Text

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)
	v • d • e

St. Lucie Unit 1 L-2002-221 Enclosure 3 Docket No. 50-335 Proposed License Amendment Spent Fuel Pool Soluble Boron Credit NON-PROPRIETARY HOLTEC LICENSE AMENDMENT REPORT (Bound Report)

HO LT E C

.mmm. Holtec Center, 555 Lincoln Drive West, Marlton, NJ 08053 Telephone (856) 797-0900 Fax (856) 797-0909 INTER-NATIO NAL 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 Remedy ThflCiI MENT NAME: St. Lucie Unit 1 SFP Storage Rack Boraflex Degradation DOCUMENT NO.: HI-2022940 CATEGORY: D- GENERIC 1237 [ PROJECT SPECIFIC PROJECT NO.:

Rev. Date Author's Rev. Date Author's Initials VIR # No. Approved Initials VIR #

No. 2 Approved 0 11/13/02 SP 118818 DOCUMENT CATEGORIZATION Holtec Quality Procedures In accordance with the Holtec Quality Assurance Manual and associated (HQPs), this document is categorized as a:

Technical Report (Per HQP 3.2)

-- Calculation Package 3 (Per HQP 3.2) (Such as a Licensing Report)

Design Specification (Per HQP 3.4)

F-- Design Criterion Document (Per HQP 3.4) --

E] Other (Specify):

DOCUMENT FORMATTING instructions of HQP 3.2.

The formatting of the contents of this document is in accordance with the DECLARATION OF PROPRIETARY STATUS This document is labeled:

E] Nonproprietary FD Holtec Proprietary 0 TOP SECRET property of Holtec International.

Documents labeled TOP SECRET contain extremely valuable intellectual/commercial without explicit approval of a company corporate officer.

They cannot be released to external organizations or entities to safeguard it Secret document bears full and undivided responsibility The recipient of Holtec's proprietary or Top against loss or duplication.

Notes approval process set forth in the Holtec Quality

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

have recorded their password-controlled electronic concurrence to the to this document will be ordered by the Project Manager and carried out if any of its contents is

2. A revision The determination as to the need for revision will be made by the materially affected during evolution of this project.

Project Manager with input from others, as deemed necessary by him.

to the document and replacing the

3. Revisions to Calculation Packages may be made by adding supplements Log".

"Table of Contents", the "Review and Certification" page and the "Revision

TABLE OF CONTENTS

SUMMARY

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

SUMMARY

OF REVISIONS 4 pages TABLE OF CONTENTS 3 pages

1.0 INTRODUCTION

7 pages 2.0 GENERAL ARRANGEMENT SOLUBLE BORON DILUTION ACCIDENT I pages 3.0 4.0 CRITICALITY SAFETY ANALYSES 46 pages APPENDIX 4A 26 pages 1 age 5.0 THERMAL-HYDRAULIC CONSIDERATIONS 1 page 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 96 pages ITOTAL R-1 1253 Holtec Report HI-2022940

TABLE OF CONTENTS 1-1

1.0 INTRODUCTION

1-3 1.1 References ............................................................................................................

2-1 2.0 GENERAL ARRANGEM ENT ...........................................................................

3-1 3.0 SOLUBLE BORON DILUTION ACCIDENT ...................................................

4-1 4.0 CRITICALITY SAFETY ANALYSES ..............................................................

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

4-5 4.2 Acceptance Criteria ..............................................................................................

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

4-6 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 ..... .. ..............................................

4-8 4.6 Analysis ................................................................................................................

4-8 4.6.1 Bounding Fuel Assemblies .................................................................................. 4-9 4.6.2 Pool W ater Temperature Effects .....................................................................

4-10 4.6.3 Uncertainties Due to M anufacturing Tolerances ...............................................

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

4-12 4.6.6 Effects of Gadolinium ........................................................................................

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

4-13 4.6.8 Eccentric Fuel Assembly Positioning ................................................................

...................... 4-13 4.6.9 Reactivity Effect of Axial Burnup and Enrichment Distribution 4-14 4.6.10 B- 10 Depletion in CEAs ....................................................................................

4-15 4.6.11 Calculation of Burnup versus Enrichment Curves .............................................

4-17 4.6.12 Interfaces ............................................................................................................

4-18 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 ..........................................................................

4-19 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 ............................................................................

4-20 4.6.13 Soluble Boron Concentration for M aximum Keff of 0.95 ..................................

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

4-21 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 ............................................................................

4-22 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 .......................................................................

............... 4-23 4.7 References.................................

"BenchmarkCalculations" Appendix 4A ...........................................................................................

i 1253 Holtec Report HI-2022940

TABLE OF CONTENTS Total of 26 Pages including 6 figures and Summary ................... ............................ 4A- 4A-3 1 41 .1I Introduction 4 Effect Effect.ofEnrichmen.....................................

of Enrichm ent .........................................................................................

4' A.2 4A -4 4' A.3 Effect of l0B Loading ........................................................................................

4A-5 4'A.4 Miscellaneous and Minor Parameters ...............................................................

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

4A-5

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

4A-5

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

4A -6 4.A .5 M O X Fuel .........................................................................................................

4A -7 4.A .6 R eferences ....................................................................................................

5-1 5.0 THERMAL-HYDRAULIC CONSIDERATIONS ..............................................

6-1 6.0 RACK SEISMIC/STRUCTURAL CONSIDERATIONS ...................................

7-1 7.0 MECHANICAL ACCIDENT ASSESSMENT ...................................................

8-1 8 .0 POOL STRUCTURE ASSESSMENT ................................................................

9-1 9 .0 RADIOLOGICAL CONSIDERATIONS ............................................................

10-1 1.0.0 ENVIRONMENTAL COST/BENEFIT ASSESSMENT ..................................

ii 1253 Holtec Report HI-2022940

TABLE OF CONTENTS Tables 2-2 2.1 Rack D esign D ata ............................................................................................................ 2-3 2.2 Table of M odule Data ................................................................................................ 2-4 Module Dimensions and Weights ...... ...............................................................

2.3 Assemblies. 4-24 4.1.1 Minimum Burnup as a Function of Enrichment for Non-Blanketed 4-25 Assemblies ..........

4.1.2 Minimum Burnup as a Function of Enrichment for Blanketed 4-26 St. Lucie Unit 1 Fuel Assembly Specifications .............................................................

4.4.1 4-27 4.4.2 Control Element Assembly (CEA) Specifications ......................................................... 4-28 4.4.3 Core Operating Parameters for Depletion Analyses ...................................................... 4-29 4.4.4 St. Lucie Unit 1 Fuel Rack Dimensions .........................................................................

