a model was generated with 16 Region 2 cells on one side of the gap, and 14 Region 1 cells on the opposite side. The calculations show that this configuration is acceptable.

Report No. HI-2022940 4-18 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

4.6.12.4 Cells facing the Pool Wall in Region 2 Racks The peripheral row of Region 2 racks that face a pool wall or the cask pit wall is designated for storage of higher reactivity fuel assemblies, regardless of the checkerboard storage configuration used for the remainder of the rack. These higher reactivity assemblies correspond to the assemblies analyzed in Case 4, without CEAs. A number of variations for this interface have been analyzed, including:

"* Rack to wall distance of 5 inches and 6 inches

"* Stainless Steel liner thickness of 0.25 inches and 0.1875 inches

"* Concrete wall (6 feet) or water layer (6 inches) behind the liner

"* Side of the rack and comer of the rack All variations of these parameters result in a reactivity value that is statistically equivalent to or lower than the reference case reactivity, with a Case 1, Case 4 or Case 5 configuration in the remainder of the rack. Placing higher reactivity fuel assemblies (Case 4, without CEAs) on the periphery of Region 2 racks so that they face the pool wall or the cask pit wall is therefore acceptable.

4.6.12.5 Fresh Fuel in Region 2 Racks For fuel assembly inspection, testing and reconstitution, it is necessary to place up to 3 assemblies and a rod basket in close proximity to each other within a Region 2 rack module. As a bounding approach, these assemblies and the rod basket are modeled as fresh assemblies with an enrichment of 4.5 wt% in the calculations. To produce satisfactory results, it is required that the four cells face-adjacent to the cell with a fresh assembly be empty. Additionally, a fresh fuel assembly must not be placed in a cell diagonally adjacent to another cell containing a fresh assembly. However, it is acceptable to place spent fuel with the highest reactivity permitted for storage in Case 1 and Case 5 checkerboards in such a diagonal position. As a bounding approach, a configuration of 4 fresh assemblies is analyzed in an infinite array of a Case 1 checkerboard, with the fresh assemblies at the closest possible approach consistent with above Report No. HI-2022940 4-19 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

requirements. The pattern is shown in Figure 4.6.7. Analysis results confirm that this configuration is acceptable.

The empty cells were modeled with a water density of 25% of the normal water density. This assumption permits the placement of non-actinide material (i.e., hardware) in these cells as long as this non-fuel hardware does not occupy more than 75% of the cell volume.

No evaluation is performed considering the placement of fresh fuel assemblies within a Case 4 storage configuration, and this condition is therefore not permitted.

4.6.13 Soluble Boron Concentration for Maximum kef of 0.95 Calculations have been performed to determine the minimum soluble boron concentration in the spent fuel pool necessary to ensure that the reactivity of the fuel pool does not exceed 0.95. For each of the five fuel checkerboard storage configurations, calculations are performed at two soluble boron levels (100 ppm and 300 ppm for Region 1; 200 ppm and 500 ppm for Region 2),

and the soluble boron concentration necessary to satisfy the regulatory requirement is then determined by linear interpolation. A target of 0.94 is used for the maximum k~ff values, which is lower, i.e. more conservative, than the regulatory limit. Note that the presence of borated water in the fuel pool results in a slightly higher delta-temperature reactivity bias for the Region 2 racks than would be calculated assuming the presence of pure water. The highest minimum soluble boron concentration calculated is 443 ppm, calculated for Case 4. The details for this calculation are shown in Table 4.6.7. For added conservatism, a minimum value of 500 ppm is specified for compliance purposes, which is larger than the calculated value of 443 ppm.

4.6.14 Abnormal and Accident Conditions The effects on reactivity of credible abnormal and accident conditions are examined in this section. None of the abnormal or accident conditions that have been identified as credible cause the reactivity of St. Lucie Unit 1 fuel pool storage racks to exceed the limiting reactivity value of Report No. HI-2022940 4-20 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

1lff = 0.95, considering the presence of soluble boron. The double contingency principle of ANSI N16.1-1975 (and the USNRC letter of April 1978) specifies that it shall require at least two unlikely, independent and concurrent events to produce a criticality accident. This principle precludes the necessity of considering the simultaneous occurrence of multiple accident conditions.

4.6.14.1 Temperature and Water Density Effects The reactivity effect of fuel pool water temperatures exceeding 150 'F has been calculated.

Temperatures up to 248 'F (120 C) are evaluated, as are local boiling conditions with void percentages up to 20%. The maximum reactivity increase compared to 150 'F is 0.0303 Ak for Region 1 and 0.0 146 Ak for Region 2. It has been determined that a soluble boron concentration of 541 ppm is required to ensure a maximum k1ff of 0.95 is not exceeded under these conditions.

4.6.14.2 Dropped Assembly - Horizontal In the event a fuel assembly is dropped on top of a storage rack module, the dropped assembly will come to rest horizontally on top of the rack with a minimum separation distance of at least 12 inches from the active region of stored fuel. This distance is sufficient to preclude neutron coupling (i.e., an effectively infinite separation). The maximum expected deformation under seismic or accident conditions will not reduce this minimum spacing to less than 12 inches.

Consequently, the horizontal fuel assembly drop accident will not significantly increase reactivity in the fuel storage racks.

4.6.14.3 Dropped Assembly - Vertical It is also possible to vertically drop an assembly into a location occupied by another assembly.

Such a vertical impact would at most cause a small compression of the stored assembly, reducing the water-to-fuel ratio and thereby potentially increasing reactivity. However, the reactivity increase would be small compared to the reactivity increase created by the misloading of a fresh Report No. HI-2022940 4-21 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

assembly discussed in the following section. The vertical drop is therefore bounded by this misloading accident and no separate calculation is performed for the drop accident.

4.6.14.4 Abnormal Location of a Fuel Assembly 4.6.14.4.1 Misloaded Fresh Fuel Assembly The misplacement of a fresh unburned fuel assembly could, in the absence of soluble poison, result in exceeding the regulatory limit (1rf of 0.95). This could possibly occur if a fresh fuel assembly of the highest permissible enrichment (4.5 wt%) were to be inadvertently misloaded into a Region 2 storage cell intended to be empty (see Section 4.6.12.5), or into a cell intended to hold a low reactivity assembly (Case 4, assembly with CEA). The reactivity consequences of these situations were investigated and it was determined that the misloading of a fresh assembly into a cell intended to remain empty is the bounding condition. The evaluation of this case is shown in Table 4.6.8. To assure that the regulatory limit of 0.95 for the maximum kef is not exceeded under this condition, a soluble boron level of 1090 ppm in the spent fuel pool is required.

4.6.14.4.2 Mislocated Fresh Fuel Assembly The mislocation of a fresh unburned fuel assembly, i.e. the accidental placement of an assembly outside of the storage rack envelope but adjacent to other fuel assemblies, has also been considered. There is one area in the pool layout in which such an accident condition could be postulated to occur; this area is near the east wall of the pool in the cut-out of the Region 2 rack.

However, the size of this cut-out is such that the mislocated assembly can face no more than 2 rack walls; an assembly positioned here would face a substantial water thickness on its other two sides. This condition is therefore bounded by the fuel misloading accident discussed earlier, since the misloading accident has a fresh assembly surrounded by two other fresh assemblies inside the Region 2 rack.

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4.7 REFERENCES

1. M.G. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91, August 1963.

2. J.F. Briesmeister, Editor, "MCNP - A General Monte Carlo N-Particle Transport Code, Version 4A," LA-12625, Los Alamos National Laboratory (1993).

3. M. Edenius, K. Ekberg, B.H. Forss~n, and D. Knott, "CASMO-4 A Fuel Assembly Burnup Program User's Manual," Studsvik/SOA-95/1, Studsvik of America, Inc. and Studsvik Core Analysis AB (proprietary).