4.6.1 Comparison of Kinf for Various Fuel assembly Types 4-30 at Representative Fuel Conditions .................................................................................

4.5wt% Enrichment and 0 Years 4.6.2 Effect of Pool Water Temperature on Kinf for Fuel of 4-31 Cooling Time at 0 ppm Soluble Boron .......................................................................... 4-32 4.6.3 Reactivity Effect of Rack and Fuel Tolerances .......................... 4-33 versus Enrichment Curves.

4.6.4 Enrichment and Cooling Time Combinations for Burnup 4-34 4.6.5 Representative Calculation for each Case ......................................................................

Case ........................................... 4-35 4.6.6 Results of Additional Selected Calculations for each of 0.95 under 4.6.7 Soluble Boron Concentration for a Maximum keff Value 4-36 N ormal Conditions .........................................................................................................

of 0.95 under 4.6.8 Soluble Boron Concentration for a Maximum keff Value 4-37 Accident C onditions .......................................................................................................

4A-9 thru 4A-13 4A. 1 Summary of Criticality Benchmark Calculations .......................................

4A.2 Comparison of MCNP4a and Keno5a Calculated Reactivities 4A -14 for V arious Enrichm ents .......................................................................................

4A.3 MCNP4a Calculated Reactivities for Critical Experiments 4A -15 w ith N eutron Absorbers ..............................................................................................

for Various 4A.4 Comparison of MCNP4a and KENO5a Calculated Reactivities 4 A -16 "B ...............................................................................................................

and Steel Reflectors ............ 4A-17 4A.5 Calculations for Critical Experiments with Thick Lead Boron 4A.6 Calculations for Critical Experiments with Various Soluble 4A -18 C oncentrations ....................................................................................................... 4A-19 4A.7 Calculations for Critical Experiments with MOX Fuel ii 1253 Holtec Report HI-2022940

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) for Case 1, Low Reactivity 4.6.1 Minimum Burnup as a Function of Initial Enrichment Assemblies Case 1, High Reactivity 4.6.2 Minimum Burnup as a Function of Initial Enrichment for Assemblies for Case 2 4.6.3 Minimum Burnup as a Function of Initial Enrichment for Case 3 4.6.4 Minimum Burnup as a Function of Initial Enrichment for Case 4 4.6.5 Minimum Burnup as a Function of Initial Enrichment for Case 5 4.6.6 Minimum Burnup as a Function of Initial Enrichment Model for Fresh Fuel Assemblies in Region 4.6.7 Schematic Configuration of the Calculational 2 Racks for Inspection and Reconstitution the Spectral Index 4A.1 MCNP Calculated k-eff Values for Various Values of of the Spectral Index 4A.2 KENOSa Calculated k-eff Values for Various Values MCNP Calculated k-eff Values at Various U-235 Enrichments 4A.3 4A.4 KENO5a Calculated k-eff Values at Various U-235 Enrichments Various Fuel 4A.5 Comparison of MCNP and KENO5a Calculations for Enrichments for Various Boron-10 4A.6 Comparison of MCNP and KENO5a Calculations Areal Densities iv 1253 Holtec Report HI-2022940

1.0 INTRODUCTION

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

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

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

1,706 fuel assemblies. The existing high St. Lucie Unit I has a current licensed storage capacity of effort performed by Holtec in 1987.

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

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

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

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

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

and are calculated with a 95% probability at calculations, including manufacturing tolerances, confidence level [5].

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

of these postulated accidents are addressed are of the SFP structure. The radiological considerations Section 8 establishes the continued adequacy prepared a cost/benefit and environmental assessment documented in Section 9. Section 10 summarizes remediation proposal.

by FPL to address the Boraflex degradation 1237 1-2 Holtec Report HI-2022940 INFORMATION SHADED AREAS DENOTE PROPRIETARY

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

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

1.1 References Boraflex Neutron Absorber in Spent Fuel

[1] NRC Information Notice 95-38, Degradation of Storage Tacks," September 1995.

Report for Reracking St. Lucie Unit 1 Fuel

[2] Holtec International Report HI-87105, "Licensing Pool," Revision 3, dated April 1987.

[31 Not Used.

Boiler & Pressure Vessel Code,Section III,

[4] American Society of Mechanical Engineers (ASME),

1989 Edition, Subsection NF, and Appendices.

Bureau of Standards, Handbook 91, August

[5] M.G. Natrella, Experimental Statistics, National 1963.

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

2.0 GENERAL ARRANGEMENT individual cells with 8.65 inch (nominal)

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

maximum protection against structural loadings The high density racks are engineered to achieve number of available storage locations, (arising from ground motion, thermal stresses, etc.), the maximum Each rack module is equipped (see Figures 2.2 and to maintain fuel assemblies in a subcritical array.

each by 3-1 inches high. The girdle bar thickness on and 2.3) with girdle bars measuring 3/4 inches thick maintained between modules. Table 2.1 gives the rack ensures that a minimum gap of 1-1/2 inches is of the two regions are of eight different types. Tables relevant design data for each region. The modules type.

2.2 and 2.3 summarize the physical data for each module 1237 2-1 Holtec Report HI-2022940 INFORMATION SHADED AREAS DENOTE PROPRIETARY

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

Table 2.2 Table of Module Data Number of Number of Cells Number of Cells Total Number of Module Modules in N-S Direction in E-W Direction Cells per Module Identification 2 9 9 81 Region 1 Al to A2 9 10 90 Region 1 BI to B2 4 13 9 117 Region 2 CI to C4 3 13 8 104 Region 2 Dl to D3 2 11 8 88 Region 2 El to E2 Region 2 1 12 8 96 F1 2 12 9 108 Region2 G1 to G2 1 13 8 96 Region 2 H1 2-3 1237 Holtec Report HI-2022940 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 2-4 1237 Holtec Report HI-2022940 SHADED AREAS DENOTE PROPRIETARY INFORMATION

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

-. 0)ý ol I TYPICAL RACK ELEVATIOQ-REGION 1 Figure 2.2 HI-2022940

04- GIRDLE BAR RACK ELEVATION-REGION 2 Figure 2.3 TYPICAL HI-2022940

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

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

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

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

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

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

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

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, to it is necessary to group together fuel assemblies having similar reactivity characteristics and 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 to reflect'different reactivity groupings.

A total of seven fuel assembly types were developed description used in this report, and its The following table lists each type by number, the type of 4.5 weight percent:

minimum burnup requirement based on an initial enrichment Fuel Storage Configurations Analyzed with different fuel assembly types were A total of five fuel storage configurations (cases) analyzed, as follows:

fuel assemblies Case 1: Region 2, Checkerboard of high and low reactivity reactivity fuel Case 2: Region 1, Checkerboard of once burned and low reactivity fuel Case 3: Region I, Checkerboard of twice burned and lower assemblies with and without CEAs Case 4: Region 2, Checkerboard of high reactivity fuel only Case 5: Region 2, Medium reactivity fuel assemblies Burnup vs Enrichment Curves each assembly type in a checkerboard array, the For each storage configuration above, and for as a function of the initial enrichment of the fuel.

minimum required burnup has been determined 1237 4-2 Report No. HI-2022940 INFORMATION SHADED AREAS DENOTE PROPRIETARY

as polynomial These functions, also termed burnup versus enrichment curves, are established 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 assemblies with axial The current inventory of irradiated fuel at St. Lucie Unit 1 contains fuel for all cases, for non blankets, as well as fuel assemblies without axial blankets. Coefficients cooling times are listed blanketed and blanketed assemblies, and for all relevant post-irradiation in Table 4.1.1 and 4.1.2, respectively.

Special Fuel Loading Rules fuel pool wall. This part of the A portion of the periphery of Region 2 storage racks faces the neutron leakage in this area.

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

i.e.,

1. Normally, each rack module will contain only one of the above listed configurations, 1 rack. However, a rack Cases 1. 4, or 5 for a Region 2 rack, and Case 2 or 3 for a Region an empty row is used to module may contain more than one permissible configuration if different configuration.

separate fuel stored in one configuration from fuel stored in a 4-3 1237 Report No. HI-2022940 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

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

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

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

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

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

4-5 1237 Report No. HI-2022940 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 the assemblies, which were used for this analysis, are given in Table 4.4.1. Table 4.4.2 shows 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 4-6 1237 Report No. HI-2022940 SHADED AREAS DENOTE PROPRIETARY INFORMATION

MCNP4a was selected because it Carlo code developed at the Los Alamos National Laboratory.

of the necessary features has been used previously and verified for criticality analyses and has all cross-section data based on for this analysis. MCNP4a calculations used continuous energy ENDF/B-V and ENDF/B-VI.

of 0.0009 with an uncertainty Benchmark calculations, presented in Appendix A, indicate a bias 95% confidence level [4.7.1].

of+/- 0.0011 for MCNP4a, evaluated with a 95% probability at the platform and cross-section libraries The calculations for this analysis utilized the same computer A.

used for the benchmark calculations discussed in Appendix to the following parameters:

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

for the aforementioned criticality parameters to be used in of 10,000 histories per cycle, a Based on these studies, the final calculations use a minimum of 100 cycles were minimum of 25 cycles were skipped before averaging, a minimum over the fueled regions (assemblies).

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

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

in the spent fuel pool racks, the tolerances, and to perform various studies. For all calculations to zero.

Xe-135 concentration in the fuel is conservatively set 4.6 ANALYSIS to determine the acceptable storage criteria This section describes the calculations that were used it summarizes their results. In addition, this for both the Region 1 and Region 2 style racks and 1

accident conditions applicable to St. Lucie Unit section discusses the postulated abnormal and fuel pool storage.

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

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

models are discussed in the appropriate sections 4.6.1 Bounding Fuel Assemblies are performed for both assembly types listed To determine the bounding assembly, calculations bound cladding thickness listed in that in Table 4.4.1, and for both the upper bound and lower cooling times and burnups, table. Further, calculations are performed for various enrichments, 4-8 1237 Report No. HI-2022940 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 with Pool water temperature effects on reactivity at 0 ppm soluble boron have been calculated 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 and (Region 1 and Region 2). In these cases, a higher temperature results in a higher reactivity, 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 all coefficient is negative, i.e. a lower temperature results in a higher reactivity. Consequently, CASMO calculations for assemblies without CEAs are evaluated at 150 OF, whereas CASMO the calculations for assemblies crediting CEAs are evaluated at 4 'C, which corresponds to highest water density. For cases containing only assemblies without CEAs (cases 1, 2, 3 and 5),

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

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

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

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

4.6.3 UncertaintiesDue to Manufacturing Tolerances tolerances on reactivity In the calculation of the final k-infinity (kinf), the effect of manufacturing Factors considered must be included. CASMO-4 was used to perform these calculations.

of the fuel dimensions include tolerances of the rack dimensions (see Table 4.4.4), tolerances Table 4.4.2). In addition to the (see Table 4.4.1) and tolerances of the CEA specifications (see of 0.05 wt% is analyzed. As was tolerances specified in these tables, an enrichment tolerance for Region 1 and Region 2 done to identify the bounding assembly, calculations are performed burnups. The reference racks, and CEAs, at a variety of enrichments, cooling times and To determine the Ak condition is the condition with nominal dimensions and properties.

for the reference condition associated with a specific manufacturing tolerance, the kinf calculated included. All of the Ak values from is compared to the kinf from a calculation with the tolerance be statistically combined (square the various tolerances represent independent effects and may allowance for manufacturing root of the sum of the squares) to determine the final reactivity reactivity) were used in the tolerances. Only the positive Ak values (signifying increasing statistical combination.

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

4-10 1237 Report No. HI-2022940 SHADED AREAS DENOTE PROPRIETARY INFORMATION

4.6.4 Uncertaintyin Depletion Calculationsand 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

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

4.6.6 Effect of Gadolinium added to the fuel.

At higher enrichments, assemblies contain up to 20 rods with Gadolinium (Gd) containing These rods are in specific locations around the control rod guide tubes. Rods Gd in the same assembly.

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

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

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

where all rods were set to a conservatively calculated planar average in slightly higher The calculations performed using the planar average enrichment result assembly enrichment reactivity values than do the calculations performed using the actual 4-12 1237 Report No. HI-2022940 SHADED AREAS DENOTE PROPRIETARY INFORMATION

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

4.6.8 Eccentric Fuel Assembly Positioning of the storage rack cell.

The fuel assembly is assumed to be normally located in the center were positioned in the corner of Nevertheless, MCNP4a calculations assumed the fuel assemblies These calculations indicated the storage rack cell (a four-assembly cluster at closest approach).

1 by up to 0.0 127 delta-k, and that eccentric fuel positioning increases the reactivity of Region the maximum difference in decreases the reactivity in Region 2. For Region 1 calculations, in the final k1f calculations.

reactivity of 0.0127 delta-k is included in the uncertainties Distribution 4.6.9 Reactivity Effect of Axial Burnup and Enrichment cosine power distribution.