4. D. Knott, "CASMO-4 Benchmark Against Critical Experiments", SOA-94/13, Studsvik of America, Inc., (proprietary).

5. D. Knott, "CASMO-4 Benchmark Against MCNP," SOA-94/12, Studsvik of America, Inc., (proprietary).

6. L.I. Kopp, "Guidance on the Regulatory Requirements for Criticality Analysis of Fuel Storage at Light-Water Reactor Power Plants," NRC Memorandum from L. Kopp to T.

Collins, August 19, 1998.

7. S.E. Turner, "Uncertainty Analysis - Burnup Distributions", presented at the DOE/SANDIA Technical Meeting on Fuel Burnup Credit, Special Session, ANS/ENS Conference, Washington, D.C., November 2, 1988.

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Table 4.1.1 Minimum Burnup as a Function of Enrichment for Non-Blanketed Assemblies Case Cooling Time Case 1, Low Reactivity Case 1, High Reactivity 0 years 12 years 15 years 20 years 0 years 12 years 15 years 20 years Case 2, 0 years Low Reactivity Case 2, High Reactivity Case 3, Low Reactivity Case 3, High Reactivity Case 4 Case 5 5 years Coefficientst A

-0.65

-0.65

-0.43 0.12

-0.41

-0.54

-0.53

-0.46

-0.74

-0.56 B

20.08 17.76 16.25 12.90 17.00 16.22 15.86 15.11 17.49 15.64 C

-16.52

-15.58

-13.84

-9.61

-21.39

-20.63

-20.07

-18.80

-19.72

-17.65 Minimum Burnups (GWd/MTU) for various Enrichments 1.9%

2.5%

3.0%

3.8%

19.29 29.62 37.87 50.40 15.82 24.76 31.85 42.52 15.48 24.10 31.04 41.70 15.33 23.39 30.17 41.14 9.43 18.55 25.92 37.29 8.24 16.55 23.17 33.21 8.15 16.27 22.74 32.54 8.25 16.10 22.39 31.98 10.84 19.38 26.09 36.06 10.04 17.95 24.23 33.70 0 years 0.00 9.31

-24.39 0.00 0.00 F 10.99 0 years 0.00 10.97

-14.71 6.13 12.72 18.20 26.98 0 years 0.00 10.51

-22.35 0.00 3.93 9.18 17.59 0 years 0 years 12 years 15 years 0.00

-0.41 0.04 0.13 0.26 10.97

-14.71 6.13 17.70

-17.97 14.18 13.10

-12.56 12.47 12.38

-11.83 12.16 11.56

-11.16 11.74 12.72 18.20 26.98 23.72 31.44 43.37 20.44 27.10 37.80 19.93 26.48 37.09 19.37 25.86 36.52 Coefficients for polynomial Function: BU = A*E 2 + B*E + C with BU = Minimum Burnup in GWD/MTU; E = Initial Enrichment in wt% 235U; A, B, C = Coefficients Report No. HI-2022940 4-24 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.1.2 Minimum Burnup as a Function of Enrichment for Blanketed Assemblies Coefficients t B

19.25 17.40 16.32 16.00 16.45 18.97 16.54 14.73 Cooling Time 0 years 5 years 10 years 15 years 20 years 0 years 5 years 10 years 15 years 20 years 0 years 5 years C

-13.42

-12.03

-11.46

-11.73

-12.81

-22.54

-19.10

-16.49 Case Case 1, Low Reactivity Case 1, High Reactivity Case 2, Low Reactivity Case 2, High Reactivity Case 3, Low Reactivity Case 3, High Reactivity Case 4 Case 5 Minimum Burnups (GWd/MTU) for 2.5 %

29.46 26.97 25.22 24.08 23.57 18.76 17.63 16.77 16.28 16.15 19.38 17.95 various Enrichments 3.0%

3.5%

4.0%

4.5%

36.77 43.67 50.14 56.20 33.69 31.56 30.24 29.70 25.55 23.86 22.57 21.78 21.51 26.09 24.23 40.05 37.58 36.06 35.46 31.85 29.73 28.08 27.06 26.67 32.43 30.23 46.05 43.26 41.55 40.83 37.66 35.22 33.31 32.10 32.10 31.62 38.40 35.95 51.69 51.69 48.62 46.70

-45.8 45.83 42.98 40.35 38.25 36.92 36.37 44.00 41.39 0.00 9.31

-24.39 0.00 3.54 8.20 12.85 17.51 0 years 0.00 10.97

-14.71 12.72 18.20 23.69 29.17 34.66 0 years 0.00 10.51

-22.35 3.93 9.18 14.44 19.69 24.95 10.97

-14.71 12.72 14.23

-10.38 23.70 13.10

-9.24 22.26 12.70

-9.27 21.04 13.02

-10.48 20.07 14.08

-12.85 19.41 18.20 30.15 28.26 26.76 25.70 25.16 23.69 36.49 34.16 32.36 31.17 30.67 29.17 42.70 39.96 37.85 36.48 35.95 34.66 48.80 45.66 43.22 41.63 40.99 Coefficients for polynomial Function: BU = A*E 2 + B*E + C with BU = Minimum Burnup in GWD/MTU; E = Initial Enrichment in wt% 235U; A, B, C = Coefficients Report No. HI-2022940 4-25 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION 13.54 12.98 17.49 15.64 0 years A

-0.84

-0.72

-0.66

-0.67

-0.76

-0.98

-0.74

-0.57

-0.46

-0.41

-0.74

-0.56

-14.70

-13.74

-19.72

-17.65 0 years 0 years 5 years 10 years 15 years 20 years 0.00

-0.24

-0.20

-0.23

-0.32

-0.47

Table 4.4.1 St. Lucie Unit 1 Fuel Assembly Specifications Parameter Value Assembly type CE 14xl4 FR 14xl4 Rod Array Size 14x14 14x14 Rod Pitch, Inches 0.580 M 0.580 Maximum Active Fuel Length, Inches 136.7 136.7 Stack Density (g/cm 3) 10.05 10.30 Total Number of Fuel Rods 176 176 Fuel Rod Outer Diameter, Inches 0.440 0.440 Cladding Thickness, Inches 0.026 - 0.028 0.028 - 0.031 Cladding Material Zr-4 Zr-4 Maximum Pellet Diameter, Inches 0.3805 0.3770 Number of Guide Tubes 5

5 Guide Tube Outer Diameter, Inches 1.115 1.115 Guide Tube Wall Thickness, Inches 0.04 0.04 Guide Tube Material Zr-4 Zr-4 Report No. HI-2022940 4-26 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.4.2 Control Element Assembly (CEA) Specifications Report No. HI-202294 0 4-27 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.4.3 Core Operating Parameters for Depletion Analyses Parameter Value Soluble Boron Concentration, ppm 750 Reactor Specific Power, MW/MtU 31.2 Core Average Fuel Temperature, 'F 1275.1 Core Average Moderator Temperature at the Top of the 600.63 Active Fuel Region, 'F in-Core Fuel Assembly Pitch, Inches 8.18 Report No. HI-2022940 4-28 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.4.4 St. Lucie Unit I Fuel Rack Dimensions Parameter Value Region I Region 2 Cell ID 8.65 8.65 Wall Thickness 0.08 F 0.08 Cell Pitch 10.12 8.86 Boraflex Gap Thickness 0.075 0.05 Sheathing Thickness 0.02 nra Sheathing Width 7.5 n/a 1237 Report No. HI-2022940 4-29 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.6.1 Comparison of kinf for Various Fuel Assembly Types at Representative Fuel Conditions Report No. HI-2022940 4-30 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.6.2 Effect of Pool Water Temperature on kinf for Fuel of 4.5 wt% Enrichment and 0 Years Cooling Time at 0 ppm Soluble Boron.