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

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

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

235 natural (0.71 wt% U) and low The St. Lucie Unit 1 plant also possesses fuel assemblies with effect the axial burnup distribution.

enriched (2.6 wt% 235U) axial blankets on the ends, which burnup and enrichment variations, and Calculations have been performed for the various axial with an assumed axially constant the results were compared with a reference case, i.e. a case indicate that, as expected, there is a burnup and enrichment. The results of this comparison and enrichment distribution at higher positive reactivity effect from considering the axial burnup for assemblies with enriched burnup and cooling times for non-blanketed assemblies and distribution is considered in the blankets. The effect of the axial burnup and enrichment by conservatively performing calculations that establish burnup vs. enrichment curves, burnup and enrichment distribution, and calculations with both a uniform and non-uniform axial the representative reactivity value.

selecting the higher of the resulting reactivity values as for conservatism.

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

further evaluation is required with these assemblies for 4.6.10 B-10 Depletion in CEAs during full power operation of the core.

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

of the B-10 in the loading of the CEA is used in the analyses. This lack of significant depletion after each reload.

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

effects on reactivity and included in the total uncertainty calculation (Section 4.6.11 Calculationof 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.

235

Fuel enrichments between 1.9 and 4.5 wt% U.

Assemblies with and without axial blankets.

Cooling times between 0 and 20 years The parameter Not all combinations of enrichment and cooling time are of practical relevance.

fuel assemblies combinations which are required to ensure that all current and future discharged burnup vs. enrichment can be safely loaded into the racks are summarized in Table 4.6.4, and indicated that it is curves are determined for these parameter combinations. Prior analysis has in order to necessary to account for the presence of the axial blankets in fuel assemblies demonstrate that all fuel assemblies can be loaded into the racks without credit for Boraflex, Currently, the since these blankets reduce the reactivity of certain high bumup fuel assemblies.

wt%. However, it is minimum enrichment of blanketed assemblies in the pool is about 3.55 in the future as a possible that blanketed assemblies with a lower enrichment could be used range for replacement for a damaged assembly unloaded from the core. The enrichment in Table 4.6.4, to blanketed assemblies has therefore been extended down to 2.5 wt%, as shown the current target cover such assemblies. This assembly average enrichment is close to 4-15 1237 Report No. HI-2022940 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.

are The coefficients of the burnup vs. enrichment curves for all conditions listed in Table 4.6.4 containing shown in Table 4.1.1 for non-blanketed assemblies, and in Table 4.1.2 for assemblies of axial blankets. These tables also provide the required minimum burnup for selected values 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 with analyzed at these conditions. Note that Case 5 is also treated as a checkerboard pattern, but the same burnup vs. enrichment curves for both assemblies in the pattern. The highest maximum limit of keff of any case with any analyzed combination of fuel parameters is below the regulatory It 1.000 applicable when considering no soluble boron to be present in the fuel pool water.

as a should be noted that the calculations contain a significant amount of additional safety margin result of the underlying conservative assumptions, such as:

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

Maximum normal temperature in the pool

Upper bound in-core moderator temperature the entire burnup /

Temperature bias and uncertainties calculated as maximum over enrichment / cooling time range curves.

No interpolation of cooling times allowed between loading calculations, and the embedded The selection of the fuel specifications for the confirmatory pool, under the assumed accident conservatisms will ensure that the actual reactivity of the always be below 1.0. All burnup vs.

condition of the loss of the soluble boron in the pool, will values below the regulatory enrichment curves are therefore acceptable and result in reactivity limit.

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

considered. However, additional special situations are permitted a homogeneous loading of

"* Cells adjacent to the pool walls in Region 2 racks are qualified for as for Case 4.

higher reactivity fuel, with the minimum burnup requirement locations for

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

and any diagonally adjacent fuel assemblies meet certain if an empty

"* A rack module may contain more than one permissible fuel storage configuration stored in another configuration.

row is used to separate fuel in one configuration from fuel across the inter-module gaps, This condition is bounded by the evaluations of the interfaces gaps.

since an empty row is much wider than any of the inter-module effect discussed in the following The results for all calculations of the interface reactivity two standard deviations), or lower subsections are statistically equivalent to (i.e. agree within This agreement demonstrates that than the result of the corresponding reference calculation.

these interface configurations are acceptable.

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

4.6.12.1 Region 1 to Region 1 Interfaces a gap of 1.5 inches, which is The four Region 1 rack modules are separated from each other by It is therefore required that only slightly larger than the cell to cell gap within each rack module.

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

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

effectively modeling an infinite array of racks with highest The results show that this configuration is acceptable.

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

configuration is acceptable.

4-18 1237 Report No. HI-2022940 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 to or All variations of these parameters result in a reactivity value that is statistically equivalent lower than the reference case reactivity, with a Case 1, Case 4 or Case 5 configuration in the on the remainder of the rack. Placing higher reactivity fuel assemblies (Case 4, without CEAs) 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 3

For fuel assembly inspection, testing and reconstitution, it is necessary to place up to As assemblies and a rod basket in close proximity to each other within a Region 2 rack module.

with a bounding approach, these assemblies and the rod basket are modeled as fresh assemblies that an enrichment of 4.5 wt% in the calculations. To produce satisfactory results, it is required fresh fuel the four cells face-adjacent to the cell with a fresh assembly be empty. Additionally, a a fresh assembly must not be placed in a cell diagonally adjacent to another cell containing for assembly. However, it is acceptable to place spent fuel with the highest reactivity permitted 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 above checkerboard, with the fresh assemblies at the closest possible approach consistent with 4-19 1237 Report No. HI-2022940 SHADED AREAS DENOTE PROPRIETARY INFORMATION

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

This The empty cells were modeled with a water density of 25% of the normal water density.

cells as long assumption permits the placement of non-actinide material (i.e., hardware) in these as this non-fuel hardware does not occupy more than 75% of the cell volume.

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

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

soluble boron levels (100 ppm and 300 ppm for Region 1; 200 ppm and is then and the soluble boron concentration necessary to satisfy the regulatory requirement k~ff values, which is determined by linear interpolation. A target of 0.94 is used for the maximum presence of borated water lower, i.e. more conservative, than the regulatory limit. Note that the bias for the Region 2 racks in the fuel pool results in a slightly higher delta-temperature reactivity minimum soluble than would be calculated assuming the presence of pure water. The highest details for this calculation boron concentration calculated is 443 ppm, calculated for Case 4. The of 500 ppm is specified for are shown in Table 4.6.7. For added conservatism, a minimum value compliance purposes, which is larger than the calculated value of 443 ppm.