Relative to 20 'C Report No. HI-2022940 4-31 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.6.3 Reactivity Effect of Rack and Fuel Tolerances Report No. HI-2022940 4-32 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.6.4:

Enrichment and Cooling Time Combinations for Burnup versus Enrichment Curves Case Non-Blanketed Assemblies Blanketed Assemblies Enrichment Cooling Time Enrichment Cooling Time 1

1.9-3.8 0, 12, 15,20 2.5 -4.5 0,5, 10, 15,20 2

1.9-3.8 0,

2.5-4.5 0,

5 (low reactivity 5 (low reactivity only) only) 3 1.9-3.8 0

2.5-4.5 0

4 1.9-3.8 0

2.5-4.5 0

5 1.9-3.8 0,12,15,20 2.5-4.5 0,5, 10, 15,20 Report No. HI-2022940 4-33 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.6.5 Representative Calculation for each Case Case 1

2 3

4 5

Region 2

1 1

2 2

Assembly 1 Enrichment 3.8 4.5 4.5 4.5 1.9 Burnup 50.4 44.0 34.7 34.7 14.1 Cooling Time 0

0 0

0 0

Assembly 2 Enrichment 3.8 4.5 3.5 3.5 1.9 Burnup 32.0 17.5 14.4 23.7 11.7 Cooling Time 20 0

0 0

20 Calculated k-eff 0.9785 0.9615 0.9636 0.9780 0.9788 Bias 0.0009 0.0009 0.0009 0.0009 0.0009 Temperature Correction 0.0037 0.0109 0.0109 0.0037 0.0037 Uncertainties Bias 0.0011 0.0011 0.0011 0.0011 0.0011 Calculationalt 0.0014 0.0014 0.0012 0.0012 0.0012 Eccentricity 0.0000 0.0127 0.0127 0.0000 0.0000 Tolerances 0.0130 0.0149 0.0149 0.0141 0.0130 Total Uncertainties 0.0131 0.0197 0.0196 0.0142 0.0131 Total Addition 0.0177 0.0315 0.0314 0.0188 0.0177 Maximum k-eff 0.9962 0.9930 0.9950 0.9968 0.9965 t Two times the standard deviation of the calculated kff Report No. HI-2022940 4-34 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.6.6 Results of Additional Selected Calculations for each Case

-Enr Enrichment in wt%; Bu = Burnup in GWD/MTU; Cool = Cooling Time in years; 1 & 2 = Two assemblies in Checkerboard Pattern Report No. HI-2022940 4-35 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.6.7 Soluble Boron Concentration for a Maximum keff Value of 0.95 under Normal Conditions.

Report No. HI-20229 4 0 4-36 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.6.8 Soluble Boron Concentration for a Maximum k1f Value of 0.95 under Accident Conditions.

Case m

1000 Reion 2

4.5% Fresh Fuel k-inf Calculated k-eff 1.1824 Assembly I Enrichment 1.9 Burnup 15.3 Cooling Time 20 Assembly 2 Enrichment 1.9 Burnup 9.4 Coolin2 Time 0

Calculated k-eff 0.9223 Bias 0.0009 Temperature Correction 0.0068 Uncertainties Bias 0.0011 Calculational 0.0014 Eccentricity 0.0000 Assembly Burnup 0.0075 Depletion 0.0130 Tolerances 0.0145 Total Uncertainties 0.0209 Total Addition 0.0286 Maximum k-eff 0.9509 Target k-max corresponding soluble boron level 1500 2

1.1098 1.9 15.3 20 1.9 9.4 0

0.8617 0.0009 0.0073 0.0011 0.0012 0.0000 0.0075 0.0124 0.0145 0.0206 0.0288 0.8905 0.94 1090 Report No. HI-2022940 4-37 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

7.5 I

0.075 8.65 1-10.12 Figure 4.4. 1:

Schematic View of Region 1 Cell (not to scale)

Report No. HI-2022940 4-38 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

8.86

'W4 Figure 4.4.2:

Schematic View of Region 2 Cell (not to scale)

Report No. HI-2022940 4-39 1237 SHADED AREAS DENOTE PROPRIETARY EINFORMATION

Case 1, Low Reactivity

--4--0 years 40.00 1~

40.00--W--12 years S15 years 20 years 3
0 0 years, Blankets S30.00

-o--5 years, Blankets SC Fig10 years, Blankets o

41-15 years, Blanketsi 20.00 20 years, Blankets 10.00 0.00 1.5 2

2.5 3

3.5 4.5 Initial Enrichment, wt%

Figure 4.6.1 Minimum Burnup as a Function of Initial Enrichment for Case 1, Low Reactivity Assemblies Report No. HI-2022940 4-40 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Case 1, High Reactivity 2

2.5 3

3.5 4

4.5 Initial Enrichment, wt%

60.00 50.00 40.00 0 30.00 d

I-1 20.00 10.00 0.00 5

Figure 4.6.2 Minimum Burnup as a Function of Initial Enrichment for Case 1, High Reactivity Assemblies Report No. HI-2022940 I - - 7 I L. I 4-41 SHADED AREAS DENOTE PROPRIETARY INFORMATION s 0 years

---

12 years 15 years 20 years

)K 0 years, Blankets

--6--5 years, Blankets

-- i-- 10 years, Blankets 15 years, Blankets 20 years, Blankets 1.5 I Z- - /

Case 2 60.00 50.00

• 40.00

= 30.00 E

S20.00 10.00 0.00 1.5 2

2.5 3

3.5 4

4.5 5

Initial Enrichment, wt%

Figure 4.6.3 Minimum Burnup as a Function of Initial Enrichment for Case 2 Report No. HI-2022940 4-42 SHADED AREAS DENOTE PROPRIETARY INFORMATION 4 Low Reactivity, 0 years Low Reactivity, 5 years High Reactivity IL.j I I Z.J/

Report No. HI-2022940

Case 3 E

E 1.5 2

2.5 3

3.5 4

4.5 Initial Enrichment, wt%

Low Reactivity LH:-:High Reactivity:

5 Figure 4.6.4 Minimum Burnup as a Function of Initial Enrichment for Case 3 Report No. HI-2022940 4-43 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Case 4 60.00-50.00

40.00 g 30.00 E

E S20.00 10.00 R 0.00 1.5 Figure 4.6.5 2

2.5 3

3.5 Initial Enrichment, wt%

Minimum Burnup as a Function of Initial Enrichment for Case 4 4

4.5 5

Report No. HI-2022940 I 

1L31 4-44 SHADED AREAS DENOTE PROPRIETARY INFORMATION IJ L./

Case 5 60.00 50.00 40.00 C

O 30.00 w

20.00 10.00 0.00 1.5 2

2.5 3

3.5 4

4.5 0 years 12 years 15 years 20 years

---

0 years, Blankets

-- o-5 years, Blankets 10 years, Blankets 15 years, Blankets 20 years, Blankets.]

5 Initial Enrichment, wt%

Figure 4.6.6 Minimum Burnup as a Function of Initial Enrichment for Case 5 Report No. HI-2022940 1,1 1 1 4-45 SHADED AREAS DENOTE PROPRIETARY INFORMATION ILO/

4 Fresh Assemblies (all separated by at least one empty cell, closest approach)

H High Reactivity L Low Reactivity F Fresh Assembly / Rod Basket Empty Cell

-.... Reflective Boundary Condition Figure 4.6.7 L

-LH.¸ H

L H

--L L

,-L H

H IL

-H ll L

H L

H L

H L

H L

H

_H H

H L

H LI H X F X F

H L

H H

H FXF H

L H IL H

SH H

L H

L H L H LHtHL H

'k-t-1"-"--E-----l-'-

Schematic Configuration of the Calculational Model for Fresh Assemblies in Region 2 Racks for Inspection and Reconstitution Report No. HI-2022940 4-46 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Appendix A Benchmark Calculations (total number of pages: 26 including this page)

Note: because this appendix was taken from a different report, the next page is labeled "Appendix 4A, Page 1".

Report No. HI-2022940 LL3 I 1237 Report No. HI-2022940

APPENDIX 4A: BENCHMARK CALCULATIONS 4A.1 INTRODUCTION AND

SUMMARY

Benchmark calculations have been made on selected critical experiments, chosen, in so far as possible, to bound the range of variables in the rack designs. Two independent methods of analysis were used, differing in cross section libraries and in the treatment of the cross sections. MCNP4a [4A. 1] is a continuous energy Monte Carlo code and KENO5a [4A.2]

uses group-dependent cross sections. For the KENO5a analyses reported here, the 238 group library was chosen, processed through the NITAWL-ll [4A.2] program to create a working library and to account for resonance self-shielding in uranium-238 (Nordheim integral treatment). The 238 group library was chosen to avoid or minimize the errorst (trends) that have been reported (e.g., [4A.3 through 4A.5]) for calculations with collapsed cross section sets.

In rack designs, the three most significant parameters affecting criticality are (1) the fuel enrichment, (2) the '°B loading in the neutron absorber, and (3) the lattice spacing (or water-gap thickness if a flux-trap design is used). Other parameters, within the normal range of rack and fuel designs, have a smaller effect, but are also included in the analyses.

Table 4A. 1 summarizes results of the benchmark calculations for all cases selected and analyzed, as referenced in the table. The effect of the major variables are discussed in subsequent sections below. It is important to note that there is obviously considerable overlap in parameters since it is not possible to vary a single parameter and maintain criticality; some other parameter or parameters must be concurrently varied to maintain criticality.

One possible way of representing the data is through a spectrum index that incorporates all of the variations in parameters. KENO5a computes and prints the "energy of the average lethargy causing fission" (EALF). In MCNP4a, by utilizing the tally option with the identical 238-group energy structure as in KENO5a, the number of fissions in each group may be collected and the EALF determined (post-processing).

t Small but observable trends (errors) have been reported for calculations with the 27-group and 44-group collapsed libraries. These errors are probably due to the use of a single collapsing spectrum when the spectrum should be different for the various cases analyzed, as evidenced by the spectrum indices.

Appendix 4A, Page 1

Figures 4A. 1 and 4A.2 show the calculated klf for the benchmark critical experiments as a function of the EALF for MCNP4a and KENO5a, respectively (UO2 fuel only). The scatter in the data (even for comparatively minor variation in critical parameters) represents experimental errort in performing the critical experiments within each laboratory, as well as between the various testing laboratories. The B&W critical experiments show a larger experimental error than the PNL criticals. This would be expected since the B&W criticals encompass a greater range of critical parameters than the PNL criticals.

Linear regression analysis of the data in Figures 4A. 1 and 4A.2 show that there are no trends, as evidenced by very low values of the correlation coefficient (0.13 for MCNP4a and 0.21 for KENO5a). The total bias (systematic error, or mean of the deviation from a kff of exactly 1.000) for the two methods of analysis are shown in the table below.

Calculational Bias of MCNP4a and KEN05a MCNP4a 0.0009 +0.0011 KENO5a-0.0030+/-0.002 The bias and standard error of the bias were derived directly from the calculated klf values in Table 4A. 1 using the following equationstt, with the standard error multiplied by the one-sided K-factor for 95 % probability at the 95 % confidence level from NBS Handbook 91 [4A. 18] (for the number of cases analyzed, the K-factor is -2.05 or slightly more than 2).

k=1 j ki (4A.1) n A classical example of experimental error is the corrected enrichment in the PNL experiments, first as an addendum to the initial report and, secondly, by revised values in subsequent reports for the same fuel rods.

These equations may be found in any standard text on statistics, for example, reference

[4A.6] (or the MCNP4a manual) and is the same methodology used in MCNP4a and in KENO5a.

Appendix 4A, Page 2

n 2 ________-___,___) ___

(4A.2) n (n-1)

Bias =(1-k) ý K o-(4A.3) where k, are the calculated reactivities of n critical experiments; or is the unbiased estimator of the standard deviation of the mean (also called the standard error of the bias (mean)); K is the one-sided multiplier for 95 % probability at the 95 % confidence level (NBS Handbook 91 [4A.18]).

Formula 4.A.3 islased on the methodology of the National Bureau of.$andards (now NIST) and is used to calculate the values presented on page 4.A-2. The first portion of the equation, ( 1-k ), is the actual bias which is added to the MCNP4a and KENO5a results.

The second term, Koj, is the uncertainty or standard error associated with the bias. The K values used were obtained from the National Bureau of Standards Handbook 91 and are for one-sided statistical tolerance limits for 95 % probability at the 95 % confidence level. The actual K values for the 56 critical experiments evaluated with MCNP4a and the 53 critical experiments evaluated with KENO5a are 2.04 and 2.05, respectively.

The bias values are used to evaluate the maximum k1 values for the rack designs.

KENO5a has a slightly larger systematic-error than MCNP4a, but both result in greater precision than published data [4A.3 through 4A.5] would indicate for collapsed cross section sets in KENO5a (SCALE) calculations.

4A.2 Effect of Enrichment The benchmark critical experiments include those with enrichments ranging from 2.46 w/o to 5.74 w/o and therefore span the enrichment range for rack designs. Figures 4A.3 and 4A.4 show the calculated krf values (Table 4A. 1) as a function of the fuel enrichment reported for the critical experiments. Linear regression analyses for these data confirms that there are no trends, as indicated by low values of the correlation coefficients (0.03 for MCNP4a and 0.38 for KENO5a). Thus, there are no corrections to the bias for the various enrichments.

Appendix 4A, Page 3

As further confirmation of the absence of any trends with enrichment, a typical configuration was calculated with both MCNP4a and KENO5a for various enrichments.

The cross-comparison of calculations with codes of comparable sophistication is suggested in Reg. Guide 3.41. Results of this comparison, shown in Table 4A.2 and Figure 4A.5, confirm no significant difference in the calculated values of k~f for the two independent codes as evidenced by the 450 slope of the curve. Since it is very unlikely that two independent methods of analysis would be subject to the same error, this comparison is considered confirmation of the absence of an enrichment effect (trend) in the bias.

4A.3 Effect of 10B Loading Several laboratories have performed critical experiments with a variety of thin absorber panels similar to the Boral panels in the rack designs. Of these critical experiments, those performed by B&W are the most representative of the rack designs. PNL has also made some measurements with absorber plates, but, with one exception (a flux-trap experiment),

the reactivity worth of the absorbers in the PNL tests is very low and any significant errors that might exist in the treatment of strong thin absorbers could not be revealed.

Table 4A.3 lists the subset of experiments using thin neutron absorbers (from Table 4A. 1) and shows the reactivity worth (Ak) of the absorbernt No trends with reactivity worth of the absorber are evident, although based on the calculations shown in Table 4A.3, some of the B&W critical experiments seem to have unusually large experimental errors. B&W made an effort to report some of their experimental errors. Other laboratories did not evaluate their experimental errors.

To further confirm the absence of a significant trend with 10B concentration in the absorber, a cross-comparison was made with MCNP4a and KENO5a (as suggested in Reg.

Guide 3.41). Results are shown in Figure 4A.6 and Table 4A.4 for a typical geometry.

These data substantiate the absence of any error (trend) in either of the two codes for the conditions analyzed (data points fall on a 450 line, within an expected 95 % probability limit).

The reactivity worth of the absorber panels was determined by repeating the calculation with the absorber analytically removed and calculating the incremental (Ak) change in reactivity due to the absorber.

Appendix 4A, Page 4

Miscellaneous and Minor Parameters 4A.4.1 Reflector Material and Spacings PNL has performed a number of critical experiments with thick steel and lead reflectors.'

Analysis of these critical experiments are listed in Table 4A.5 (subset of data in Table 4A. 1). There appears to be a small tendency toward overprediction of kly at the lower spacing, although there are an insufficient number of data points in each series to allow a quantitative determination of any trends. The tendency toward overprediction at close spacing means that the rack calculations may be slightly more conservative than otherwise.

4A.4.2 Fuel Pellet Diameter and Lattice Pitch The critical experiments selected for analysis cover a range of fuel pellet diameters from 0.311 to 0.444 inches, and lattice spacings from 0.476 to 1.00 inches. In the rack designs, the fuel pellet diameters range from 0.303 to 0.3805 inches O.D. (0.496 to 0.580 inch lattice spacing) for PWR fuel and from 0.3224 to 0.494 inches O.D. (0.488 to 0.740 inch lattice spacing) for BWR fuel. Thus, the critical experiments analyzed provide a reasonable representation of power reactor fuel. Based on the data in Table 4A. 1, there does not appear to be any observable trend with either fuel pellet diameter or lattice pitch, at least over the range of the critical experiments applicable to rack designs.