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

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

accident precludes the necessity of considering the simultaneous occurrence of multiple conditions.

4.6.14.1 Temperature and Water Density Effects been calculated.

The reactivity effect of fuel pool water temperatures exceeding 150 'F has conditions with void Temperatures up to 248 'F (120 C) are evaluated, as are local boiling compared to 150 'F is 0.0303 Ak for percentages up to 20%. The maximum reactivity increase boron concentration Region 1 and 0.0 146 Ak for Region 2. It has been determined that a soluble under these conditions.

of 541 ppm is required to ensure a maximum k1ff of 0.95 is not exceeded 4.6.14.2 Dropped Assembly - Horizontal the dropped assembly In the event a fuel assembly is dropped on top of a storage rack module, distance of at least will come to rest horizontally on top of the rack with a minimum separation to preclude neutron 12 inches from the active region of stored fuel. This distance is sufficient deformation under coupling (i.e., an effectively infinite separation). The maximum expected to less than 12 inches.

seismic or accident conditions will not reduce this minimum spacing increase reactivity Consequently, the horizontal fuel assembly drop accident will not significantly in the fuel storage racks.

4.6.14.3 Dropped Assembly - Vertical by another assembly.

It is also possible to vertically drop an assembly into a location occupied stored assembly, reducing Such a vertical impact would at most cause a small compression of the the reactivity the water-to-fuel ratio and thereby potentially increasing reactivity. However, by the misloading of a fresh increase would be small compared to the reactivity increase created 4-21 1237 Report No. HI-2022940 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.

Report No. HI-2022940 4-22 1237 SHADED AREAS DENOTE PROPRIETARY INFORMATION

4.7 REFERENCES

Handbook 91, August

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

Carlo N-Particle Transport Code,

2. J.F. Briesmeister, Editor, "MCNP - A General Monte (1993).

Version 4A," LA-12625, Los Alamos National Laboratory "CASMO-4 A Fuel Assembly

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

SOA-94/13,

4. D. Knott, "CASMO-4 Benchmark Against Critical Experiments",

Studsvik of America, Inc., (proprietary).

Studsvik of America,

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

for Criticality Analysis of Fuel

6. L.I. Kopp, "Guidance on the Regulatory Requirements Memorandum from L. Kopp to T.

Storage at Light-Water Reactor Power Plants," NRC Collins, August 19, 1998.

presented at the

7. S.E. Turner, "Uncertainty Analysis - Burnup Distributions",

Special Session, ANS/ENS DOE/SANDIA Technical Meeting on Fuel Burnup Credit, Conference, Washington, D.C., November 2, 1988.

4-23 1237 Report No. HI-2022940 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.1.1 Assemblies Minimum Burnup as a Function of Enrichment for Non-Blanketed Minimum Burnups (GWd/MTU) for Coefficientst Case Cooling various Enrichments Time B C 1.9% 2.5% 3.0% 3.8%

A

-16.52 19.29 29.62 37.87 50.40 0 years -0.65 20.08 Case 1, -15.58 15.82 24.76 31.85 42.52 12 years -0.65 17.76 Low -13.84 15.48 24.10 31.04 41.70 15 years -0.43 16.25 Reactivity -9.61 15.33 23.39 30.17 41.14 20 years 0.12 12.90

-21.39 9.43 18.55 25.92 37.29 0 years -0.41 17.00 Case 1, -20.63 8.24 16.55 23.17 33.21 12 years -0.54 16.22 High -20.07 8.15 16.27 22.74 32.54 15 years -0.53 15.86 Reactivity 15.11 -18.80 8.25 16.10 22.39 31.98 20 years -0.46 17.49 -19.72 10.84 19.38 26.09 36.06

-0.74 33.70 Case 2, 0 years 15.64 -17.65 10.04 17.95 24.23 5 years -0.56 Low Reactivity Case 2, 0 years 0.00 9.31 -24.39 0.00 0.00 F 10.99 High Reactivity Case 3, 0 years 0.00 10.97 -14.71 6.13 12.72 18.20 26.98 Low Reactivity 10.51 -22.35 0.00 3.93 9.18 17.59 Case 3, 0 years 0.00 High Reactivity 12.72 18.20 26.98 0 years 0.00 10.97 -14.71 6.13 Case 4 17.70 -17.97 14.18 23.72 31.44 43.37 0 years -0.41 27.10 37.80 Case 5 12.47 20.44 0.04 13.10 -12.56 12 years 19.93 26.48 37.09 0.13 12.38 -11.83 12.16 15 years 19.37 25.86 36.52 0.26 11.56 -11.16 11.74 2 B*E + C with BU = Minimum Burnup in Coefficients for polynomial Function: BU = A*E +

235 U; A, B, C = Coefficients GWD/MTU; E = Initial Enrichment in wt%

4-24 1237 Report No. HI-2022940 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.1.2 Minimum Burnup as a Function of Enrichment for Blanketed Assemblies Case Cooling Coefficients t Minimum Burnups (GWd/MTU) for Time various Enrichments 3.0% 3.5% 4.0% 4.5%

A B C 2.5 %

19.25 -13.42 29.46 36.77 43.67 50.14 56.20 Case 1, 0 years -0.84

-0.72 17.40 -12.03 26.97 33.69 40.05 46.05 51.69 51.69 Low 5 years 43.26 48.62 16.32 -11.46 25.22 31.56 37.58 Reactivity 10 years -0.66 24.08 30.24 36.06 41.55 46.70 15 years -0.67 16.00 -11.73 35.46 -45.8 40.83 45.83 20 years -0.76 16.45 -12.81 23.57 29.70 18.76 25.55 31.85 37.66 42.98 0 years -0.98 18.97 -22.54 Case 1, 23.86 29.73 35.22 40.35

-0.74 16.54 -19.10 17.63 High 5 years 33.31 38.25 14.73 -16.49 16.77 22.57 28.08 Reactivity 10 years -0.57

-0.46 -14.70 16.28 21.78 27.06 32.10 32.10 36.92 15 years 13.54 26.67 31.62 36.37