4A.4.3 Soluble Boron Concentration Effects Various soluble boron concentrations were used in the B&W series of critical experiments and in one PNL experiment, with boron concentrations ranging up to 2550 ppm. Results of MCNP4a (and one KENO5a) calculations are shown in Table 4A.6. Analyses of the very high boron concentration experiments (> 1300 ppm) show a tendency to slightly overpredict reactivity for the three experiments exceeding 1300 ppm. In turn, this would suggest that the evaluation of the-racks with higher soluble boron concentrations could be slightly conservative.

Parallel experiments with a depleted uranium reflector were also performed but not included in the present analysis since they are not pertinent to the Holtec rack design.

Appendix 4A, Page 5 4A.4

4A.5 MOX Fuel The number of critical experiments with PuO2 bearing fuel (MOX) is more limited than for U0 2 fuel. However, a number of MOX critical experiments have been analyzed and the results are shown in Table 4A.7. Results of these analyses are generally above a kff of 1.00, indicating that when Pu is present, both MCNP4a and KENO5a overpredict the reactivity. This may indicate that calculation for MOX fuel will be expected to be conservative, especially with MCNP4a. It may be noted that for the larger lattice spacings, the KENO5a calculated reactivities are below 1.00, suggesting that a small trend may exist with KENO5a. It is also possible that the overprediction in kff for both codes may be due to a small inadequacy in the determination of the Pu-241 decay and Am-241 growth. This possibility is supported by the consistency in calculated klf over a wide range of the spectral index (energy of the average lethargy causing fission).

Appendix 4A, Page 6

4A.6 References

[4A. 1] J.F. Briesmeister, Ed., "MCNP4a - A General Monte Carlo N Particle Transport Code, Version 4A; Los Alamos National Laboratory, LA-12625-M (1993).

[4A.2] SCALE 4.3, 'A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation", NUREG-0200 (ORNL-NUREG-CSD-2/U2/R5, Revision 5, Oak Ridge National Laboratory, September 1995.

[4A.3] M.D. DeHart and S.M. Bowman, "Validation of the SCALE Broad Structure 44-G Group ENDFIB-Y Cross-Section Library for Use in Criticality Safety Analyses", NUREG/CR-6102 (ORNL/TM-12460)

Oak Ridge National Laboratory, September 1994.

[4A.4] W.C. Jordan et al., "Validation of KENOV.a", CSD/TM-238, Martin Marietta Energy Systems, Inc., Oak Ridge National Laboratory, December 1986.

[4A.5]

O.W. Hermann et al., "Validation of the Scale System for PWR Spent Fuel Isotopic Composition Analysis", ORNL-TM-12667, Oak Ridge National Laboratory, undated.

[4A.6] R.J. Larsen and M.L. Marx, An Introduction to Mathematical Statistics and its Applications, Prentice-Hall, 1986.

[4A.7] M.N. Baldwin et al., Critical Experiments Supporting Close Proximity Water Storage of Power Reactor Fuel, BAW-1484-7, Babcock and Wilcox Company, July 1979.

[4A.8]

G.S. Hoovier et al., Critical Experiments Supporting Underwater Storage of Tightly Packed Configurations of Spent Fuel Pins, BAW 1645-4, Babcock & Wilcox Company, November 1991.

[4A.9] L.W. Newman et al., Urania Gadolinia: Nuclear Model Development and Critical Experiment Benchmark, BAW-1810, Babcock and Wilcox Company, April 1984.

Appendix 4A, Page 7

[4,A.10] J.C. Manaranche et al., "Dissolution and Storage Experimental Program with 4.75 wlo Enriched Uranium-Oxide Rods," Trans.

Am. Nuel. Soe. 33: 362-364 (1979).

[4A. 11] S.R. Bierman and E.D. Clayton, Criticality Experiments with Subcritical Clusters of 2.35 wlo and 4.31 w/o 235U Enriched U02 Rods in Water with Steel Reflecting Walls, PNL-3602, Battelle Pacific Northwest Laboratory, April 1981.

[4A. 12] S.R. Biennan et al., Criticality Experiments with Subcritical Clusters of 2.35 w/o and 4.31 w/o z35 Enriched UO2 Rods in Water with Uranium or Lead Reflecting Walls, PNL-3926, Battelle Pacific Northwest Laboratory, December, 1981.

[4A'*3] S.R. Bierman et al., Critical Separation Betweeji Subcritical S*'Clusters of 4.31 w/o z35U Enriched UO2 Rods in Water with Fixed Neutron Poisons, PNL-2615, Battelle Pacific Northwest Laboratory, October 1977.

[4A. 14] S.R. Bierman, Criticality Experiments with Neutron Flux Traps Containing Voids, PNL-7167, Battelle Pacific Northwest Laboratory, April 1990.

[4A.15] B.M. Durst et al., Critical Experiments with 4.31 wt % z~5U Enriched UO2 Rods in Highly Borated Water Lattices, PNL-4267, Battelle Pacific Northwest Laboratory, August 1982.

[4A. 16] S.R. Bierman, Criticality Experiments with Fast Test Reactor Fuel Pins in Organic Moderator, PNL-5803, Battelle Pacific Northwest Laboratory, December 1981.

[4A. 17] E.G. Taylor et al., Saxton Plutonium Program Critical Experiments for the Saxton Partial Plutonium Core, WCAP-3385-54, Westinghouse Electric Corp., Atomic Power Division, December 1965.

[4A. 18] M.G. Natrella, Experimental Statistics, National Bureau of Standards, Handbook 91, August 1963.

Appendix 4A, Page 8

Table 4A.1 Summary of Criticality Benchmark Calculations Calculated k.

U Identification Enrich.

MCNP4a KENO5a EALF t (eV)

MCNP4a KENO5a 1

B&W-1484 (4A.7)

Core I 2.46 0.9964 +/- 0.0010 0.9898+/- 0.0006 0.1759 0.1753 2

B&W-1484 (4A.7)

Core II 2.46 1.0008 +/- 0.0011 1.0015 +/- 0.0005 0.2553 0.2446 3

B&W-1484 (4A.7)

Core Ell 2.46 1.0010 +/- 0.0012 1.0005 +/- 0.0005 0.1999 0.1939 4

B&W-1484 (4A.7)

Core IX 2.46 0.9956 +/- 0.0012 0.9901 +/- 0.0006 0.1422 0.1426 5

B&W-1484 (4A.7)

Core X 2.46 0.9980 +/- 0.0014 0.9922 +/- 0.0006 0.1513 0.1499 6

B&W-1484 (4A.7)

Coie X 2.46 0.9978 +/- 0.0012 1.0005 +/- 0.0005 0.2031 0.1947 7

B&W-1484 (4A.7)

Core XII 2.46 0.9988 +/- 0.0011 0.9978 +/- 0.0006 0.1718 0.1662 8

B&W-1484 (4A.7)

Core XII 2.46 1.0020 +/- 0.0010 0.9952 +/- 0.0006 0.1988 0.1965 9

B&W-1484 (4A.7)

Core XIV 2.46 0.9953 +/- 0.0011 0.9928 +/- 0.0006 0.2022 0.1986 10 B&W-1484 (4A.7)

Core XV I

2.46 0.9910 +/- 0.0011 0.9909 +/- 0.0006 0.2092 0.2014 11 B&W-1484 (4A.7)

Core XVI tV 2.46 0.9935 +/- 0.0010 0.9889 +/- 0.0006 0.1757 0.1713 12 B&W-1484 (4A.7)

Core XVII 2.46 0.9962 +/- 0.0012 0.9942 +/- 0.0005 0.2083 0.2021 13 B&W-1484 (4A.7)

Core XVMI 2.46 1.0036 +/- 0.0012 0.9931 +/- 0.0006 0.1705 0.1708

'b--,9A

.. 1° A A fl..,,

i Appendhxwtjage.

Table 4A.1 Summary of Criticality Benchmark Calculations Calculatedlkf Oaf aranrn Ird~ntifleatonn Enrich.

MCNP4a KENO5a EALFt (eV-)

MCNP4a KENO5a 14 B&W-1484 (4A.7)

Core XIX 2.46 0.9961 +/- 0.0012 0.9971 +/- 0.0005 0.2103 0.2011 15 B&W-1484 (4A.7)

Core XX 2.46 1.0008 +/- 0.0011 0.9932 +/- 0.0006 0.1724 0.1701 16 B&W-1484 (4A.7)

Core XXI 2.46 0.9994 +/- 0.0010 0.9918 +/- 0.0006 0.1544 0.1536 17 B&W-1645 (4A.8)

S-type Fuel, w/886 ppm B 2.46 0.9970 +/- 0.0010 0.9924 +/- 0.06 1.4475 1.4680 18 B&W-1645 (4A.8)

S-type Fuel, w/746 ppm B 2.46 0.9990 +/- 0.0010 0.9913 +/- 0.0006 1.5463 1.5660 19 B&W-1645 (4A.8)

SO-type Fuel, w/1156 ppm B 2.46 0.9972 +/- 0.0009 0.9949 +/- 0.0005 0.4241 0.4331 20 B&W-1810 (4A.9)

Case 1 1337 ppm B 2.46 1.0023 +/- 0.0010 NC 0.1531 NC 21 B&W-1810 (4A.9)

Case 12 1899 ppm B 2.46/4.02 1.0060 +/- 0.0009 NC 0.4493 NC 22 Firench (4A.10)

Water Moderator 0 gap 4.75 0.9966 +/- 0.0013 NC 0.2172 NC 23 French (4A.10)

Water Moderator 2.5 cm gap 4.75 0.9952 +/- 0.0012 NC 0.1778 NC 24 French (4A.10)

Water Moderator 5 cm gap 4.75 0.9943 +/- 0.0010 NC 0.1677 NC 25 French (4A.10)

Water Moderator 10 cm gap 4.75 0.9979 +/- 0.0010 NC 0.1736 NC 26 PNL-3602 (4A.11)

Steel Reflector, 0 separation 2.35 NC 1.0004 +/- 0.0006 NC 0.1018 Appendix 4A, Page 10

Table 4A.1 Summary of Criticality Benchmark Calculations Calculated kif Tudpntflpgtinf~r Enrich.

MCNP4a KENO5a EALF t (eV)

MCNP4a KENO5a 27 PNL-3602 (4A.11)

Steel Reflector, 1.321 cm sepn.

2.35 0.9980 +/- 0.0009 0.9992 +/- 0.0006 0.1000 0.0909 28 PNL-3602 (4A.11)

Steel Reflector, 2.616 cm sepn 2.35 0.9968 +/- 0.0009 0.9964 +/- 0.0006 0.0981 0.0975 29 PNL-3602 (4A.11)

Steel Reflector, 3.912 cm sepn.

2.35 0.9974 +/- 0.0010 0.9980 +/- 0.0006 0.0976 0.0970 30 PNL-3602 (4A.11)

Steel Reflector, Infinite sepn.

2.35 0.9962 +/- 0.0008 0.9939 +/- 0.0006 0.0973 0.0968 31 PNL-3602 (4A.11)

Steel Reflector, 0 cm sepn.

4.306 NC 1.0003 +/- 0.0007 NC 0.3282 32 PNL-3602 (4A.11)

Steel Reflector, 1.321 cm sepn.

4.306 0.9997 +/- 0.0010 1.0012 +/- 0.0007 0.3016 0.3039 33 PNL-3602 (4A.11)

Steel Reflector, 2.616 cm sepn.

4.306 0,9994 +/- 0.0012 0.9974 +/- 0.0007 0.2911 0.2927 34 PNL-3602 (4A.11)

Steel Reflector, 5.405 cm sepn.

4.306 0.9969 +/- 0.0011 0.9951 +/- 0.0007 0.2828 0.2860 35 PNL-3602 (4A.11)

Steel Reflector, Infinite sepn. It 4.306 0.9910 +/- 0.0020 0.9947 +/- 0.0007 0.2851 0.2864 36 PNL-3602 (4A.11)

Steel Reflector, with Boral Sheets 4.306 0.9941 +/- 0.0011 0.9970 +/- 0.0007 0.3135 0.3150 37 PNL-3926 (4A.12)

Lead Reflector, 0 cm sepn.

4.306 NC 1.0003 +/- 0.0007 NC 0.3159 38 PNL-3926 (4A.12)

Lead Reflector, 0.55 cm sepn.

4.306 1.0025 +/- 0.0011 0.9997 +/- 0.0007 0.3030 0.3044 39 PNL-3926 (4A.12)

Lead Reflector, 1.956 cm sepn.

4.306 1.0000 +/- 0.0012 0.9985 +/- 0.0007 0.2883 0.2930 Appendix 4A, Page 11

Table 4A.1 Summary of Criticality Benchmark Calculations Calculated k,_

Identification Enrich.

MCNP4a KENO5a EALF ' (eV)

MCNP4a KENOSa 40 PNL-3926 (4A.12)

Lead Reflector, 5.405 cm sepn.

4.306 0.9971 +/- 0.0012 0.9946 +/- 0.0007 0.2831 0.2854 41 PNL-2615 (4A.13)

Experiment 004/032 - no absorber 4.306 0.9925 +/- 0.0012 0.9950 +/- 0.0007 0.1155 0.1159 42 PNL-2615 (4A.13)

Experiment 030

- Zr plates 4.306 NC 0.9971 +/- 0.0007 NC 0.1154 43 PNL-2615 (4A.13)

Experiment 013

- Steel plates 4.306 NC 0.9965 +/- 0.0007 NC 0.1164 44 PNL-2615 (4A.13)

Experiment 014

- Steel plates 4.306 NC 0.9972 +/- 0.0007 NC 0.1164 45 PNL-2615 (4A.13)

Exp. 009 1.05% Boron-Steel plates 4.306 0.9982 +/- 0.0010 0.9981 +/- 0.0007 0.1172 0.1162 46 PNL-2615 (4A.13)

Exp. 012 1.62% Boron-Steel plates 4.306 0.9996 +/- 0.0012 0.9982 +/- 0.0007 0.1161 0.1173 47 PNL-2615 (4A.13)

Exp. 031 - Boral plates 4.306 0.9994 +/- 0.0012 0.9969 +/- 0.0007 0.1165 0.1171 48 PNL-7167 (4A.14)

Experiment 214R - with flux trap 4.306 0.9991 +/- 0.0011 0.9956 +/- 0.0007 0.3722 0.3812 49 PNL-7167 (4A.14)

Experiment 214V3 - with flux trap 4.306 0.9969 +/- 0.0011 0.9963 +/- 0.0007 0.3742 0.3826 50 PNL-4267 (4A.15)

Case 173 - 0 ppm B 4.306 0.9974 +/- 0.0012 NC 0.2893 NC 51 PNL-4267 (4A.15)

Case 177 - 2550 ppm B 4.306 1.0057 +/- 0.0010 NC 0.5509 NC 52 PNL-5803 (4A.16)

MOX Fuel - Type 3.2 Exp. 21 20% Pu 1.0041 +/- 0.0011 1.0046 +/- 0.0006 0.9171 0.8868 Appendix 4A, Page 12 Reference

Table 4A.1 Summary of Criticality Benchmark Calculations Calculated kf Identification Enrich.

MCNF4a.