-0.41 12.98 -13.74 16.15 21.51 20 years 32.43 38.40 44.00

-0.74 17.49 -19.72 19.38 26.09 Case 2, 0 years 35.95 41.39 5 years -17.65 17.95 24.23 30.23 Low -0.56 15.64 Reactivity Case 2, 0 years 0.00 9.31 -24.39 0.00 3.54 8.20 12.85 17.51 High Reactivity Case 3, 0 years 0.00 10.97 -14.71 12.72 18.20 23.69 29.17 34.66 Low Reactivity Case 3, 0 years 0.00 10.51 -22.35 3.93 9.18 14.44 19.69 24.95 High Reactivity 23.69 29.17 34.66 0.00 10.97 -14.71 12.72 18.20 Case 4 0 years 36.49 42.70 48.80

-0.24 14.23 -10.38 23.70 30.15 Case 5 0 years 34.16 39.96 45.66

-0.20 13.10 -9.24 22.26 28.26 5 years 37.85 43.22 21.04 26.76 32.36 10 years -0.23 12.70 -9.27 25.70 31.17 36.48 41.63 15 years -0.32 13.02 -10.48 20.07 25.16 30.67 35.95 40.99 20 years -0.47 14.08 -12.85 19.41 in Coefficients for polynomial Function: BU = A*E + B*E + C with BU = Minimum Burnup 2

235U; A, B, C = Coefficients GWD/MTU; E = Initial Enrichment in wt%

4-25 1237 Report No. HI-2022940 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.4.1 St. Lucie Unit 1 Fuel Assembly Specifications Value Parameter CE 14xl4 FR 14xl4 Assembly type 14x14 14x14 Rod Array Size 0.580 M 0.580 Rod Pitch, Inches 136.7 136.7 Maximum Active Fuel Length, Inches 10.05 10.30 Stack Density (g/cm 3 )

176 176 Total Number of Fuel Rods 0.440 0.440 Fuel Rod Outer Diameter, Inches 0.026 - 0.028 0.028 - 0.031 Cladding Thickness, Inches Zr-4 Zr-4 Cladding Material 0.3805 0.3770 Maximum Pellet Diameter, Inches 5 5 Number of Guide Tubes 1.115 1.115 Guide Tube Outer Diameter, Inches 0.04 0.04 Guide Tube Wall Thickness, Inches Zr-4 Zr-4 Guide Tube Material 4-26 1237 Report No. HI-2022940 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.4.2 Control Element Assembly (CEA) Specifications 1237 4 4-27 Report No. HI-20229 0 INFORMATION SHADED AREAS DENOTE PROPRIETARY

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

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

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

Table 4.6.2 Effect of Pool Water Temperature on kinf ppm Soluble Boron.

for Fuel of 4.5 wt% Enrichment and 0 Years Cooling Time at 0 Relative to 20 'C 4-31 1237 Report No. HI-2022940 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.6.3 Reactivity Effect of Rack and Fuel Tolerances 4-32 1237 Report No. HI-2022940 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) 1.9-3.8 0 2.5-4.5 0 3

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 4-33 1237 Report No. HI-2022940 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 4-34 1237 Report No. HI-2022940 SHADED AREAS DENOTE PROPRIETARY INFORMATION

Table 4.6.6 Case Results of Additional Selected Calculations for each

= Cooling Time in years;

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

Table 4.6.7 Value of 0.95 under Normal Conditions.

Soluble Boron Concentration for a Maximum keff 1237 4 4-36 Report No. HI-20229 0 INFORMATION SHADED AREAS DENOTE PROPRIETARY

Table 4.6.8 0.95 under Accident Conditions.

Soluble Boron Concentration for a Maximum k1f Value of 1500 Case m 1000 2 Reion 2 4.5% Fresh Fuel k-inf 1.1098 Calculated k-eff 1.1824 Assembly I 1.9 Enrichment 1.9 15.3 Burnup 15.3 20 Cooling Time 20 Assembly 2 1.9 Enrichment 1.9 9.4 Burnup 9.4 0 Coolin2 Time 0 0.8617 Calculated k-eff 0.9223 0.0009 Bias 0.0009 Temperature Correction 0.0068 0.0073 Uncertainties 0.0011 Bias 0.0011 0.0012 Calculational 0.0014 0.0000 Eccentricity 0.0000 0.0075 Assembly Burnup 0.0075 0.0124 Depletion 0.0130 0.0145 Tolerances 0.0145 0.0206 Total Uncertainties 0.0209 0.0288 Total Addition 0.0286 0.8905 Maximum k-eff 0.9509 0.94 Target k-max 1090 corresponding soluble boron level 4-37 1237 Report No. HI-2022940 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) 4-38 1237 Report No. HI-2022940 SHADED AREAS DENOTE PROPRIETARY INFORMATION

8.86

'W4 Figure 4.4.2: Schematic View of Region 2 Cell (not to scale) 4-39 1237 Report No. HI-2022940 SHADED AREAS DENOTE PROPRIETARY EINFORMATION

Case 1, Low Reactivity

--4--0 years years 40.00 40.00--W--12 1~

S15 years 20 years 30 0 years, Blankets S30.00 -o--5 years, Blankets SC Fig10 years, Blankets o 41- 15 years, Blanketsi

20 years, Blankets 20.00 10.00 0.00 3 3.5 4.5 1.5 2 2.5 Initial Enrichment, wt%

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

Case 1, High Reactivity 60.00 50.00 s 0 years 40.00 ---- 12 years 15 years 20 years

)K 0 years, Blankets 0 30.00 --6--5 years, Blankets d -- i-- 10 years, Blankets I-1 - 15 years, Blankets 20 years, Blankets 20.00 10.00 0.00 2 2.5 3 3.5 4 4.5 5 1.5 Initial Enrichment, wt%

Figure 4.6.2 Minimum Burnup as a Function of Initial Enrichment for Case 1, High Reactivity Assemblies II L.

- - 7I Report No. HI-2022940 4-41 I Z-- /

SHADED AREAS DENOTE PROPRIETARY INFORMATION

Case 2 60.00 50.00

40.00 4 Low Reactivity, 0 years Low Reactivity, 5 years

= 30.00 High Reactivity 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 IL.j I 4-42 I Z.J/

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

Case 3 Low Reactivity LH:-:High Reactivity:

E E

3.5 4 4.5 5 1.5 2 2.5 3 Initial Enrichment, wt%

Figure 4.6.4 Minimum Burnup as a Function of Initial Enrichment for Case 3 4-43 1237 Report No. HI-2022940 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 4 4.5 5 1.5 2 2.5 3 3.5 Initial Enrichment, wt%

Figure 4.6.5 Minimum Burnup as a Function of Initial Enrichment for Case 4 1L31 I

Report No. HI-2022940 4-44 IJL./

SHADED AREAS DENOTE PROPRIETARY INFORMATION

Case 5 60.00 50.00 40.00 -- 0 years 12 years 15 years 20 years C

0 years, Blankets O 30.00 -- o- 5 years, Blankets

--- 10 years, Blankets w - 15 years, Blankets 20.00 20 years, Blankets.]

10.00 0.00 2.5 3 3.5 4 4.5 5 1.5 2 Initial Enrichment, wt%

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

SHADED AREAS DENOTE PROPRIETARY INFORMATION

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

H High Reactivity -LH.¸ --L ,-L H -H L H L H L H IL ll L H L Low Reactivity L H L H L H L H F Fresh Assembly / Rod Basket

_H H H L H LI Empty Cell HX F X F H L H

-.... Reflective Boundary Condition H FXFH H L H IL H SH H L H LH L H LHtHL H

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

2 Racks for Inspection and Figure 4.6.7 Schematic Configuration of the Calculational Model for Fresh Assemblies in Region Reconstitution 4-46 1237 Report No. HI-2022940 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".

LL3 I Report No. HI-2022940 1237

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=1n j ki (4A.1)

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

4A.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.5 MOX Fuel The number of critical experiments with PuO 2 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 U0 2 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~5 U 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. EALF t (eV)

MCNP4a KENO5a U Identification Enrich. MCNP4a KENO5a 2.46 0.9964 +/- 0.0010 0.9898+/- 0.0006 0.1759 0.1753 1 B&W-1484 (4A.7) Core I 2.46 1.0008 +/- 0.0011 1.0015 +/- 0.0005 0.2553 0.2446 2 B&W-1484 (4A.7) Core II 2.46 1.0010 +/- 0.0012 1.0005 +/- 0.0005 0.1999 0.1939 3 B&W-1484 (4A.7) Core Ell 2.46 0.9956 +/- 0.0012 0.9901 +/- 0.0006 0.1422 0.1426 4 B&W-1484 (4A.7) Core IX 2.46 0.9980 +/- 0.0014 0.9922 +/- 0.0006 0.1513 0.1499 5 B&W-1484 (4A.7) Core X 2.46 0.9978 +/- 0.0012 1.0005 +/- 0.0005 0.2031 0.1947 6 B&W-1484 (4A.7) Coie X 2.46 0.9988 +/- 0.0011 0.9978 +/- 0.0006 0.1718 0.1662 7 B&W-1484 (4A.7) Core XII 2.46 1.0020 +/- 0.0010 0.9952 +/- 0.0006 0.1988 0.1965 8 B&W-1484 (4A.7) Core XII 2.46 0.9953 +/- 0.0011 0.9928 +/- 0.0006 0.2022 0.1986 9 B&W-1484 (4A.7) Core XIV Core XV I 2.46 0.9910 +/- 0.0011 0.9909 +/- 0.0006 0.2092 0.2014 10 B&W-1484 (4A.7)

Core XVI tV 2.46 0.9935 +/- 0.0010 0.9889 +/- 0.0006 0.1757 0.1713 11 B&W-1484 (4A.7)

Core XVII 2.46 0.9962 +/- 0.0012 0.9942 +/- 0.0005 0.2083 0.2021 12 B&W-1484 (4A.7)

Core XVMI 2.46 1.0036 +/- 0.0012 0.9931 +/- 0.0006 0.1705 0.1708 13 B&W-1484 (4A.7)

'b--,9A .. .. 1° A A fl..,, i Appendhxwtjage.

Table 4A.1 Summary of Criticality Benchmark Calculations Calculatedlkf EALFt (eV-)

KENO5a MCNP4a KENO5a Oaf aranrn Ird~ntifleatonn Enrich. MCNP4a Core XIX 2.46 0.9961 +/- 0.0012 0.9971 +/- 0.0005 0.2103 0.2011 14 B&W-1484 (4A.7)

Core XX 2.46 1.0008 +/- 0.0011 0.9932 +/- 0.0006 0.1724 0.1701 15 B&W-1484 (4A.7) 2.46 0.9994 +/- 0.0010 0.9918 +/- 0.0006 0.1544 0.1536 16 B&W-1484 (4A.7) Core XXI 2.46 0.9970 +/- 0.0010 0.9924 +/- 0.06 1.4475 1.4680 17 B&W-1645 (4A.8) S-type Fuel, w/886 ppm B 2.46 0.9990 +/- 0.0010 0.9913 +/- 0.0006 1.5463 1.5660 18 B&W-1645 (4A.8) S-type Fuel, w/746 ppm B 2.46 0.9972 +/- 0.0009 0.9949 +/- 0.0005 0.4241 0.4331 19 B&W-1645 (4A.8) SO-type Fuel, w/1156 ppm B 2.46 1.0023 +/- 0.0010 NC 0.1531 NC 20 B&W-1810 (4A.9) Case 1 1337 ppm B 2.46/4.02 1.0060 +/- 0.0009 NC 0.4493 NC 21 B&W-1810 (4A.9) Case 12 1899 ppm B 4.75 0.9966 +/- 0.0013 NC 0.2172 NC 22 Firench (4A.10) Water Moderator 0 gap 4.75 0.9952 +/- 0.0012 NC 0.1778 NC 23 French (4A.10) Water Moderator 2.5 cm gap 4.75 0.9943 +/- 0.0010 NC 0.1677 NC 24 French (4A.10) Water Moderator 5 cm gap 4.75 0.9979 +/- 0.0010 NC 0.1736 NC 25 French (4A.10) Water Moderator 10 cm gap 2.35 NC 1.0004 +/- 0.0006 NC 0.1018 26 PNL-3602 (4A.11) Steel Reflector, 0 separation Appendix 4A, Page 10

Table 4A.1 Summary of Criticality Benchmark Calculations Calculated kif EALF t (eV)

Enrich. MCNP4a KENO5a Tudpntflpgtinf~r 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 Steel Reflector, Infinite sepn. 2.35 0.9962 +/- 0.0008 0.9939 +/- 0.0006 0.0973 0.0968 30 PNL-3602 (4A.11)

Steel Reflector, 0 cm sepn. 4.306 NC 1.0003 +/- 0.0007 NC 0.3282 31 PNL-3602 (4A.11)

Steel Reflector, 1.321 cm sepn. 4.306 0.9997 +/- 0.0010 1.0012 +/- 0.0007 0.3016 0.3039 32 PNL-3602 (4A.11)

Steel Reflector, 2.616 cm sepn. 4.306 0,9994 +/- 0.0012 0.9974 +/- 0.0007 0.2911 0.2927 33 PNL-3602 (4A.11)

Steel Reflector, 5.405 cm sepn. 4.306 0.9969 +/- 0.0011 0.9951 +/- 0.0007 0.2828 0.2860 34 PNL-3602 (4A.11) 35 PNL-3602 (4A.11) Steel Reflector, Infinite sepn. It 4.306 0.9910 +/- 0.0020 0.9947 +/- 0.0007 0.2851 0.2864 Steel Reflector, with Boral Sheets 4.306 0.9941 +/- 0.0011 0.9970 +/- 0.0007 0.3135 0.3150 36 PNL-3602 (4A.11)

Lead Reflector, 0 cm sepn. 4.306 NC 1.0003 +/- 0.0007 NC 0.3159 37 PNL-3926 (4A.12)

Lead Reflector, 0.55 cm sepn. 4.306 1.0025 +/- 0.0011 0.9997 +/- 0.0007 0.3030 0.3044 38 PNL-3926 (4A.12) 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,_ EALF ' (eV)

Reference Identification Enrich. MCNP4a KENO5a 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

Table 4A.1 Summary of Criticality Benchmark Calculations Calculated kf EALFt (eV)

KENO5a MCNP4a KENO5a Reference Identification Enrich. MCNF4a .

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 53 MOX Fuel - Type 3.2 Exp. 13 20% Pu 1.0083 +/- 0.0011 0.9989 +/- 0.0006 0.1665 0.1706 54 PNL-5803 (4A.16) 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 Saxton Case 56 U 0.56" pitch 5.74 0.9994 +/- 0.0011 0.9967 +/- 0.0007 0.2923 0.2954 60 WCAP-3385 (4A.17) 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

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 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 tEALF is the energy of the average lethargy causing fission.

Appendix 4A, Page 15

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 1.010 I

1.005 N-I 1.000 0

0.995 0.990 0.1 1 Energy of Average Lethargy Causing Fission (Log Scale)

FIGURE 4A.1 MCNP CALCULATED k-eff VALUES for VARIOUS VALUES OF THE SPECTRAL INDEX

Linear Regression with Correlation Coefficient of 0.21 1.010 1.005 4.-

1.000

-3 0.995 C,

0 0.990 0.985 0.1 1 Energy of Average Lethargy Causing Fission (Log Scale)

FIGURE 4A.2 KENO5a CALCULATED k-eff VALUES FOR VARIOUS VALUES OF THE SPECTRAL INDEX

-- -Linear Regression with Correlation Coefficient of 0.03

__________ r r r 1.010 I) T 4 + + ti 1,005 (D ..(

u i)

-o N- 1.000 4 1 1---t C)

"0 g F! 4t AP1rl ,:

t11

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

+/-

J CO 03 C.,

0.995 -- ______________ I I I T 'f 1

) (

z z

)

I I-J-4 0.990 I - I ---- i -I ]-I I f 1 1 1 I !I I II I I1 -1 i I ii i i i i i i i i F1 11 1 1!

4.0 4.5 5.0 5.5 6.0

2. 0 2.5 3.0 3.5 Enrichment, w/o U-235 FIGURE 4A.3 MCNP CALCULATED k eff VALUES AT VARIOUS U-235 ENRICHMENTS

Linear Regression with Correlation Coefficient of 0.38 1.010 1,005 0*1.000

"-4 4

2 0.995

-o 0.990 0.985 Enrichment, w/o U-235 FIGURE 4A.4 KENO CALCULATED. k-eff VALUES AT VARIOUS U-235 ENRICHMENTS

0.94 E 0.92 C,)

0 0.90 70 C) 0 u) 0 0.88 In 0

z ILJ 0.86 0.84 MCNP k-eff Calculations FIGURE 4A.5 COMPARISON OF MCNP AND KENO5A CALCULATIONS FOR VARIOUS FUEL ENRICHMENTS

1.04-1.03 1.01 C

3: 0.99

~0.98 "0

S 0.97

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

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

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

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

performed rack seismic/structural evaluations and reported results remain 6-1 1237 Holtec Report HI-2022940 SHADED AREAS DENOTE PROPRIETARY INFORMATION

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

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

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

accident evaluations for postulated fuel drops and the associated 1237 7-1 Holtec Report HI-2022940 SHADED AREAS DENOTE PROPRIETARY INFORMATION

8.0 POOL STRUCTURE ASSESSMENT The proposed change does not require any physical modifications to fuel, storage racks, or plant will structural systems. No new equipment is required to be installed. Changes in water coolant density All loading not significantly affect any of the evaluations previously performed for the pool structure.

previously conditions and load combinations previously considered remain valid. Therefore, the performed rack pool structure evaluations and reported results remain valid.

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

9.0 RADIOLOGICAL CONSIDERATIONS modifications to fuel or to the fuel storage racks.

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

within the pool.

not in any way affect the soluble boron already present and the minimum allowed post-irradiation fuel The maximum allowed fuel enrichment, fuel rod burnup, fuel fuel source terms remain unchanged. The revised cooling time remain unchanged. Thus, the spent fuel.

the location of source terms represented by spent storage configuration will not significantly affect exterior walls will not change. Thus, the radiation The proximity of fuel to the pool water surface and the water inventory remains unchanged. Therefore, attenuation provided by the walls and the fuel pool their reported results remain valid. Dose levels previously performed radiological evaluations and significantly after implementing the proposed change.

surrounding the SFP are not expected to change 1237 9-1 Holtec Report HI-2022940 SHADED AREAS DENOTE PROPRIETARY INFORMATION

10.0 ENVIRONMENTAL/COST BENEFIT ASSESSMENT benefit assessment (enclosure of the Florida Power and Light has prepared an environmental and cost This assessment examines the underlying 10CFR50.92 evaluation) of the proposed license amendment.

at St. Lucie Unit 1 and it considers need for actions to mitigate the consequences of Boraflex degradation the proposed change. This assessment also the thermal and radiological impacts on the environment of the proposed license amendment is considers the occupational exposure that will be incurred as methods of managing the storage of implemented. In the assessment FPL identified several alternative economic consequences of each candidate irradiated nuclear fuel and it examined the environmental and of a "no action" alternative.

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

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

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

Contents

Text

SUMMARY

SUMMARY

1.0 INTRODUCTION

1.0 INTRODUCTION

1.0 INTRODUCTION

4.7 REFERENCES

SUMMARY

Navigation menu

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

Text

SUMMARY

SUMMARY

1.0 INTRODUCTION

1.0 INTRODUCTION

1.0 INTRODUCTION

4.7 REFERENCES

SUMMARY

Navigation menu

Search