KENO5a EALFt (eV)

MCNP4a KENO5a 53 PNL-5803 (4A.16)

MOX Fuel - Type 3.2 Exp. 43 20% Pu 1.0058 +/- 0.0012 1.0036 +/- 0.0006 0.2968 0.2944 54 PNL-5803 (4A.16)

MOX Fuel - Type 3.2 Exp. 13 20% Pu 1.0083 +/- 0.0011 0.9989 +/- 0.0006 0.1665 0.1706 55 PNL-5803 (4A.16)

MOX Fuel - Type 3.2 Exp. 32 20% Pu 1.0079 +/- 0.0011 0.9966 +/- 0.0006 0.1139 0.1165 56 WCAP-3385 (4A.17)

Saxton Case 52 PuO2 0.52" pitch 6.6% Pu 0.9996 +/- 0.0011 1.0005 +/- 0.0006 0.8665 0.8417 57 WCAP-3385 (4A.17)

Saxton Case 52 U 0.52" pitch 5.74 1.0000 +/- 0.0010 0.9956 +/- 0.0007 0.4476 0.4580 58 WCAP-3385 (4A.17)

Saxton Case 56 PuO2 0.56" pitch 6.6% Pu 1.0036 +/- 0.0011 1.0047 +/- 0.0006 0.5289 0.5197 59 WCAP-3385 (4A.17)

Saxton Case 56 borated PuO2 6.6% Pu 1.0008 +/- 0.0010 NC 0.6389 NC 60 WCAP-3385 (4A.17)

Saxton Case 56 U 0.56" pitch 5.74 0.9994 +/- 0.0011 0.9967 +/- 0.0007 0.2923 0.2954 61 WCAP-3385 (4A.17)

Saxton Case 79 PuO2 0.79" pitch 6.6% Pu 1.0063 +/- 0.0011 1.0133 +/- 0.0006 0.1520 0.1555 62 WCAP-3385 (4A.17)

Saxton Case 79 U 0.79" pitch 5.74 1.0039 +/- 0.0011 1.0008 +/- 0.0006 0.1036 0.1047 Notes: NC stands for not calculated.

t EALF is the energy of the average lethargy causing fission.

It These experimental results appear to be statistical outliers (> 3 a) suggesting the possibility of unusually large experimental error. Although they could justifiably be excluded, for conservatism, they were retained in determining the calculational basis.

Appendix 4A, Page 13 Reference

Table 4A.2 COMPARISON OF MCNP4a AND KENO5a CALCULATED REACTIVITIESt FOR VARIOUS ENRICHMENTS Calculated ke+_ +/- lo Enrichment MCNP4a KENO5a 3.0 0.8465 +/- 0.0011 0.8478 + 0.0004 3.5 0.8820 +/- 0.0011 0.8841 +/- 0.0004 3.75 0.9019 + 0.0011 0.8987 +/- 0.0004 4.0 0.9132 +/- 0.0010 0.9140 - 0.0004 4.2 0.9276 + 0.0011 0.9237 + 0.0004 4.5 0.9400 +/- 0.0011 0.9388 +/- 0.0004 f

Based on the GE 8x8R fuel assembly.

Appendix 4A, Page 14

Table 4A.3 MCNP4a CALCULATED REACTIVITIES FOR CRITICAL EXPERIMENTS WITH NEUTRON ABSORBERS tEALF is the energy of the average lethargy causing fission.

Appendix 4A, Page 15 Ak MCNP4a Worth of Calculated EALF t Ref.

Experiment Absorber k*

(eV) 4A.13 PNL-2615 Boral Sheet 0.0139 0.9994+/-0.0012 0.1165 4A.7 B&W-1484 Core XX 0.0165 1.0008+/-0.0011 0.1724 4A.13 PNL-2615 1.62% Boron-steel 0.0165 0.9996+/-0.0012 0.1161 4A.7 B&W-1484 Core XIX 0.0202 0.9961+/-0.0012 0.2103 4A.7 B&W-1484 Core XXI 0.0243 0.9994+/-0.0010 0.1544 4A.7 B&W-1484 Core XVII 0.0519 0.9962+/-0.0012 0.2083 4A.1I PNL-3602 Boral Sheet 0.0708 0.9941+/-0.0011 0.3135 4A.7 B&W-1484 Core XV 0.0786 0.9910+/-0.0011 0.2092 4A.7 B&W-1484 Core XVI 0.0845 0.9935+/-0.0010 0.1757 4A.7 B&W-1484 Core XIV 0.1575 0.9953+/-0.0011 0.2022 4A.7 B&W-1484 Core XIII 0.1738 1.0020+/-0.0011 0.1988 4A. 14 PNL-7167 Expt 214R flux trap 0.1931 0.9991+/-0.0011 0.3722

Table 4A.4 COMPARISON OF MCNP4a AND KENO5a CALCULATED REACTIVITIF~t FOR VARIOUS 'B LOADINGS Calculated kI +/- la

'0B, g/cm2 MCNP4a KENO5a 0.005 1.0381 +/- 0.0012 1.0340 +/- 0.0004 0.010 0.9960 +/- 0.0010 0.9941 + 0.0004 0.015 0.9727 +/- 0.0009 0.9713 + 0.0004 0.020 0.9541 +/- 0.0012 0.9560 +/- 0.0004 0.025 0.9433 +/- 0.0011 0.9428 +/- 0.0004 0.03 0.9325 +/- 0.0011 0.9338 + 0.0004 0.035 0.9234 + 0.0011 0.9251 +/- 0.0004 0.04 0.9173 +/- 0.0011 0.9179 + 0.0004 t

Based on a 4.5% enriched GE 8x8R fuel assembly.

Appendix 4A, Page 16

Table 4A.5 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH THICK LEAD AND STEEL REFLECTORSt Separation, Ref.

Case E, wt%

cm MCNP4a KIf KENO5a k.,

4A.11 Steel 2.35 1.321 0.9980+/-0.0009 0.9992+/-0.0006 Reflector 2.35 2.616 0.9968+/-0.0009 0.9964+/-0.0006 2.35 3.912 0.9974+/-0.0010 0.9980+/-0.0006 2.35 00 0.9962+/-0.0008 0.9939+/-0.0006 4A.11 Steel 4.306 1.321 0.9997+/-0.0010 1.0012+/-0.0007 Reflector 4.306 2.616 0.9994+/-0.0012 0.9974+/-0.0007 4.306 3.405 0.9969+/-0.0011 0.9951+/-0.0007 4.306 Co 0.9910+/-0.0020 0.9947+/-0.0007 4A. 12 Lead 4.306 0.55 1.0025+/-0.0011 0.9997+/-0.0007 Reflector 4.306 1.956 1.0000+/-0.0012 0.9985+/-0.0007 4.306 5.405 0.9971+/-0.0012 0.9946+/-0.0007 t

Arranged in order of increasing reflector-fuel spacing.

Appendix 4A, Page 17

Table 4A.6 CALCULATIONS FOR CRITICAL EXPERJMENTS WITH VARIOUS SOLUBLE BORON CONCENTRATIONS Calculated k, Boron Concentration, Reference Experiment ppm MCNP4a KENO5a 4A.15 PNL-4267 0

0.9974 +/- 0.0012 4A.8 B&W-1645 886 0.9970 +/- 0.0010 0.9924 +/- 0.0006 4A.9 B&W-1810 1337 1.0023 + 0.0010 4A.9 B&W-1810 1899 1.0060 +/- 0.0009 4A.15 PNL-4267 2550 1.0057 +/- 0.0010 Appendix 4A, Page 18

Table 4A.7 CALCULATIONS FOR CRITICAL EXPERIMENTS WITH MOX FUEL MCNP4a KENOSa Reference Case*

kd E

• kf EAU"I PNL-5803 MOX Fuel - Exp. No. 21 1.0041+/-0.0011 0.9171 1.0046+/-0.0006 0.8868

[4A.161 MOX Fuel - Exp. No. 43 1.0058+/-0.0012 0.2968 1.0036+/-0.0006 0.2944 MOX Fuel - Exp. No. 13 1.0083+/-0.0011 0.1665 0.9989+/-0.0006 0.1706 MOX Fuel - Exp. No. 32 1.0079+/-0.0011.

0.1139 0.9966+/-0.0006 0.1165 WCAP-Saxton @ 0.52" pitch 0.9996+/-0.0011 0.8665 1.0005+/-0.0006 0.8417 3385-54

[4A.17]

Saxton @ 0.56" pitch 1.0036+/-0.0011 0.5289 1.0047+/-0.0006 0.5197 Saxton @ 0.56" pitch borated 1.0008+/-0.0010 0.6389 NC NC Saxton @ 0.79" pitch 1.0063+/-0.0011 0.1520 1.0133+/-0.0006 0.1555 Note: NC stands for not calculated t

Arranged in order of increasing lattice spacing.

tt EALF is the energy of the average lethargy causing fission.

Appendix 4A, Page 19

Linear Regression with Correlation Coefficient of 0.13 Energy of Average Lethargy Causing Fission (Log Scale)

FIGURE 4A.1 MCNP CALCULATED k-eff VARIOUS VALUES OF THE VALUES for SPECTRAL INDEX 1.010 1.005 N-I 0

1.000 0.995 0.990 0.1 1

I

Linear Regression with Correlation Coefficient of 0.21 Energy of Average Lethargy Causing Fission (Log Scale)

FIGURE 4A.2 KENO5a CALCULATED VARIOUS VALUES OF k-eff VALUES THE SPECTRAL 1.010 1.005 4.-

-3 C,

0 1.000 0.995 0.990 0.985 0.1 1

FOR INDEX

Correlation Coefficient of 0.03 I) r r

r T

1.010 1,005 1.000 0.995 0.990 4t AP1rl,:

i) 4 1

1---t t11

_I-------------~i

+/-

I I

I T

'f 1

3.0 3.5 Enrichment, w/o

)

4.0 U-235

(

I ! I I I I I I 1 -1 i I i i i i i i i i i i 4.5 5.0 5.5 6.0 FIGURE 4A.3 MCNP CALCULATED k AT VARIOUS U-235 eff VALUES ENRICHMENTS (D

C)

-o N-

"0 CO 03 C.,

4 +

+

ti F!

z z

2.

F1 11 1 1!

0 2.5 g

I

-J-4 I ----

i

-I

]-I I f 1 1 1

-- -Linear Regression with

..(

u J

)

)

I I

Linear Regression with Correlation Coefficient of 0.38 Enrichment, w/o U-235 FIGURE 4A.4 KENO CALCULATED. k-eff VALUES AT VARIOUS U-235 ENRICHMENTS 1.010 1,005 0*1.000

"-4 4

-o 2 0.995 0.990 0.985

MCNP k-eff Calculations FIGURE 4A.5 COMPARISON OF MCNP AND KENO5A CALCULATIONS FOR VARIOUS FUEL ENRICHMENTS 0.94 0.92 E

0.90 0.88 C,)

0 70 C) 0 0

u)

In 0

z ILJ 0.86 0.84

1.04-1.03 1.01 C

3: 0.99

• ~0.98 S

0.97 "0

• 0.96 0.020 g/c i

0.95 0.025

/Gmzq o.94

e. 9.3..030 g//am3 0.93 0.035 g/amiuq 0.92 0.04 g/lo aq

0. 91 0.900 0.920 0.940 0.960 0.980 1.000 1.020 1.040 Reactivity Calculated with KENO5a FIGURE 4A.6 COMPARISON OF MCNP AND KENO5ct CALCULATIONS FOR VARIOUS BORON-I 0 AREAL DENSITIES

THERMAL-HYDRAULIC CONSIDERATIONS The proposed change does not entail any physical modifications to fuel, storage racks, or plant cooling systems. No changes in fresh fuel enrichment limits, constraints on maximum fuel rod burnup, or cooling time restrictions prior to the manipulation of irradiated fuel are being proposed. There will be no changes to the spent fuel decay heat load or to the SFP cooling system capabilities. Therefore, the previous thermal-hydraulic evaluations performed for the SFP remain valid.

Holtec Report HI-2022940 5-1 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION 5.0

6.0 RACK SEISMIC/STRUCTURAL CONSIDERATIONS The proposed change does not require any physical modifications to fuel, the fuel storage racks, or to plant structural systems. No new equipment is required to be installed. Changes in water coolant density will not significantly affect any of the evaluations previously performed for the racks. All loading conditions and load combinations previously considered remain valid. Therefore, the previously performed rack seismic/structural evaluations and reported results remain valid.

Holtec Report HI-2022940 6-1 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

MECHANICAL ACCIDENT ASSESSMENT A spent fuel pool boron dilution analysis has been performed by Florida Power and Light and is included as an enclosure to the license amendment request. This dilution analysis includes a discussion of certain postulated accident conditions that can increase the pool water inventory (i.e., break in a makeup line) from the perspective of fuel pool reactivity.

The proposed change does not require any physical modifications to fuel, storage racks, or plant structures. The proposed change does not require installation of new equipment or require the removal of any existing plant equipment. The change does not produce any new potential accident conditions, because no changes to fuel handling techniques or fuel handling equipment are required to implement the proposed license amendment. The change does not produce any greater potential for previously postulated accident conditions to occur; fuel weight is not increased, the interface between fuel and the hoist grapple apparatus is not changed, and no other aspects of equipment used to perform fuel or control rod manipulation are changed.

The proposed license amendment change does not cause an increase in consequences of any postulated accident, because no changes in fresh fuel enrichment, the limitations on maximum fuel rod burnup, or minimum post-irradiation cooling times are proposed. Therefore, the previously performed mechanical accident evaluations for postulated fuel drops and the associated reported results remain valid.

Holtec Report HI-2022940 7-1 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION 7.0

POOL STRUCTURE ASSESSMENT The proposed change does not require any physical modifications to fuel, storage racks, or plant structural systems. No new equipment is required to be installed. Changes in water coolant density will not significantly affect any of the evaluations previously performed for the pool structure. All loading conditions and load combinations previously considered remain valid. Therefore, the previously performed rack pool structure evaluations and reported results remain valid.

Holtec Report HI-2022940 8-1 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION 8.0

RADIOLOGICAL CONSIDERATIONS The proposed change does not require any physical modifications to fuel or to the fuel storage racks.

The proposed change does not increase the amount of fuel stored in the fuel storage racks or cause the quantity of other activated material to increase. As a result, no new radiological source terms need to be considered. Changes in water coolant density will not significantly affect any of the radiological evaluations previously performed for the racks. It is noted that the credit taken for soluble boron does not in any way affect the soluble boron already present within the pool.

The maximum allowed fuel enrichment, fuel rod burnup, and the minimum allowed post-irradiation fuel cooling time remain unchanged. Thus, the spent fuel source terms remain unchanged. The revised fuel storage configuration will not significantly affect the location of source terms represented by spent fuel.

The proximity of fuel to the pool water surface and exterior walls will not change. Thus, the radiation attenuation provided by the walls and the fuel pool water inventory remains unchanged. Therefore, the previously performed radiological evaluations and their reported results remain valid. Dose levels surrounding the SFP are not expected to change significantly after implementing the proposed change.

Holtec Report HI-2022940 9-1 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION 9.0

10.0 ENVIRONMENTAL/COST BENEFIT ASSESSMENT Florida Power and Light has prepared an environmental and cost benefit assessment (enclosure of the 1 OCFR50.92 evaluation) of the proposed license amendment. This assessment examines the underlying need for actions to mitigate the consequences of Boraflex degradation at St. Lucie Unit 1 and it considers the thermal and radiological impacts on the environment of the proposed change. This assessment also considers the occupational exposure that will be incurred as the proposed license amendment is implemented. In the assessment FPL identified several alternative methods of managing the storage of irradiated nuclear fuel and it examined the environmental and economic consequences of each candidate alternative. Finally, the assessment considered the ramifications of a "no action" alternative.

The conclusion of the environmental assessment is that none of the alternatives examined has a lower overall impact on the environment than the proposed alternative, which credits the presence of soluble boron in the fuel pool and the repositioning of stored irradiated fuel. The occupational exposure plant workers can expect to receive during the fuel repositioning campaign has been conservatively estimated and is a small fraction of the St. Lucie site's annual radiation exposure budget. Finally, the assessment concluded that none of the alternatives considered, including the "no action" alternative is economically superior to the chosen alternative.

Holtec Report HI-2022940 10-1 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION