ML18310A154
| ML18310A154 | |
| Person / Time | |
|---|---|
| Site: | NuScale |
| Issue date: | 11/05/2018 |
| From: | Bergman T NuScale |
| To: | Document Control Desk, Office of New Reactors |
| Shared Package | |
| ML18310A153 | List: |
| References | |
| AF-1018-61992, LO-1018-61991 TR-0816-49833, Rev 1 | |
| Download: ML18310A154 (148) | |
Text
LO-1018-61991 November 5, 2018 Docket No.52-048 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738
SUBJECT:
NuScale Power, LLC Submittal of "Fuel Storage Rack Analysis," TR-0816-49833, Revision 1
REFERENCES:
- 1. Letter from NuScale Power, LLC to U.S. Nuclear Regulatory Commission, "Submittal of Technical Report 'Fuel Storage Rack Analysis," dated December 30, 2016 (ML17005A112)
- 2. NuScale Technical Report, "Fuel Storage Rack Analysis,"
Revision 0, TR-0816-49833, dated September 2016 (ML17005A128)
NuScale Power, LLC (NuScale) hereby submits Revision 1 of the "Fuel Storage Rack Analysis," TR-0816-49833, Revision 1. The purpose of this submittal is to request that the NRC review and approve the methodology and results presented in the report. As noted within the report, the structural analysis has been intentionally omitted. The report, as well as corresponding changes within Rev 2 of the DCA, indicate that the COL applicant shall provide the NRC a structural evaluation of the spent fuel storage racks which is dependent on a vendor specific design and the as-built configuration of spent fuel storage racks. contains the proprietary version of the report entitled "Fuel Storage Rack Analysis,"
Revision 1. NuScale requests that the proprietary version be withheld from public disclosure in accordance with the requirements of 10 CFR § 2.390. The enclosed affidavit (Enclosure 3) supports this request. Enclosure 2 contains the non proprietary version of the report entitled "Fuel Storage Rack Analysis," Revision 1.
This letter makes no regulatory commitments and no revisions to any existing regulatory commitments.
If you have any questions, please contact Carrie Fosaaen at 541-452-7126 or at cfosaaen@nuscalepower.com if you have any questions.
Sincerely, Thomas Distribution: Samuel Lee, NRC, OWFN-8G9A Greg Cranston, NRC, OWFN-8G9A Getachew Tesfaye, NRC, OWFN-8G9A NuScale Power, LLC 1100 NE Circle Blvd , Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com
LO-1018-61991 Page 2 of 2 11/05/2018 Enclosure 1: Fuel Storage Rack Analysis, TR-0816-49833-P, Revision 1, proprietary version Enclosure 2: Fuel Storage Rack Analysis, TR-0816-49833-NP, Revision 1, nonproprietary version Enclosure 3: Affidavit of Thomas A. Bergman, AF-1018-61992 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com
LO-1018-61991 :
Fuel Storage Rack Analysis, TR-0816-49833-P, Revision 1, proprietary version NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com
LO-1018-61991 :
Fuel Storage Rack Analysis, TR-0816-49833-NP, Revision 1, nonproprietary version NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Licensing Technical Report Fuel Storage Rack Analysis November 2018 Revision 1 Docket: 52-048 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 www.nuscalepower.com
© Copyright 2018 by NuScale Power, LLC
© Copyright 2018 by NuScale Power, LLC i
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Licensing Technical Report COPYRIGHT NOTICE This report has been prepared by NuScale Power, LLC and bears a NuScale Power, LLC, copyright notice. No right to disclose, use, or copy any of the information in this report, other than by the U.S.
Nuclear Regulatory Commission (NRC), is authorized without the express, written permission of NuScale Power, LLC.
The NRC is permitted to make the number of copies of the information contained in this report that is necessary for its internal use in connection with generic and plant-specific reviews and approvals, as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by NuScale Power, LLC, copyright protection notwithstanding. Regarding nonproprietary versions of these reports, the NRC is permitted to make the number of copies necessary for public viewing in appropriate docket files in public document rooms in Washington, DC, and elsewhere as may be required by NRC regulations. Copies made by the NRC must include this copyright notice and contain the proprietary marking if the original was identified as proprietary.
© Copyright 2018 by NuScale Power, LLC ii
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Licensing Technical Report Department of Energy Acknowledgement and Disclaimer This material is based upon work supported by the Department of Energy under Award Number DE-NE0008820.
This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof.
© Copyright 2018 by NuScale Power, LLC iii
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Licensing Technical Report CONTENTS Abstract ........................................................................................................................... 1 Executive Summary ........................................................................................................ 2 1.0 Introduction ........................................................................................................... 3 1.1 Abbreviations ............................................................................................. 3 1.2 General Description of the Fuel Storage Racks ......................................... 4 1.3 Purpose ................................................................................................... 17 2.0 Background ........................................................................................................ 18 2.1 Analysis Application ................................................................................. 18 2.2 Codes, Standards, and Regulatory Requirements ................................... 18 3.0 Analysis .............................................................................................................. 20 3.1 Structural Analysis (Intentionally Omitted) ............................................... 20 3.1.1 Fuel Storage Detailed and Simplified Structural Model Development (Intentionally Omitted) ............................................. 20 3.1.2 Time-History Development for Fuel Storage Rack Floor Response Spectra (Intentionally Omitted) ........................... 20 3.1.3 Load-Drop Analysis for Fuel Racks (Intentionally Omitted) ........................................................................................ 20 3.1.4 Whole-Pool Analysis and Fuel Storage Rack Interaction (Intentionally Omitted) ................................................. 20 3.1.5 Fuel Storage Rack Detailed Stress Analysis and Design (Intentionally Omitted) ....................................................... 20 3.1.6 Sensitivity Analysis - Whole-Pool Analysis with Partially Loaded Racks and Varying Friction Coefficients (Intentionally Omitted) ............................................... 20 3.2 Thermal-Hydraulic Analysis ..................................................................... 20 3.2.1 Maximum SFP Bulk Temperature Calculation .............................. 21 3.2.2 Spent Fuel Pool Computational Fluid Dynamics Analysis ......................................................................................... 23 3.3 Criticality Analysis .................................................................................... 68 3.3.1 Methodology ................................................................................. 69 3.3.2 Assumptions ................................................................................. 73 3.3.3 Configuration ................................................................................. 74
© Copyright 2018 by NuScale Power, LLC iv
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Licensing Technical Report 3.3.4 Initial Conditions, Boundary Conditions, and Limitations ..................................................................................... 77 3.3.5 Software Use and Qualification ..................................................... 79 3.3.6 Analysis, Evaluation, and Data...................................................... 80 3.3.7 Benchmark Analysis ...................................................................... 90 3.3.8 Summary of Criticality Evaluations .............................................. 121 3.4 Material Analysis .................................................................................... 122 3.4.1 Material Evaluation for NuScale Fuel Racks ............................... 122 4.0 Summary and Conclusions ............................................................................... 128
5.0 REFERENCES
................................................................................................. 129 5.1 SOURCES OF INFORMATION RELIED UPON .................................... 129
5.2 REFERENCES
CITED........................................................................... 130
© Copyright 2018 by NuScale Power, LLC v
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Licensing Technical Report TABLES Table 1-1 Abbreviations ................................................................................................... 3 Table 1-2 Definitions ........................................................................................................ 4 Table 3-1 Heat exchanger data sheet summary ............................................................ 22 Table 3-2 Analysis cases and peak liquid temperatures ............................................... 29 Table 3-3 Fuel assembly specification for criticality analysis ......................................... 71 Table 3-4 Fuel rack design parameters for criticality analysis ....................................... 72 Table 3-5 Material properties ......................................................................................... 73 Table 3-6 System sensitivity to nominal temperature .................................................... 78 Table 3-7 Operational tolerance impact on neutron multiplication ................................. 79 Table 3-8 Spent fuel pool tolerance analysis with unborated moderator ....................... 81 Table 3-9 SFP tolerance analysis with moderator at 800 ppm dissolved boron ............ 83 Table 3-10 Fuel storage damaged fuel analysis .............................................................. 85 Table 3-11 Factors for k95/95 calculation for fuel storage .................................................. 86 Table 3-12 Spent fuel pool dropped fuel assembly analysis ........................................... 87 Table 3-13 k-effective and k95/95 for a seismic event in the fuel storage .......................... 88 Table 3-14 Parameter range for critical experiment selection ......................................... 91 Table 3-15 Benchmark experiments selected for enrichment trend ................................ 97 Table 3-16 Benchmark experiments selected for fuel rod pitch trend ............................. 98 Table 3-17 Benchmark experiments selected for fuel assembly separation trend .......... 99 Table 3-18 Benchmark experiments selected for soluble boron trend ............................ 99 Table 3-19 Benchmark experiments selected for separator plate boron areal density trend ................................................................................................ 100 Table 3-20 Benchmark experiments selected for moderator to fuel area ratio trend ............................................................................................................ 101 Table 3-21 Benchmark experiments selected for neutron spectrum ............................. 101 Table 3-22 Critical experiment parameters and KENO-V.a results ............................... 102 Table 3-23 Regression analysis for possible bias trending variables ............................ 108 Table 3-24 Data used for bias and bias uncertainty ...................................................... 117 Table 3-25 Area of applicability for bias and bias uncertainty ....................................... 121 Table 3-26 Summary of criticality analysis results ......................................................... 122 Table 3-27 NuScale spent fuel pool water chemistry .................................................... 123
© Copyright 2018 by NuScale Power, LLC vi
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Licensing Technical Report FIGURES Figure 11 Fuel storage rack general configuration .......................................................... 7 Figure 11a Bottom Grid ..................................................................................................... 8 Figure 11b Bottom Grid, Cross Bars, Outer Grid Lower Bands, and Corner Posts ........... 9 Figure 11c Mid Band Assembly ....................................................................................... 10 Figure 11d Top Inner Grid ............................................................................................... 11 Figure 11e Top Outer Grid............................................................................................... 12 Figure 12 Spent fuel pool general arrangement of racks ............................................... 13 Figure 12a Cross-sectional view looking north ................................................................ 14 Figure 12b Cross-sectional view looking west ................................................................. 15 Figure 13 Fuel storage rack width ................................................................................. 16 Figure 14 Fuel storage rack cross section with fuel assembly ...................................... 17 Figure 31 Spent fuel pool heat exchanger concept ....................................................... 22 Figure 32 Top view with rack numbers .......................................................................... 28 Figure 33 Temperature contour at the upper surface of the active fuel region, case 1 ............................................................................................................ 30 Figure 34 Temperature contour at the lower surface of the active fuel region, case 1 ............................................................................................................ 31 Figure 35 Temperature contour on vertical planes, case 1 ........................................... 32 Figure 36 Streamlines from the cooling water inlets, case 1 ......................................... 33 Figure 37 Streamlines from the upper surface of the active fuel region, case 1 ............ 34 Figure 38 Plot of the solution residuals, case 1 ............................................................. 35 Figure 39 Plot of the solution peak spent fuel pool temperature, case 1 ....................... 36 Figure 310 Temperature contour at the upper surface of the active fuel region, case 2 ............................................................................................................ 37 Figure 311 Temperature contour at the lower surface of the active fuel region, case 2 ............................................................................................................ 38 Figure 312 Temperature contour on vertical planes, case 2 ........................................... 39 Figure 313 Streamlines from the cooling water inlets, case 2 ......................................... 40 Figure 314 Streamlines from the upper surface of the active fuel region, case 2 ............ 41 Figure 315 Plot of the solution residuals, case 2 ............................................................. 42 Figure 316 Plot of the solution peak spent fuel pool temperature, case 2 ....................... 43 Figure 317 Temperature contour at the upper surface of the active fuel region, case 3 ............................................................................................................ 44
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Licensing Technical Report Figure 318 Temperature contour at the lower surface of the active fuel region, case 3 ............................................................................................................ 45 Figure 319 Temperature contour on vertical planes, case 3 ........................................... 46 Figure 320 Streamlines from the cooling water inlets, case 3 ......................................... 47 Figure 321 Streamlines from the upper surface of the active fuel region, case 3 ............ 48 Figure 322 Plot of the solution residuals, case 3 ............................................................. 49 Figure 323 Plot of the solution peak spent fuel pool temperature, case 3 ....................... 50 Figure 324 Temperature contour at the upper surface of the active fuel region, case 4 ............................................................................................................ 51 Figure 325 Temperature contour at the lower surface of the active fuel region, case 4 ............................................................................................................ 52 Figure 326 Temperature contour on vertical planes, case 4 ........................................... 53 Figure 327 Streamlines from the cooling water inlets, case 4 ......................................... 54 Figure 328 Streamlines from the upper surface of the active fuel region, case 4 ............ 55 Figure 329 Plot of the solution residuals, case 4 ............................................................. 56 Figure 330 Plot of the solution peak spent fuel pool temperature, case 4 ....................... 57 Figure 331 Temperature contour at the upper surface of the active fuel region, case 5 ............................................................................................................ 58 Figure 332 Temperature contour at the lower surface of the active fuel region, case 5 ............................................................................................................ 59 Figure 333 Temperature contour on vertical planes, case 5 ........................................... 60 Figure 334 Streamlines from the cooling water inlets, case 5 ......................................... 61 Figure 335 Streamlines from the upper surface of the active fuel region, case 5 ............ 62 Figure 336 Plot of the solution residuals, case 5 ............................................................. 63 Figure 337 Plot of the solution peak spent fuel pool temperature, case 5 ....................... 64 Figure 338 Velocity in active fuel region of rack 11 for case 3 ........................................ 65 Figure 339 Mesher settings and statistics for 3 mesh .................................................... 66 Figure 340 KENO-V.a model of single rack location in fuel storage ................................ 75 Figure 341 KENO-V.a model of a single fuel storage rack .............................................. 76 Figure 342 KENO-V.a model of the spent fuel pool......................................................... 77 Figure 343 System sensitivity to nominal temperature .................................................... 79 Figure 344 k95/95 for a seismic event in the fuel storage .................................................. 89 Figure 345 Regression analysis of U-235 enrichment ................................................... 110 Figure 346 Regression analysis of fuel rod pitch ........................................................... 111 Figure 347 Regression analysis of fuel assembly separation ........................................ 112
© Copyright 2018 by NuScale Power, LLC viii
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Licensing Technical Report Figure 348 Regression analysis of dissolved boron concentration ................................ 113 Figure 349 Regression Analysis of 10B Area Density in Absorber Plates ...................... 114 Figure 350 Regression analysis of moderator to fuel area ratio .................................... 115 Figure 351 Regression analysis of neutron spectrum ................................................... 116
© Copyright 2018 by NuScale Power, LLC ix
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Abstract This technical report presents a summary of the documents, analytical inputs, interpretations, and methodologies used to design and analyze the fuel storage racks as a basis to demonstrate compliance with the applicable regulations to support the NuScale design certification. This report considers regulatory requirements in the areas of nuclear criticality, thermal hydraulics, and materials analysis in the fuel storage rack design. Structural Analysis has intentionally been omitted.
© Copyright 2018 by NuScale Power, LLC 1
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Executive Summary This technical report contains a summary of the documents, analytical inputs, interpretations, and methodologies used to design and analyze the fuel storage racks as a basis to demonstrate compliance with the applicable regulations to support the NuScale Power design certification.
Regulatory Guide 1.29 (Reference 1) classifies fuel storage racks as Seismic Category I structures. The fuel storage racks are designed per 10 CFR 50, Appendix B (Reference 3) and ASME NQA-1 (Reference 4) requirements.
The design of fuel storage racks are in compliance with regulatory requirements in the areas of thermal hydraulics, nuclear criticality, and materials analysis.
The structural analysis of the fuel storage racks evaluates the design for the event of a fuel assembly drop, seismic and structural integrity. The fuel assembly drop is evaluated to the design criteria in NuScale Design Specific Review Standard (DSRS) 3.8.4 (Reference 11). The seismic design criteria for the fuel racks are specified in DSRS 3.7.1 (Reference 8) and ASCE/SEI 43 (Reference 9). The structural integrity of the fuel storage racks are evaluated to criteria specified in ASME Code Section III, Division I, Subsection NF (Reference 12). The structural analysis within this report has been intentionally omitted. The COL applicant shall provide the NRC a structural evaluation of the spent fuel storage racks. This evaluation is dependent on a vendor specific design and the as-built configuration of spent fuel storage racks.
The thermal-hydraulic analysis evaluates the ability to cool the spent fuel according to the criteria recommended in DSRS 9.1.2 (Reference 2), III Review Procedure 4.I. The thermal-hydraulic analysis of the flow through the fuel racks is adequate for decay heat removal from the spent fuel assemblies with two trains of spent fuel pool (SFP) cooling in operation and upon loss of one cooling train. Furthermore, the thermal-hydraulic analysis shows adequate flow circulation of the coolant during anticipated operating conditions (one failed SFP cooling train), including full core-offloads during refueling, to prevent nucleate boiling for all fuel assemblies.
The nuclear criticality analysis of the fuel storage racks evaluates the designs to ensure criticality control is provided to meet the criticality-related portions of 10 CFR Part 50 (Reference 3),
Section 50.68. The fuel storage racks are in compliance with regulatory requirements in the area of nuclear criticality.
The structural components of the fuel storage racks are at risk to degrade through corrosion mechanisms. To mitigate these risks, stainless steel is chosen for the structural components of the racks. These materials have a proven positive performance in light water reactors and are expected to perform well in the NuScale SFP environment for the design lifetime of 60 years.
Based on the evaluations summarized within this document, the fuel storage rack design is in compliance with the regulatory requirements. The fuel storage racks allow placement of fuel assemblies with a maximum initial enrichment of 5 percent U-235 and a maximum burn-up of
((2(a),(c) without restrictions for zoning or loading patterns after 72 hours post shutdown of the reactor. © Copyright 2018 by NuScale Power, LLC 2
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 1.0 Introduction The fuel storage racks are nonsafety-related and not risk-significant. However, the fuel storage racks are designed to Seismic Category I and 10 CFR 50, Appendix B requirements. Safety-related qualified software is used in the design and meets the requirements of ASME NQA-1. This report is prepared in accordance with the guidance provided by the NuScale Design Specific Review Standard (DSRS). The fuel storage racks are designed as modular rack systems for storing, loading, and unloading fuel assemblies (FAs) in the spent fuel pool (SFP). The design incorporates a flux trap and fixed neutron absorbers. The racks are able to store both irradiated and un-irradiated fuel in the water-filled SFP. The design is a general placement design developed for the long term storage of irradiated fuel and new (unirradiated) fuel. The analyses performed do not impose any restriction for FA placement in the SFP. The fuel storage rack is an 11x11 configuration, allowing up to 121 FAs to be safely stored per rack. The fuel storage racks can safely store at most 1404 FAs vertically in the SFP, factoring in the maximum reach of the fuel handling machine path. Because of the reach of the fuel handling machine (red line shown in Figure 12), not all fuel cells can safely be reached. Therefore, a maximum of 1404 FAs can be placed in the accessible storage locations but more or fewer stored FAs are also considered in some of the analyses as described below. 1.1 Abbreviations Table 1-1 Abbreviations Term Definition ALE Arbitrary-Lagrangian-Eulerian ASME American Society of Mechanical Engineers BOL beginning of life CEUS Central and Eastern United States CFD computational fluid dynamics CRA control rod assembly DSRS Design Specific Review Standard EOL end of life FA Fuel Assembly FEA finite element analysis HF high frequency, refers to high-frequency target-response spectra, i.e., HF1 and HF2 HX heat exchanger © Copyright 2018 by NuScale Power, LLC 3
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Term Definition ID inner diameter IR interaction ratio ISRS in-structure response spectra LF low frequency, refers to low frequency target response spectra, i.e., LF1, LF2, LF3 and LF4 LWR light water reactor MMC metal matrix composite MWD/MTU megawatt day per metric ton, uranium NRC Nuclear Regulatory Commission OD outer diameter PNNL Pacific Northwest National Laboratories PSD power spectrum density RS response spectra SCC stress-corrosion cracking SFP spent fuel pool SRSS square root of the sum of the squares SSE safe-shutdown earthquake WUS Western United States ZPA zero period acceleration 95/95 95 percent probability with 95 percent confidence Table 1-2 Definitions Term Definition Of, being, or related to a designated or theoretical value that Nominal may vary from the actual value 1.2 General Description of the Fuel Storage Racks The fuel storage racks are classified as nonsafety-related and not risk-significant. However, the fuel storage racks are designed to Seismic Category I and 10 CFR 50, Appendix B requirements. The design, materials, and procurement of the racks are in accordance with American Society of Mechanical Engineers (ASME) NQA-1a-2009, 2008 Revision with 2009 Addenda (Reference 4). All structural materials meet ASME Boiler and Pressure Vessel Code (Code) Section II (Reference 5). Design of the fuel storage racks is in accordance with applicable sections the DSRS and applicable requirements in the ASME Code Section III, Division I, Subsection NF. Additionally, guidance specified in ANS © Copyright 2018 by NuScale Power, LLC 4
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 57.2 (Reference 33) and ANS 57.3 (Reference 34) that pertains to the design of the fuel storage racks is used. The rack design presented in this report can be used to store either unirradiated (new) or irradiated (spent) FAs. Unirradiated assemblies and irradiated assemblies with a maximum enrichment of 5 wt% U-235 and (( }}2(a),(c) average assembly burnup are stored in the SFP. The fuel design is derived from AREVAs 17x17 pressurized water reactor design for Westinghouse-type reactors. Many of the design features are common to the Advanced W17 HTP' fuel design currently in operation, but the length is scaled down to meet the NuScale design. The overall length of the FA from the outside shoulders of the end fittings is 94 inches and has an active fuel length of 78.74 inches. The size of the FA is 8.426 x 8.426 inches and weighs 830 pounds. The fuel storage design consists of fourteen 11x11 free-standing modules within the SFP with a 121 usable storage spaces divided into arrays of (( }}2(a),(c) inner dimension square tubes arranged on an 11.22 pitch providing a flux trap to store and maintain criticality control of the new and spent FAs (see Figure 11, Figure 13, and Figure 14). Expanded views of the rack assemblies are provided in Figure 11a through Figure 11e to supplement Figure 11. ((
}}2(a),(c)
The principal structural load-carrying material used in the design is ((
}}2(a),(c) stainless steel fabricated from a mixture of sheet steel, bar, and rolled steel stock of standard and readily available sizes. The ((
}}2(a),(c) stainless steel meets the corrosion requirements (( }}2(a),(c) while providing the minimum mechanical properties (( }}2(a),(c) . The individual tubes surrounding the FAs are fabricated from the same (( }}2(a),(c) stainless steel and include a beveled lead-in to aid in the placement of FAs. The lead-in is designed to allow identification markers to be embossed for tracking FAs. The lead-ins are welded along the edges, but remain open at the intersection of four adjacent cells. The bottom of the rack is constructed of a thick baseplate ( (( }}2(a),(c) ) providing structural rigidity and support against FA drop scenarios.
The rack modules are used in wet conditions to allow the receipt and storage of new fuel and freshly discharged FAs in the SFP. © Copyright 2018 by NuScale Power, LLC 5
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 The wet storage racks stand freely on the SFP floor liner, separated by a small gap between each other considering the main body (i.e., fuel tubes, and upper and lower grids) and a larger gap between the racks and the pool walls (see Figure 12). ((
}}2(a),(c) The fuel tubes are supported by a base plate with an area larger than the area of an 11x11 matrix of tubes, which is used to set the racks in appropriate proximity of each other (see Figure 13 and Figure 14). The racks are butted up against each other, which leaves no spacing between the racks, considering the baseplate and mid-level cross stiffener beam, but are not physically joined (see Figure 12 and Figure 13). The layout and gaps are made to prevent an accidental placement of an FA between racks (see Figure 12a and Figure 12b). The travel limitations of the fuel handling machine prevent misplacement of an FA between a wall and a rack. Administrative controls are imposed to prevent placing FAs in the open space at the fuel elevator. Each of the rack modules is supported by (( }}2(a),(c) independently adjustable feet to ensure level installation.
Standard material sizes are chosen to facilitate fabrication to the extent possible. The procurement and quality control procedures used in the fabrication are in accordance with the requirements of the ASME Code Section III, Subsection NF (Reference 12). The installation of the racks in the pool may be performed in any sequence which best suits the erection contractor because there is no fuel present that would require special control of rack placement for criticality concerns. All feet of the rack are vertically adjustable and the racks are leveled prior to installation of the other racks. The feet are secured after the rack is level. The racks are butted up against each other and leave basically no spacing between the racks at the baseplate level. All racks are installed prior to filling water in the SFP. © Copyright 2018 by NuScale Power, LLC 6
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 11 Fuel storage rack general configuration
© Copyright 2018 by NuScale Power, LLC 7
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 11a Bottom Grid © Copyright 2018 by NuScale Power, LLC 8
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 11b Bottom Grid, Cross Bars, Outer Grid Lower Bands, and Corner Posts © Copyright 2018 by NuScale Power, LLC 9
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 11c Mid Band Assembly © Copyright 2018 by NuScale Power, LLC 10
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 11d Top Inner Grid © Copyright 2018 by NuScale Power, LLC 11
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 11e Top Outer Grid © Copyright 2018 by NuScale Power, LLC 12
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Figure 12 Spent fuel pool general arrangement of racks © Copyright 2018 by NuScale Power, LLC 13
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Figure 12a Cross-sectional view looking north © Copyright 2018 by NuScale Power, LLC 14
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Figure 12b Cross-sectional view looking west © Copyright 2018 by NuScale Power, LLC 15
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 13 Fuel storage rack width
© Copyright 2018 by NuScale Power, LLC 16
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 14 Fuel storage rack cross section with fuel assembly 1.3 Purpose This document defines the documents, analytical inputs, interpretations, and methodologies used to design and analyze the fuel storage racks. This document also contains summaries of the results necessary to comply with the regulatory requirements for the design of fuel storage racks. However, the structural analysis within this report has been intentionally omitted. The COL applicant shall provide the NRC a structural evaluation of the spent fuel storage racks. This evaluation is dependent on a vendor specific design and the as-built configuration of spent fuel storage racks.
© Copyright 2018 by NuScale Power, LLC 17
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 2.0 Background The rack designs in this report are designed to store either unirradiated (new) or irradiated (spent) FAs. Unirradiated assemblies and irradiated assemblies with a maximum enrichment of 5 wt% U-235 and (( }}2(a),(c) assembly average burnup are stored in the fuel storage racks in the SFP. 2.1 Analysis Application The design and analysis presented in this report uses fundamental approaches and industry-accepted software. Conservatism is used in each analysis. Thermal-hydraulic calculations use the industry-accepted ANSYS finite element analysis (FEA) code. Criticality evaluations use KENO-V.a as controlled by the CSAS5 module of SCALE. SCALE is a widely used software code, which provides a tool set for criticality safety, reactor physics, radiation shielding, radioactive source term characterization, and sensitivity and uncertainty analysis. 2.2 Codes, Standards, and Regulatory Requirements
- 10 CFR 50, Appendix A, General Design Criteria for Nuclear Power Plants, latest edition.
- 10 CFR 50, Appendix B, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants.
- 10 CFR 20.1101 (b), Radiation Protection Programs.
- 10 CFR 50.68, Criticality Accident Requirements.
- Regulatory Guide 1.13, Spent Fuel Storage Facility Design Basis, Revision 2.
- Regulatory Guide 1.29, Seismic Design Classification for Nuclear Power Plants, Revision 5.
- Regulatory Guide 1.61, Damping Values for Seismic Design of Nuclear Power Plants, Revision 1.
- Regulatory Guide 1.92, Combining Modal Responses and Spatial Components in Seismic Response Analysis, Revision 2.
- Regulatory Guide 8.8, Information Relevant to Maintaining Occupational Radiation Exposure as Low as is Reasonably Achievable," Revision 3, June 1976.
- Spent Fuel Project Office Interim Staff Guidance - 9, Storage of Components Associated with FAs, Revision 1.
- NUREG/CR-0612, Control of Heavy Loads at Nuclear Power Plants, July 1980.
© Copyright 2018 by NuScale Power, LLC 18
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1
- NUREG/CR-6801, Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses, March 2003.
- NUREG/CR-6665 ORNL/TM-1999/303, Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel, Prepared by C. V. Parks, M. D. DeHart, and J. C. Wagner.
- ASME NQA-1a-2009, Quality Assurance Program for Nuclear Facilities," 2008 Revision with 2009 Addenda.
- ASME Boiler and Pressure Vessel Code, Section II, Materials, 2007 Edition.
- ASME Boiler and Pressure Vessel Code, Section III, Division I, Sub-section NF, Class 3, 2007 Edition.
- ASME Boiler and Pressure Vessel Code, Section V, Nondestructive Examination, 2007 Edition and Addenda.
- ASME Boiler and Pressure Vessel Code, Section IX, Welding and Brazing Qualifications, 2007 Edition and Addenda.
- ANSI/ANS 8.17, 2004, Criticality Safety Criteria for Handling, Storage, and Transportation of LWR Fuel Outside Reactors.
- ANSI/ANS 57.1, 1992 (reaffirmed 2005), Design Requirements for Light Water Reactor Fuel Handling Systems.
- ANSI/ANS 57.2, 1983, Design Requirements for Light Water Reactor Spent Fuel Storage facilities at Nuclear Power Plants, Withdrawn in 1993 (reaffirmation in progress).
- ANSI/ANS 57.3, 1983, Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants, Withdrawn in 1993 (reaffirmation in progress).
- ANSI/ANS-5.1-1979, American National Standard for Decay Heat Power in Light Water Reactors.
- ANSI N 14.6-1993, Special Lifting Devices for Shipping Containers Weighing 10,000 Pounds (4,500 kg) or More.
© Copyright 2018 by NuScale Power, LLC 19
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 3.0 Analysis 3.1 Structural Analysis (Intentionally Omitted) 3.1.1 Fuel Storage Detailed and Simplified Structural Model Development (Intentionally Omitted) 3.1.2 Time-History Development for Fuel Storage Rack Floor Response Spectra (Intentionally Omitted) 3.1.3 Load-Drop Analysis for Fuel Racks (Intentionally Omitted) 3.1.4 Whole-Pool Analysis and Fuel Storage Rack Interaction (Intentionally Omitted) 3.1.5 Fuel Storage Rack Detailed Stress Analysis and Design (Intentionally Omitted) 3.1.6 Sensitivity Analysis - Whole-Pool Analysis with Partially Loaded Racks and Varying Friction Coefficients (Intentionally Omitted) 3.2 Thermal-Hydraulic Analysis The purpose of the thermal-hydraulic analysis is to evaluate SFP temperature according to the criteria recommended in DSRS 9.1.2 (Reference 2, Review Procedure 2.I) for normal and anticipated operating condition (loss of one cooling train). Specifically, the thermal-hydraulic design requirement used to assess the fuel storage racks within this report must show that the peak fuel cladding temperature is maintained below the local saturation temperature for the water at the top of the FA under all design conditions (with cooling available). The thermal-hydraulic analysis is to demonstrate with computational fluid dynamics (CFD) analysis that there is adequate cooling flow at a safe level through the FAs and SFP racks and that the decay heat removal is sufficient to prevent nucleate boiling on the fuel rod cladding during normal operating and anticipated operating conditions (loss of one cooling train). Decay heat removal during accident conditions is provided by the ultimate heat sink design. © Copyright 2018 by NuScale Power, LLC 20
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 3.2.1 Maximum SFP Bulk Temperature Calculation 3.2.1.1 Methodology Heat exchanger (HX) temperature data at several values of SFP water inlet temperature and heat load are used to calculate a performance coefficient for the HX. The performance coefficient is then used to predict the temperature difference required to remove the specified decay heat load. The decay heat load for the SFP is based on the maximum allowed design burnup for the fuel using ANSI/ANS-5.1-1979 (Reference 18). 3.2.1.2 Assumptions The combination of forced flow from one cooling train and natural circulation induced by differential fluid density is sufficient to cause full mixing of the cooled fluid entering the SFP into the bulk SFP water (i.e., there is no localized hot spots in the SFP, and therefore, the HX performance can be determined based on averaged pool conditions). This is justified based on the results of the CFD analysis. 3.2.1.3 Configuration The calculation is performed based on one train of SFP cooling in operation. 3.2.1.4 Initial Conditions, Boundary Conditions, and Limitations ((
}}2(a),(c)
The maximum HX inlet cooling water temperature (site cooling water temperature) is 91 degrees F. 3.2.1.5 Software Use and Qualification No specialized calculation software is used; only hand calculations are performed. 3.2.1.6 Analysis, Evaluation, and Data The HX data sheets reflect HX performance at various values of cooling water temperature and heat load duty. From these data sheets, the values of heat removal rate, tube-side inlet and outlet temperatures, site cooling water temperature (shell-side inlet temperature), and the tube-side and overall temperature differences, are summarized below in Table 3-1. A conceptual diagram of the SFP heat exchanger is shown in Figure 31. © Copyright 2018 by NuScale Power, LLC 21
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Table 3-1 Heat exchanger data sheet summary ((
}}2(a),(c)
Figure 31 Spent fuel pool heat exchanger concept When only one train of SFP cooling is in service due to a postulated single failure of the other train, the entire heat load of (( }}2(a),(c) must be removed by the single operating train of SFP cooling. The overall HX T to transfer this heat is ((
}}2(a),(c) . Because the maximum shell-side inlet temperature (site cooling water temperature) is 91.0 degrees F, the HX tube-side inlet temperature would have to be
(( }}2(a),(c) . By engineering judgement this value is increased to ((
}}2(a),(c)
For a nominal (no extra margin) HX tube side inlet temperature of ((
}}2(a),(c), with one train of SFP cooling operating the tube-side outlet temperature would be (( }}2(a),(c) because the T across the HX is
(( }}2(a),(c) . 3.2.1.7 Implementation and Use Maximum SFP bulk temperatures have been determined using a conservative heat load with one train of SFP cooling in service.
© Copyright 2018 by NuScale Power, LLC 22
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 For a nominal (no extra margin) HX tube side inlet temperature of ((
}}2(a),(c) , with one train of SFP cooling operating the tube side outlet temperature would be (( }}2(a),(c).
3.2.2 Spent Fuel Pool Computational Fluid Dynamics Analysis 3.2.2.1 Methodology ((
}}2(a),(c)
Once the maximum water temperature has been determined, a hand calculation is used to determine the film heat transfer coefficient at the outer surface of the fuel rod cladding and the cladding surface temperature. An axial peaking factor is applied to the heat flux used in the hand calculation to maximize the cladding temperature. The saturation temperature at the top of the FAs is calculated using the pressure due to the assumed ambient overpressure and the elevation head resulting from the minimum water level for the SFP. The calculated maximum fuel rod cladding temperature must be less than this saturation temperature. 3.2.2.2 Assumptions
- 1. No credit is taken for heat loss through the pool walls.
- 2. No credit is taken for evaporative cooling from the pool surface.
- 3. No credit is taken for convective cooling to ambient air.
- 4. The pressure above the fuel pool is slightly negative, for conservatism a pressure of -
1/2 in wg (0.018 psig) is assumed.
- 5. The SFP volume is considered self-contained and there is no communication with any other water source by way of the weir wall opening, which includes the refueling pool and reactor pool. The reactor pool cooling system has sufficient capacity for cooling the heat loads in the reactor and refueling pools so that no heat load from these two pools needs to be added to the SFP.
- 6. ((
}}2(a),(c)
© Copyright 2018 by NuScale Power, LLC 23
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c) 3.2.2.3 Configuration The spent fuel pool is analyzed using a CFD model of the pool water, the cooling system connections, and the structures and fuel assemblies in the pool.
The CFD model is created using ANSYS Design Modeler. The CFD model is meshed using ANSYS Mesher. The calculation of fluid flow patterns and fluid temperatures is performed using ANSYS FLUENT. ((
}}2(a),(c)
The FLUENT code is configured to use the superficial velocity in its calculations, as opposed to the physical (actual) velocity. The superficial velocity is the velocity that would exist if there is no blockage (porosity = 1.0), but the same flow rate. Therefore, the superficial velocity is less than the actual velocity that would exist. The pressure loss factors are scaled to accommodate this reduced velocity, so the only effect on the model is the static pressure changes due to acceleration of the fluid. Because the velocities in the SFP are relatively small, this effect is insignificant. ((
}}2(a),(c)
© Copyright 2018 by NuScale Power, LLC 24
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Once a suitable pressure loss coefficient correlation has been determined, the conversion process consists of plotting the pressure loss as a function of velocity (i.e., Reynolds number) and curve fitting the results with a second order polynomial equation to separate loss factors into their viscous and inertial resistance components. The friction loss coefficient associated with flow along surfaces (i.e., the fuel rods and interior sides of the storage rack cell) is determined from the standard Darcy equation. ((
}}2(a),(c)
© Copyright 2018 by NuScale Power, LLC 25
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
The walls and floor of the SFP are left with the default configuration of adiabatic (zero heat flux) and zero slip. © Copyright 2018 by NuScale Power, LLC 26
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 3.2.2.4 Initial Conditions, Boundary Conditions, and Limitations The cooling water flow rate with a single train in operation is 450,000 lbm/hr. This flow rate is evenly distributed among the four cooling water inlet faces, which are configured as mass flow rate boundary conditions. The actual flow rate of water through the fuel assemblies that are stored in the spent fuel pool is only mildly dependent on the cooling water flow rate. This is because the cooling water inlet and outlet pipes are located high on the pool walls, so there is no forced cooling of the fuel assemblies. The flow through the fuel assemblies is buoyancy driven natural circulation flow that enters the storage racks through (( }}2(a),(c) diameter holes near the bottom, and exits the top of the racks. After exiting the racks the heated fluid must pass through a large quantity of less heated fluid, giving up a large amount of energy in the process. The amount of mixing of the fluid exiting the racks with the inlet cooling water flow will determine the temperature of the water entering the bottom of the racks. The cooling flow rate is based on a single train in operation. The expected cooling water inlet temperature (nominal HX - with no extra margin - outlet temperature) is 111.2 degrees F when one cooling water train is in operation. This temperature is applied to all four inlet boundary conditions. The four outlet faces are configured as outflows, allowing the FLUENT code to determine the flow rates and temperatures such that mass and energy are conserved. For each outflow boundary condition, the fraction of the total outflow assigned to that boundary condition is specified to be 0.25. The uniform distribution discussion above for the inlets applies to the outlets as well. ((
}}2(a),(c) 3.2.2.5 Software Use and Qualification ANSYS/FLUENT Version 15.0.7, an ASME/NQA-1 2008/2009a-approved computer code for safety-related application, is used to prepare, mesh, and solve the CFD model.
3.2.2.6 Analysis, Evaluation, and Data 3.2.2.6.1 Computational Fluid Dynamics Case Runs ((
}}2(a),(c)
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Five different locations for the storage rack with the core offload FAs are analyzed. The storage racks are identified with a two-digit number that encodes the rack matrix position within the SFP. The rack nearest the northwest corner is identified as Rack #11, the rack nearest the northeast corner is identified as Rack #15, and the rack in the southeast corner is identified as Rack #35. The identifiers are shown in Figure 32. Figure 32 Top view with rack numbers Using this identification technique, the five cases are defined in Table 3-2. This table includes the highest liquid temperature on the upper surface of the active fuel region in the rack with the core offload FAs, as shown on the temperature contour plots (see note below regarding Case 4). © Copyright 2018 by NuScale Power, LLC 28
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Table 3-2 Analysis cases and peak liquid temperatures ((
}}2(a),(c)
Each case is provided with seven results figures: The first two figures of each case (Figure 33, Figure 34, Figure 310, Figure 311, Figure 317, Figure 318, Figure 324, Figure 325, Figure 331, and Figure 332) are temperature contours at the top and bottom of the active fuel regions. Because there is a uniform heat addition in the porous regions, the direction of flow can be determined by observing the temperature change from bottom to top. The third figure of each case (Figure 35, Figure 312, Figure 319, Figure 326, and Figure 333) shows the temperature contours on two vertical planes that are aligned with the X (east-west) and Z (north-south) axes. One of the planes is positioned to pass through the middle of the storage rack with the higher heat loading. The other plane is positioned to pass through the centerline of one of the cooling water inlets. This contour clearly shows how the cooling water turns downward and sinks through the warmer pool water. The fourth and fifth figures of each case (Figure 36, Figure 37, Figure 313, Figure 314, Figure 320, Figure 321, Figure 327, Figure 328, Figure 334, and Figure 335) show streamlines that are color coded by temperature. The fourth figure shows streamlines that originate at the cooling water inlets. These streamlines also show how the cooling water turns downward and sinks. The fifth figure shows streamlines that pass through the upper face of the active fuel region of the storage rack with the higher heat load. The sixth and seventh figures of each case (Figure 38, Figure 39, Figure 315, Figure 316, Figure 322, Figure 323, Figure 329, Figure 330, Figure 336, and Figure 337) show plots of values that are recorded during the process of iterating to a solution. The sixth plot gives the solver residuals and the seventh plot gives the maximum temperature at the top of the active fuel region for the rack with the high heat load.
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 33 Temperature contour at the upper surface of the active fuel region, case 1
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 34 Temperature contour at the lower surface of the active fuel region, case 1
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 35 Temperature contour on vertical planes, case 1
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 36 Streamlines from the cooling water inlets, case 1
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 37 Streamlines from the upper surface of the active fuel region, case 1
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 38 Plot of the solution residuals, case 1
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 39 Plot of the solution peak spent fuel pool temperature, case 1
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 310 Temperature contour at the upper surface of the active fuel region, case 2
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 311 Temperature contour at the lower surface of the active fuel region, case 2
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 312 Temperature contour on vertical planes, case 2
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 313 Streamlines from the cooling water inlets, case 2
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 314 Streamlines from the upper surface of the active fuel region, case 2
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 315 Plot of the solution residuals, case 2
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 316 Plot of the solution peak spent fuel pool temperature, case 2
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 317 Temperature contour at the upper surface of the active fuel region, case 3
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 318 Temperature contour at the lower surface of the active fuel region, case 3
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 319 Temperature contour on vertical planes, case 3
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 320 Streamlines from the cooling water inlets, case 3
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 321 Streamlines from the upper surface of the active fuel region, case 3
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 322 Plot of the solution residuals, case 3
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 323 Plot of the solution peak spent fuel pool temperature, case 3
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 324 Temperature contour at the upper surface of the active fuel region, case 4
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 325 Temperature contour at the lower surface of the active fuel region, case 4
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 326 Temperature contour on vertical planes, case 4
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 327 Streamlines from the cooling water inlets, case 4
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 328 Streamlines from the upper surface of the active fuel region, case 4
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 329 Plot of the solution residuals, case 4
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 330 Plot of the solution peak spent fuel pool temperature, case 4
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 331 Temperature contour at the upper surface of the active fuel region, case 5
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 332 Temperature contour at the lower surface of the active fuel region, case 5
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 333 Temperature contour on vertical planes, case 5
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 334 Streamlines from the cooling water inlets, case 5
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 335 Streamlines from the upper surface of the active fuel region, case 5
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 336 Plot of the solution residuals, case 5
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 337 Plot of the solution peak spent fuel pool temperature, case 5 The case with the lowest peak temperature, Case 3, would also have the highest flow rate through the storage rack with the higher heat load. A contour of the velocity in the active fuel region of storage Rack #11 for Case 3 is shown in Figure 338. This shows the maximum superficial velocity is ((
}}2(a),(c) Both of these velocities are within the range of velocities covered in the pressure-loss coefficient evaluations. © Copyright 2018 by NuScale Power, LLC 64
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 338 Velocity in active fuel region of rack 11 for case 3 3.2.2.6.2 Verification of Mesh Independence To verify that the results are not dependent on the formation of the CFD mesh, an additional mesh with about twice the number of elements is prepared. This mesh uses the same geometry as the base mesh, but the mesh nominal sizing parameter is set to 3 inches. The resulting mesh has about (( }}2(a),(c) elements and a minimum element quality of about (( }}2(a),(c) . The meshing settings and some of the mesh statistics are shown in Figure 339.
© Copyright 2018 by NuScale Power, LLC 65
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 339 Mesher settings and statistics for 3 mesh This mesh is used for a repeat of Case 1. The resulting peak temperature at the upper surface of the active fuel region of storage Rack #15 is (( }}2(a),(c) . This is within (( }}2(a),(c) of the result for Case 1, and is lower, indicating the results have an insignificant dependence on the size of the mesh. © Copyright 2018 by NuScale Power, LLC 66
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 3.2.2.6.3 Peak Fuel Cladding Temperature The peak fuel cladding temperature is computed based upon the decay heat dissipated from the FA, the peak local water temperature determined in the CFD case runs, and the computed film heat transfer coefficient between the fuel rod and the water. ((
}}2(a),(c)
Computing the Nusselt number and solving for the heat transfer coefficient yields a value of (( }}2(a),(c) . The combined thermal resistance of the oxide layer and the fluid convective film coefficient is (( }}2(a),(c) . To account for the radial peaking, the temperature rise across the FA must be increased by the peaking factor. Applying the convective heat transfer coefficient with the calculated peak local SFP water temperature of (( }}2(a),(c) , the maximum fuel rod cladding temperature for an FA dissipating the core offload decay heat is (( }}2(a),(c) . The saturation temperature for water at the top of the FA with a minimum pool water level of 55 feet and an ambient pressure of -0.018 psig is the acceptance criteria for this temperature. The density of water at 160 degrees F and the ambient pressure is 61.00 lbm/ft3. The saturation pressure at the top of the FA is 34.39 psia. The saturation temperature at this pressure is 258 degrees F. Thus, the storage rack design is shown to provide sufficient buoyancy-driven water flow to meet the criteria of maintaining the peak FA rod temperature below the local saturation temperature. 3.2.2.7 Implementation and Use These results show that placing the core offload FAs in the rack nearest the southeast corner of the SFP produces the highest local liquid temperatures. The temperature contours given in Figure 33, Figure 310, Figure 317, Figure 324, and Figure 331 indicate that, for cases other than Case 4, the rack nearest the southeast corner has downward flow through it. This is caused by the proximity of the rack to the cooling water inlets, because the cooler, denser water naturally sinks once it enters the SFP. This natural tendency for downflow in the area near the inlets is sufficient to overpower the buoyancy-driven flow within the storage rack near the inlets, when the lower © Copyright 2018 by NuScale Power, LLC 67
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 heat load is applied. Therefore, with the higher heat load applied near the inlets, the buoyancy-driven flow is partially canceled by the downflow from above, and a lower flow rate through the rack is produced. This lower flow rate results in a higher local liquid temperature. ((
}}2(a),(c) 3.3 Criticality Analysis The fuel storage racks provide criticality control to meet the following acceptance criteria of 10 CFR 50.68 applicable to storage of FAs in the SFP:
- k-effective must not exceed 1.0 with racks flooded with unborated water and fuel of the maximum permissible reactivity at a 95 percent probability with 95 percent confidence (95/95).
- k-effective must not exceed 0.95 with racks flooded with a minimum boron concentration, CB,min, (800 ppm) and fuel of the maximum permissible reactivity at a 95/95.
The racks are designed for storing 17X17 fuel assemblies with a 0.374 in. fuel rod diameter and 0.496 in. fuel rod pitch. A single rack design with a flux trap is employed in the SFP to store either new or burned FAs. The rack module uses poison plates made of boron carbide-aluminum metal matrix composite (MMC) as the neutron poison material. This material is ideal for long-term use in the chemical, radiation, and thermal environments of wet storage racks. ((
}}2(a),(c)
© Copyright 2018 by NuScale Power, LLC 68
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 3.3.1 Methodology KENO-V.a controlled by the CSAS5 module of SCALE Version 6.1.3 is used to calculate the effective multiplication factor (k-effective) of the fuel stored in the fuel storage rack. The CSAS5 module allows simplified data input to the functional modules BONAMI-S, CENTRM/PMC, and KENO-V.a. These modules process the required cross sections and calculate the k-effective of the system. The 238-group ENDF/B-VII.0 Library is selected for this calculation. BONAMI-S performs the resonance self-shielding calculations for nuclides that have Bondarenko data associated with their cross sections. CENTRM/PMC computes continuous-energy neutron spectra in zero- or one-dimensional systems, by solving the Boltzmann Transport Equation using a combination of point-wise and multi-group nuclear data. Finally, KENO-V.a calculates the k-effective of a three-dimensional system. A sufficiently large number of neutron histories are run that the standard deviation is acceptably low for all calculations and passes the chi-square test for normality at the 5-percent level. The equation below is used to evaluate k-effective at the 95 percent probability and 95 percent confident level as required by the regulations.
/
/ = + + + + + + Equation 1 where, keff = the KENO-V.a calculated result ksys = summation of k values associated with the variation of system and base case modeling parameters (e.g., moderator temperature) biasm = bias associated with the calculation methodology compared to benchmarks.
The bias is always calculated as a negative number and is used here as the absolute value of the number. C = 2, consistent with NUREG/CR-6698, Equation 2 k = standard deviation of the KENO-V.a calculated result sys = standard deviation of ksys ktol = sum of statistically independent manufacturing tolerances tol = standard deviation of ktol © Copyright 2018 by NuScale Power, LLC 69
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Criticality benchmark calculations are performed to establish the value of biasm. These calculations benchmark the ability of the criticality code to predict the reactivity of a system based on comparison to critical experiments. The criticality benchmark calculations and their applicability to the fuel rack analyses are documented in Section 3.3.7. The fuel storage rack analyses assume the racks contain 17X17 fuel assemblies with a 0.374 in. fuel rod diameter and 0.496 in. fuel rod pitch. The FA parameters used for this analysis are shown in Table 3-3. The fuel storage rack analyses assume fresh fuel at the maximum allowable enrichment of 5 wt% U-235. The analyses do not take credit for burnup, zoning, or a loading pattern. The design parameters for the racks that comprise the fuel storage are shown in Table 3-4. The material compositions are shown in Table 3-5. The analysis for the fuel storage uses a water density at 67.73 degrees F of 0.9982 g/cm3. A bias term, defined in Section 3.3.4.1, is included in the calculation of k95/95 to adjust the system operating temperature to 40 degrees F and a density of 1.0 g/cm3. © Copyright 2018 by NuScale Power, LLC 70
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Table 3-3 Fuel assembly specification for criticality analysis Description Nominal Value Tolerance Overall size 8.426 in. x 8.426 in. Overall weight 830 lb. Overall length 94 in. Rod array 17x17 Number of fuel rods 264 Number of guide tubes 24 Number of instrument tubes 1 Instrument tube inside diameter 11.43 mm (( }}2(a),(c) (ID) Instrument tube outside diameter 12.24 mm (( }}2(a),(c) (OD) Instrument tube material Zircaloy-4 Fuel pellet theoretical density 96 percent (( }}2(a),(c) Fuel pellet diameter 0.3195 in. (( }}2(a),(c) Fuel rod clad ID 0.326 in. (( }}2(a),(c) Furl rod clad OD 0.374 in. (( }}2(a),(c) Fuel active length 78.74 in. (( }}2(a),(c) Fuel rod pitch 0.496 in. ((
}}2(a),(c)
Fuel rod clad material M5 Guide tube ID 11.43 mm (( }}2(a),(c) Guide tube OD 12.24 mm (( }}2(a),(c) © Copyright 2018 by NuScale Power, LLC 71
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Table 3-4 Fuel rack design parameters for criticality analysis Description Nominal Value Tolerance(Note 2) ((
}}2(a),(c)
Storage tube pitch 11.22 in. ((
}}2(a),(c)
Fuel storage rack array size 11x11 Fuel storage temperature range 40 to 212°F ((
}}2(a),(c) © Copyright 2018 by NuScale Power, LLC 72
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Table 3-5 Material properties ((
}}2(a),(c)
B4C 2.52 g/cm3 Aluminum 2.702 g/cm3
- 1. The ASTM standard in 2006 and prior years gave this range as 18.0 to 20.0 wt%. This range is still widely used in industry references. ((
}}2(a),(c) 3.3.2 Assumptions Unless otherwise stated, the following assumptions are common to the analysis of the fuel storage racks.
- 1. No burnable poisons are modeled and the U-235 enrichment remains at the maximum value, rather than the cutback value that is realistically present in a fuel rod containing certain types of burnable poison.
- 2. No axial blankets are modeled. The U-235 enrichment remains at the maximum value throughout the active fuel length.
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1
- 3. The fuel pellet density is not reduced by either the chamfer and or the dish. This conservatively increases the amount of fissionable material in the model.
- 4. The U-235 enrichment is assumed to be 5 wt percent, the maximum allowed by 10 CFR 50.68 (b)(7). Enrichment is not included in the tolerance analysis.
- 5. Sensitivity studies were performed to determine the effect of the structural elements in the model. From these studies, an additional bias and uncertainty in the bias of 0.00222 +/- 0.00042 is included in the determination of the k95/95 values for the racks.
- 6. Only the active fuel length is modeled, with a water reflector at the axial boundary.
This is conservative in that FA structure and rack structure, including some poison plate length, is replaced by moderator. Because the amount of poison plate that is omitted is much larger than the tolerance on the length of the poison plate, the poison plate length is not included in the tolerance analysis.
- 7. The radial boundary conditions for the fuel storage single rack cases include a water gap that is governed by the dimensions of the base plate that supports the rack, with a periodic boundary condition that simulates an infinite array of fuel storage racks.
- 8. The radial boundary conditions for the fuel storage whole-pool cases include a water gap that is governed by the nominal distance to the pool wall, with a concrete reflector consistent with the pool wall.
- 9. The whole-pool cases for the fuel storage assume all rack locations are used, even though some locations are physically inaccessible.
3.3.3 Configuration 3.3.3.1 Fuel Storage Rack Model The rack for the fuel storage is modeled in KENO-V.a with the nominal FA dimensions shown in Table 3-3 and the nominal rack dimensions shown in Table 3-4. The KENO-V.a model of a single rack location for the fuel storage is shown in Figure 340. The KENO-V.a model of a single fuel storage rack is shown in Figure 341. The fuel storage rack is modeled with a radial reflector region that is half the minimum rack-to-rack spacing, with periodic radial boundary conditions that simulate an infinite array of racks. The regions directly above and below the fuel region are modeled as 13 inches of water followed by a water boundary condition. The fuel storage whole pool is modeled as a collection of 14 individual racks, spaced with the minimum distance between the outer cells of adjacent racks, and with the nominal spacing to the concrete pool walls. The whole pool model is used for the dropped assembly analysis and damaged FA analysis. The KENO-V.a model of the whole pool, with a dropped assembly, is shown in Figure 342. © Copyright 2018 by NuScale Power, LLC 74
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Figure 340 KENO-V.a model of single rack location in fuel storage © Copyright 2018 by NuScale Power, LLC 75
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Figure 341 KENO-V.a model of a single fuel storage rack © Copyright 2018 by NuScale Power, LLC 76
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Figure 342 KENO-V.a model of the spent fuel pool 3.3.4 Initial Conditions, Boundary Conditions, and Limitations 3.3.4.1 Fuel Storage Model The normal operating and accident condition temperature for the fuel storage racks is 40 degrees F to 212 degrees F (277.59 degrees K to 373.2 degrees K). The lowest available temperature in the 238 group ENDF/B-VII cross section library is 67.73 degrees F (293 degrees K). Therefore, to evaluate the sensitivity of the system to changes in temperature, cases are evaluated across a temperature range of 67.73 degrees F to 212 degrees F for 0 ppm boron and 800 ppm boron. The moderator density changes with the temperature for these cases, so the variation is due to both the temperature effect on the cross-section as well as the change in moderator density. The uncertainty terms for these cases are calculated using the root sum of squares of the base case and an uncertainty for the extrapolated value which was calculated by the statistical propagation of the errors of the curve fit coefficients. The results of the temperature sensitivity calculations and the extrapolated data points are provided in Table 3-6. Figure 343 shows how the system responds, with a slightly negative gradient (-1x10-4 slope) to increases in temperature, with or without boron in the moderator. © Copyright 2018 by NuScale Power, LLC 77
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Table 3-6 System sensitivity to nominal temperature ((
}}2(a),(c) © Copyright 2018 by NuScale Power, LLC 78
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 343 System sensitivity to nominal temperature The operational tolerances for the 0 ppm and 800 ppm boron cases are calculated as the differential between the base (67.73 degrees F) case and the 40 degrees F extrapolated case. The uncertainty terms for these cases are calculated using the root sum of squares of the base case and an assumed uncertainty for the 40 degrees F extrapolated case of 0.00142. These operational tolerance values are applied in the k95/95 calculation to account for the positive reactivity bias associated with the lower temperature of 40 degrees F. Table 3-7 Operational tolerance impact on neutron multiplication Conditions ksys ksys Operational tolerance, 0 ppm 0.00412 0.00142 Operational tolerance, 800 ppm 0.00376 0.00154 3.3.5 Software Use and Qualification KENO-V.a controlled by the CSAS5 module of SCALE Version 6.1.3 is qualified for use. SCALE Version 6.1.3 is the software used to perform safety-related calculations in accordance with AREVA procedures.
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 3.3.6 Analysis, Evaluation, and Data 3.3.6.1 Fuel Storage Tolerance Analysis The fuel storage tolerance analysis for both the fuel and the rack shown in Table 3-8 is performed with unborated moderator and the tolerance analysis shown in Table 3-9 is performed with the moderator at 800 ppm dissolved boron. The poison plate thickness does not affect the areal density of B-10. The fuel storage tolerance analysis shows there is no significant trend of k-effective versus poison plate motion in the direction parallel to the face of the plate. The flux trap cases for the fuel storage tolerance analysis are shown in Table 3-8 and Table 3-9. The delta of (( }}2(a),(c) for the flux trap is determined by the tolerance on the thickness of the (( }}2(a),(c) square tube that defines the minimum water gap thickness. This value is still larger than the spacing restrictions imposed by the grid assemblies at the top and bottom of the rack. The FA position tolerance analysis for the fuel storage consists of a single case where all assemblies move diagonally to the common corner of each group of four locations. The last rows of both Table 3-8 and Table 3-9 show the value of ktol and tol, for the respective boron concentration in the moderator. ktol is calculated by summing all the positive values of k, while tol is calculated as the statistical propagation (SRSS) of the respective values of k, where the corresponding value of k is positive. Separate sensitivity studies were performed to assess the effect of the rack internal structural elements that include the elements located above and below the active fuel and the support tubes located in the flux trap region. The studies showed that modeling the structural material above and below the fuel as an unborated reflector would provide an adequate representation of those components. The results also showed that there was a nearly statistically equal value when compared to the water only case. In the cases that examined the support bars or tubes in the flux trap region, the results showed that modeling the tubes would increase the reactivity by approximately 0.223% with an uncertainty of 0.0004 in an unborated or moderately borated (800 ppm boron) pool. From these studies, the additional bias and uncertainty in the bias is 0.00222 +/- 0.00042 for 800 ppm boron, and 0.00223 +/- 0.00038 for no boron. © Copyright 2018 by NuScale Power, LLC 80
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Table 3-8 Spent fuel pool tolerance analysis with unborated moderator ((
}}2(a),(c)
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Table 3-9 SFP tolerance analysis with moderator at 800 ppm dissolved boron ((
}}2(a),(c)
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c) 3.3.6.2 Damaged Fuel Assembly The fuel storage is required to hold up to five damaged FAs. For this analysis a damaged FA is assumed to have a cladding failure in 100 percent of the fuel rods where the gap between the fuel pellet and the clad is flooded with water.
The damaged fuel analysis is run with the whole pool model as shown in Figure 342, less the extra assembly shown for the drop analysis. Four scenarios with damaged FAs are simulated, as described below. The analyses are run with both an unborated moderator and with a dissolved boron concentration of 800 ppm. © Copyright 2018 by NuScale Power, LLC 84
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1
- Damaged fuel is concentrated at the center of a fuel storage rack.
- Damaged fuel is positioned in corner locations within the fuel storage rack where the poison plates are slightly narrower.
- Damaged fuel is concentrated around a single corner of a single rack.
- All FAs are assumed to be damaged.
The results are shown in Table 3-10. The values of k95/95 shown in this table are calculated using Equation 1 in Section 3.3.1, using the appropriate values shown in Table 3-11. Table 3-10 Fuel storage damaged fuel analysis Scenario keff k95/95 Limit No damaged fuel, full pool, 800 ppm boron 0.86710 0.00009 0.92191 0.95 5 damaged FAs, in center of rack, 800 ppm boron 0.86740 0.00008 0.92220 0.95 4 damaged FAs, in corner locations, 800 ppm boron 0.86717 0.00008 0.92197 0.95 5 damaged FAs, in one corner, 800 ppm boron 0.86710 0.00009 0.92191 0.95 All damaged FAs, 800 ppm boron 0.87376 0.00009 0.92857 0.95 No damaged fuel, full pool, 0 ppm boron 0.91627 0.00009 0.96616 1.0 5 damaged FAs, in center of rack, 0 ppm boron 0.91654 0.00010 0.96642 1.0 4 damaged FAs, in corner locations, 0 ppm boron 0.91606 0.00009 0.96594 1.0 5 damaged FAs, in one corner, 0 ppm boron 0.91623 0.00009 0.96612 1.0 All damaged FAs, 0 ppm boron 0.92441 0.00010 0.97429 1.0 © Copyright 2018 by NuScale Power, LLC 85
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Table 3-11 Factors for k95/95 calculation for fuel storage Symbol Description Value Source biasm Bias associated with the calculation methodology -0.02346 Section 3.3.7 compared to benchmark C C, confidence factor 2 Section 3.3.7 ktol Manufacturing tolerance, no boron 0.01974 Table 3-8 tol Sigma, manufacturing tolerance, no boron 0.00109 Table 3-8 ktol Manufacturing tolerance, 800 ppm 0.02496 Table 3-9 tol Sigma, manufacturing tolerance, 800 ppm 0.00159 Table 3-9 ksys Operational bias for temperature, no boron 0.00412 Section 3.3.4.1 sys Sigma, operational bias for temperature, no boron 0.00142 Section 3.3.4.1 ksys Operational bias for temperature, 800 ppm 0.00376 Section 3.3.4.1 sys Sigma, operational bias for temperature, 800 ppm 0.00154 Section 3.3.4.1 ksys Operational bias for structural material, no boron 0.00223 Section 3.3.6.1 sys Sigma, operational bias for structural material, no 0.00038 Section boron 3.3.6.1 ksys Operational bias for structural material, 800 ppm 0.00222 Section 3.3.6.1 sys Sigma, operational bias for structural material, 800 0.00042 Section ppm 3.3.6.1 3.3.6.3 Assembly Dropped on Top of Rack A dropped assembly lying on top of the rack in a horizontal position the fuel storage is at least (( }}2(a),(c) from the top of the active fuel region of the assemblies stored in the rack. © Copyright 2018 by NuScale Power, LLC 86
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 For the fuel storage in water this distance represents approximately ((
}}2(a),(c), which is more than sufficient to neutronically uncouple the dropped assembly from the assemblies stored in the rack. Thus, an assembly dropped on top of a fuel storage rack in the SFP is not a criticality concern.
3.3.6.4 Assembly Dropped Outside of the Fuel Storage Rack The SFP consists of 14 individual racks, arranged in three rows of five racks, with one corner rack omitted to allow for the FA elevator. For the nominal configuration, there is not enough space to drop a FA between two racks. The only location where an assembly can be dropped adjacent to a rack is in the empty corner with the elevator. In this scenario, the dropped FA is placed as close to the surrounding filled storage racks as possible. Figure 342 provides a general representation of the relative position for the dropped fuel assembly. Several cases were analyzed that considered a dropped FA, including located in the corner near the three racks, as well as directly adjacent (face-to-face) and half-way between two adjacent fuel assemblies in a rack. The cases also spanned the full width of the rack. The location of the dropped assembly was shown to be statistically insignificant. The results are shown in Table 3-12. The values of k95/95 shown in this table are calculated using Equation 1 in Section 3.3.1, using the appropriate values shown in Table 3-11. Thus, an assembly dropped outside of a fuel storage rack in the SFP is not a criticality concern. Table 3-12 Spent fuel pool dropped fuel assembly analysis Scenario keff k95/95 Full pool, no dropped fuel 0.86710 0.92191 0.00009 1 dropped fuel assembly in fuel 0.86738 0.00015 0.92219 elevator area 3.3.6.5 Misloaded Fuel Assembly The fuel storage analysis conservatively assumes the racks are completely loaded with FAs at the maximum reactivity. There are no restrictions on loading patterns, therefore, there is no possibility of misloading an assembly in the fuel storage racks. © Copyright 2018 by NuScale Power, LLC 87
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 3.3.6.6 Fuel Storage Racks Seismic Event The seismic event is analyzed with the single fuel storage rack model and periodic radial boundary conditions. The spacing between storage tubes is reduced in a series of cases to determine the maximum deformation that maintains k95/95 less than the limit of 0.95. The results are shown in Table 3-13 and Figure 344. The values of k95/95 shown in this table are calculated using Equation 1 in Section 3.3.1, using the appropriate values shown in Table 3-11. A COL applicant shall provide the NRC a structural evaluation of the spent fuel storage racks. The structural evaluation shall include the mechanical analysis of the fuel storage racks for a seismic event to demonstrate that the racks do not undergo permanent or transient deformation in excess of the maximum allowed deformation, and thus demonstrate that there is no possibility of a seismic event causing a criticality in the SFP. Table 3-13 k-effective and k95/95 for a seismic event in the fuel storage ((
}}2(a),(c)
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c)
Figure 344 k95/95 for a seismic event in the fuel storage
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 3.3.7 Benchmark Analysis The software codes used for criticality analysis are validated against benchmark experiments. The validation determines the calculation bias and the uncertainty of the bias that is associated with the modeling methodology, the code, and the cross-section library that are used to perform the criticality analysis. Guidance for the validation is provided in NUREG/CR-6698 (Reference 20). Criticality codes are verified by comparing benchmark calculations to actual critical benchmark experiments. The difference between the calculated reactivity and the experimental reactivity is referred to as calculational bias. This bias may be a function of system parameters, such as fuel lattice separation, fuel enrichment, neutron absorber properties, reflector properties, or fuel/moderator volume ratio, or there may be no specific correlation with system parameters. The purpose of this validation is to statistically determine the magnitude of the calculational bias and the uncertainty of the bias, and whether any such dependencies exist so they may be properly accounted for in the criticality analysis. The parameter range for the selection of critical experiments for the benchmark cases is shown in Table 3-14. © Copyright 2018 by NuScale Power, LLC 90
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Table 3-14 Parameter range for critical experiment selection Parameter Critical Experiment Requirement Fissile material Uranium Isotopic composition 3.5 to 6.5 wt% Physical form Pellets of UO2 clad with Zircaloy 4 Temperature -22 to 157 degrees F Soluble Boron concentration 0 to 2000 ppm Moderator to fuel area ratio 1.99 Moderation material H2O Moderator density 1.0 g/cm3 +/-10 wt% Neutron absorbing material Should be within 20% of the atom density ((
}}2(a),(c)
(( }}2(a),(c) ((
}}2(a),(c)
Neutron spectrum The neutron energy is to cover the same energy range and similar spectrum. Spectrum can be quantified by Energy of Average Lethargy of Fission (EALF) 3.3.7.1 Methodology 3.3.7.1.1 Bias and Bias Uncertainty When comparing the experimentally measured k-effective (kexp) to the calculated k-effective (kcalc), the values are normalized as shown in NUREG/CR-6698, Equation 9 (Reference 20): calc norm = exp The measured k-effective is adjusted for some experiments where the published benchmark model has a known omission, which causes a small but non-negligible modeling bias. © Copyright 2018 by NuScale Power, LLC 91
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 In addition, the errors are combined statistically as shown in NUREG/CR-6698, Equation 3:
= calc + exp The weighted mean value of k-effective is calculated with the following set of equations as shown in NUREG/CR-6698, Equations 4 - 7:
Variance about the mean: 1 1 norm norm 1
=
1 1 Average total uncertainty:
=
1 Weighted mean knorm: 1 norm norm = 1 Square root of the pooled variance:
= +
Finally, the bias is determined as: Bias = norm 1 if knorm is less than one; otherwise Bias = 0. And the uncertainty on the bias is Where C is the 95/95 tolerance factor dependent on the sample size. © Copyright 2018 by NuScale Power, LLC 92
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 3.3.7.1.2 Normality The use of the pooled variance and single-sided tolerance limit or the use of the tolerance band (NUREG/CR-6698 Section 2.4.4) is predicated on the requirement for the data to be representative of a normal distribution. This document uses two tests to determine if a given distribution may be represented by a normal distribution. In the details given in NUREG/CR-6698, a Shapiro-Wilk test is used, but as described it is limited to sample sizes between 10 and 50. This document uses an expanded Shapiro-Wilk test (Royston, 1992, Reference 38) which is suitable for sample sizes between 12 and 2000. This document also uses the DAgostino test from ANSI N15.15 (Reference 22). As described in ANSI N15.15 the test is for sample sizes between 50 and 2000. For the expanded Shapiro-Wilk test: Sort the data in ascending order, x1 xn Define the values m1, , mn as mi = NORMINV((i-.375)/(n+.25)) Determine m=the sum of (mi)2 Set u=1/SQRT(n) and define the coefficients a1, , an: an = -2.706056u5 + 4.434685u4 - 2.071190u3 - 0.147981u2 + 0.221157u + mn*m-0.5 an-1 = -3.582633u5 + 5.682633u4 - 1.752461u3 - 0.293762u2 + 0.042981u + mn-1*m-0.5 ai = mi/SQRT() for 2 < i < n-1 a2=-an-1 and a1=-an where: 2 2
=
1 2 2 ( )2 The W statistic is defined as = )2 ( The value ln(1-W) is approximately normally distributed with the following mean and standard deviation
= 0.0038915(ln()) 0.083751(ln()) 0.31082(ln()) 1.5861
. ( ( )) . ( ( )) .
=
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ( ) And the statistic = can be tested against the standard normal distribution. If the p value is less than the desired level of significance (0.05 is used here), the hypothesis that the data is normal is rejected. For the DAgostino test:
=
Where from ANSI N15.15-1974, Section 7.2.2: (n + 1) T= i x 2 And from ANSI N15.15-1974, Section 4.2.2: ( )
=
Critical values for the 95th percentile would be P(0.025) and P(0.975) for the two-tailed test. 3.3.7.1.3 Tolerance Band If statistically significant trends are discovered, then a one-sided lower tolerance band may be used. The equation for the one-sided lower tolerance band is taken from NUREG/CR-6698, Equation 23: ( , ) 1 ( ) ( 2)
= () 2 + +
( ) , Where
- Kfit(x) is the functional fit of keff versus parameter x, derived in the trend analysis.
Note that for Kfit(x) > 1, use Kfit(x) = 1
- SPfit is the pooled variance for the experiments used in the fit.
- P is the desired confidence level (0.95)
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ( , )
- is the F distribution percentile with degree of fit of 2 and n-2 degrees of freedom (assuming Kfit is a linear fit). Calculated with the Excel function F.INV.RT(1-p,2,n-2). Note that NUREG/CR-6698 states that the function FINV should be used; F.INV.RT is the preferred equivalent function for the obsolete FINV function.
- n is the number of experiments used for the trend analysis of parameter x
- x is the independent variable of the regression
- xi is value of the independent variable for the ith experiment
- is the weighted mean of the independent variables
- Z2P-1 is the symmetric percentile of the normal distribution that contains the P fraction. Calculated with the Excel function NORM.S.INV(P). Note that NUREG/CR-6698 states that the function NORMSINV should be used; NORM.S.INV is the preferred equivalent function for the obsolete NORMSINV function.
- is (1-P)/2
* , is the upper Chi-square percentile. Calculated with the Excel function CHISQ.INV.RT(1-,n-2). Note that NUREG/CR-6698 states that the function CHIINV should be used; CHISQ.INV.RT is the preferred equivalent function for the obsolete CHIINV function.
3.3.7.1.4 Non-parametric Analysis If the benchmark data is found to not be normally distributed, then a non-parametric analysis can be used to determine the limiting neutron multiplication factor, KL, taken from NUREG/CR-6698, Equation 33: KL = Smallest keff value - Uncertainty for Smallest keff - Non-parametric Margin (NPM) Or, if the smallest keff value is greater than unity: KL = 1 - SP - NPM Where: Uncertainty for the Smallest keff, taken as total for the experiment. NPM = Non-Parametric Margin, from Table 2.2 of NUREG/CR-6698 Sp = Square root of the pooled variance © Copyright 2018 by NuScale Power, LLC 95
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 The following equation determines the confidence that a fraction of the population is above the value of rank m as taken from NUREG/CR-6698, Equation 31:
=1 (1 ) ( )
! ( )!
Where: q = the desired population fraction (normally 0.95) n = the number of data in one data sample m = the rank order indexing from the smallest sample to the largest (m=1 for the smallest sample m = 2 for the second smallest sample, etc.). For a desired population fraction of 0.95 and a rank order of m=1 (the smallest sample in the data set), the equation reduces to NUREG/CR-6698, Equation 32:
= 1 0.95 The Non-Parametric Margin is taken from NUREG/CR-6698, Table 2.2, and assumes that for > 0.90, the Non-Parametric Margin is zero. This occurs for n greater than or equal to 45 with m=1 (using the smallest sample).
From NUREG/CR-6698, Equation 31, for n greater than or equal to 93 and m=2, is greater than 0.95. This means that for data sets with greater than 93 points, the second smallest value of keff may be used for KL while maintaining a confidence level of 95/95. 3.3.7.2 Selection of Experiments A large set of experiments is selected for either the determination of the bias and bias uncertainty, or for the determination of trends, or for both uses. The complete list of experiments used for the determination of the bias and bias uncertainty is provided in Table 3-22. 3.3.7.2.1 Selection of Experiments for Enrichment Trend The experiments shown in Table 3-15 are chosen to sample the calculation of k-effective with U-235 enrichment that is representative of the fuel storage racks. The range of U-235 enrichments was from 3.0 wt% to 6.5 wt%. While this is slightly outside the desired area of applicability (3.5-6.5 wt%), it does not adversely bias the results with its inclusion, rather it lends statistical significance to the sample. © Copyright 2018 by NuScale Power, LLC 96
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Table 3-15 Benchmark experiments selected for enrichment trend Experiment Name Reference Cases Selected # of Cases LEU-COMP-THERM-002 Ref. 23 1-5 5 LEU-COMP-THERM-009 Ref. 23 1 - 4, 9 5 LEU-COMP-THERM-010 Ref. 23 5, 16 - 19 5 LEU-COMP-THERM-013 Ref. 23 1, 2, 3 3 LEU-COMP-THERM-014 Ref. 23 1, 2, 5, 6, 7 5 LEU-COMP-THERM-039 Ref. 23 1 - 17 17 LEU-COMP-THERM-048 Ref. 23 1-5 5 LEU-COMP-THERM-050 Ref. 23 1-7 7 LEU-COMP-THERM-070 Ref. 23 1 - 12 12 LEU-COMP-THERM-075 Ref. 23 4, 5, 6 3 PAT80 Ref. 21 SS1, SS2 2 Total Cases 69 3.3.7.2.2 Selection of Experiments for Fuel Rod Pitch Trend The experiments shown in Table 3-16 are chosen to sample the calculation of k-effective with fuel rod pitch that is representative of the fuel storage racks. The range of fuel rod pitch was from 1.26 cm - 2.54 cm. This adequately covers the expected fuel pitch of (( }}2(a),(c) for the fuel. In order to obtain a wide range of pitch values, it is necessary to expand the area of applicability for enrichment to enrichments as low as 2.35 wt%. The result of expanding the area of applicability serves to improve the statistical significance of the trend. © Copyright 2018 by NuScale Power, LLC 97
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Table 3-16 Benchmark experiments selected for fuel rod pitch trend Experiment Name Reference Cases Selected # of Cases LEU-COMP-THERM-001 Ref. 23 1-8 8 LEU-COMP-THERM-002 Ref. 23 1-5 5 LEU-COMP-THERM-008 Ref. 23 1 - 4, 16, 17 6 LEU-COMP-THERM-009 Ref. 23 1 - 4, 9 5 LEU-COMP-THERM-010 Ref. 23 5, 16 - 19 5 LEU-COMP-THERM-012 Ref. 23 2-5 4 LEU-COMP-THERM-013 Ref. 23 1, 2, 3 3 LEU-COMP-THERM-014 Ref. 23 1, 2, 5, 6, 7 5 LEU-COMP-THERM-016 Ref. 23 12 - 14 3 LEU-COMP-THERM-039 Ref. 23 1 - 17 17 LEU-COMP-THERM-042 Ref. 23 3 1 LEU-COMP-THERM-048 Ref. 23 1-5 5 LEU-COMP-THERM-050 Ref. 23 1-7 7 LEU-COMP-THERM-051 Ref. 23 10 - 19 10 PAT80 Ref. 21 SS1, SS2 2 Total Cases 86 3.3.7.2.3 Selection of Experiments for Fuel Assembly Separation Trend The experiments shown in Table 3-17 are chosen to sample the calculation of k-effective with FA separation that is representative of the fuel storage racks. The FA separation ranged from zero cm to 15.393 cm. This adequately covers the expected FA separation for the fuel storage racks, which is approximately 7 cm. In order to obtain a wide range of assembly separation values, it is necessary to expand the area of applicability for enrichment to enrichments as low as 2.35 wt%. The result of expanding the area of applicability serves to improve the statistical significance of the trend. © Copyright 2018 by NuScale Power, LLC 98
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Table 3-17 Benchmark experiments selected for fuel assembly separation trend Experiment Name Reference Cases Selected # of Cases LEU-COMP-THERM-001 Ref. 23 1-8 8 LEU-COMP-THERM-009 Ref. 23 1 - 4, 9 5 LEU-COMP-THERM-010 Ref. 23 5, 16 - 19 5 LEU-COMP-THERM-012 Ref. 23 2-5 4 LEU-COMP-THERM-013 Ref. 23 1, 2, 3 3 LEU-COMP-THERM-016 Ref. 23 12, 13, 14 3 LEU-COMP-THERM-042 Ref. 23 3 1 LEU-COMP-THERM-051 Ref. 23 10 - 19 10 PAT80 Ref. 21 SS1, SS2 2 Total Cases 41 3.3.7.2.4 Selection of Experiments for Soluble Boron Trend The experiments shown in Table 3-18 are chosen to sample the calculation of k-effective with soluble boron concentration that is representative of the fuel storage racks. The soluble boron concentration ranged from 0 ppm to 5030 ppm. This adequately covers the expected soluble boron concentration and the modeled scenario of low-boron concentration for the fuel storage racks, of 2000 ppm and 800 ppm, respectively. In order to obtain a wider range of boron concentration values, it is necessary to expand the area of applicability for enrichment to enrichments as low as 2.35 wt%. Without this extension, there are only 12 experiments in the range of the nominal boron concentration and none in the range of the low-boron concentration used. Table 3-18 Benchmark experiments selected for soluble boron trend Experiment Name Reference Cases Selected # of Cases LEU-COMP-THERM-008 Ref. 23 1 - 4, 16, 17 6 LEU-COMP-THERM-014 Ref. 23 2, 5, 6, 7 4 LEU-COMP-THERM-050 Ref. 23 1-7 7 LEU-COMP-THERM-051 Ref. 23 10 - 19 10 Total Cases 27 © Copyright 2018 by NuScale Power, LLC 99
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 3.3.7.2.5 Selection of Experiments for Boron Separator Plate Areal Density Trend The experiments shown in Table 3-19 are chosen to sample the calculation of k-effective with boron separator plates that are representative of the NuScale fuel storage racks. The boron atom density ranged from 1.93E-5 to about 4.03E-3 at/barn. This adequately covers the expected boron density for the fuel storage racks, of approximately 1.35E-3 at/barn. In order to obtain a wider range of boron densities, it is necessary to expand the area of applicability for enrichment to enrichments as low as 2.35 wt%. Additionally, it is not possible to analyze the impact of boron areal density within the suggested +/-20 percent, so the range of boron concentration in the separator plates is increased until a minimal sample size is obtained. Without this extension, there are only five experiments in the range of the nominal boron density, which is far too small a sample to determine if a trending bias is present. Also, there are Boroflex' experiments that are not chosen, as this material is specifically excluded from the construction of the fuel storage racks. Cases that use stainless steel separator plates are included, even when impregnated with boron, because it is expected that the construction will make use of large amounts of stainless steel, if not for the poison plate substrate, it will still separate the FAs. Table 3-19 Benchmark experiments selected for separator plate boron areal density trend Experiment Name Reference Cases Selected # of Cases LEU-COMP-THERM-009 Ref. 23 9 1 LEU-COMP-THERM-012 Ref. 23 2-5 4 LEU-COMP-THERM-013 Ref. 23 2, 3 2 LEU-COMP-THERM-016 Ref. 23 12, 13, 14 3 LEU-COMP-THERM-042 Ref. 23 3 1 LEU-COMP-THERM-051 Ref. 23 10 - 19 10 PAT80 Ref. 21 SS1, SS2 2 Total Cases 23 3.3.7.2.6 Selection of Experiments for Moderator to Fuel Area Ratio Trend The experiments shown in Table 3-20 are chosen to sample the calculation of k-effective with moderator-to-fuel ratios that are representative of the fuel storage racks. The range of ratios is from 0.996 to 3.882. This adequately covers the expected ratio for the fuel storage racks of approximately 1.99. In most cases, the moderator-to-fuel ratio is not given in any reference, so a simple calculation is performed to compare the area of the fuel to the area of the moderator. © Copyright 2018 by NuScale Power, LLC 100
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Table 3-20 Benchmark experiments selected for moderator to fuel area ratio trend Experiment Name Reference Cases Selected # of Cases LEU-COMP-THERM-002 Ref. 23 1-5 5 LEU-COMP-THERM-009 Ref. 23 1 - 4, 9 5 LEU-COMP-THERM-010 Ref. 23 5, 16 - 19 5 LEU-COMP-THERM-013 Ref. 23 1-3 3 LEU-COMP-THERM-014 Ref. 23 1, 2, 5 - 7 5 LEU-COMP-THERM-039 Ref. 23 1 - 17 17 LEU-COMP-THERM-048 Ref. 23 1-5 5 LEU-COMP-THERM-050 Ref. 23 1-7 7 Total Cases 52 3.3.7.2.7 Selection of Experiments for Neutron Spectrum The experiments shown in Table 3-21 are chosen to sample the calculation of keff with the neutron spectrum, as represented by the Energy of Average Lethargy of Fission (EALF), that was representative of the NuScale fuel storage racks. The EALF ranged from 0.11 eV to 0.748 eV. This adequately covers the expected neutron spectrum of 0.187 to 0.512 eV. Table 3-21 Benchmark experiments selected for neutron spectrum Experiment Name Reference Cases Selected # of Cases LEU-COMP-THERM-002 Ref. 23 1-5 5 LEU-COMP-THERM-009 Ref. 23 1-4, 9 5 LEU-COMP-THERM-010 Ref. 23 5,16-19 5 LEU-COMP-THERM-013 Ref. 23 1-3 3 LEU-COMP-THERM-014 Ref. 23 1,2,5,6,7 5 LEU-COMP-THERM-039 Ref. 23 1-17 17 LEU-COMP-THERM-048 Ref. 23 1-5 5 LEU-COMP-THERM-050 Ref. 23 1-7 7 PAT80 Ref. 21 SS1, SS2 2 Total Cases 54 © Copyright 2018 by NuScale Power, LLC 101
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 3.3.7.3 Results of Benchmark Calculations A summary of the pertinent parameters for each experiment is shown in Table 3-22 with the results of each KENO-V.a case. Table 3-22 Critical experiment parameters and KENO-V.a results Benchmark Model Pin Soluble 10B area (Note 1) Encrichment Pitch Separation Boron density EALF
# Name (%) (cm) (cm) (ppm) (at/barn) H/U (eV) keff exp keff calc 1 LEU-COMP-THERM-001-001 2.35 2.032 0 2.918 9.67E-02 0.9998 0.0030 0.9980 0.00100 2 LEU-COMP-THERM-001-002 2.35 2.032 11.92 2.918 9.58E-02 0.9998 0.0030 0.9983 0.00110 3 LEU-COMP-THERM-001-003 2.35 2.032 8.41 2.918 9.50E-02 0.9998 0.0030 0.9983 0.00110 4 LEU-COMP-THERM-001-004 2.35 2.032 10.05 2.918 9.54E-02 0.9998 0.0030 0.9984 0.00095 5 LEU-COMP-THERM-001-005 2.35 2.032 6.39 2.918 9.43E-02 0.9998 0.0030 0.9959 0.00089 6 LEU-COMP-THERM-001-006 2.35 2.032 8.01 2.918 9.52E-02 0.9998 0.0030 0.9967 0.00100 7 LEU-COMP-THERM-001-007 2.35 2.032 4.46 2.918 9.36E-02 0.9998 0.0031 0.9972 0.00130 8 LEU-COMP-THERM-001-008 2.35 2.032 7.57 2.918 9.45E-02 0.9998 0.0031 0.9970 0.00096 9 LEU-COMP-THERM-002-001 4.306 2.54 3.882 1.13E-01 0.9997 0.0020 0.9976 0.00056 10 LEU-COMP-THERM-002-002 4.306 2.54 3.882 1.13E-01 0.9997 0.0020 0.9987 0.00029 11 LEU-COMP-THERM-002-003 4.306 2.54 3.882 1.13E-01 0.9997 0.0020 0.9983 0.00026 12 LEU-COMP-THERM-002-004 4.306 2.54 3.882 1.12E-01 0.9997 0.0018 0.9982 0.00028 13 LEU-COMP-THERM-002-005 4.306 2.54 3.882 1.10E-01 0.9997 0.0019 0.9968 0.00026 14 LEU-COMP-THERM-008-001 2.459 1.636 1511 1.841 2.80E-01 1.0007 0.0012 0.9975 0.00029 15 LEU-COMP-THERM-008-002 2.459 1.636 1334 1.944 2.47E-01 1.0007 0.0012 0.9985 0.00031 16 LEU-COMP-THERM-008-003 2.459 1.636 1337 1.944 2.47E-01 1.0007 0.0012 0.9983 0.00028 17 LEU-COMP-THERM-008-004 2.459 1.636 1183 1.934 2.46E-01 1.0007 0.0012 0.9978 0.00030 18 LEU-COMP-THERM-008-016 2.459 1.636 1158 2.026 2.28E-01 1.0007 0.0012 0.9973 0.00031 19 LEU-COMP-THERM-008-017 2.459 1.636 921 2.205 1.99E-01 1.0007 0.0012 0.9972 0.00037 20 LEU-COMP-THERM-009-001 4.306 2.54 8.58 3.882 1.13E-01 1.0000 0.0021 0.9988 0.00027 21 LEU-COMP-THERM-009-002 4.306 2.54 9.65 3.882 1.12E-01 1.0000 0.0021 0.9980 0.00029 22 LEU-COMP-THERM-009-003 4.306 2.54 9.22 3.882 1.13E-01 1.0000 0.0021 0.9981 0.00028 23 LEU-COMP-THERM-009-004 4.306 2.54 9.76 3.882 1.12E-01 1.0000 0.0021 0.9984 0.00028 24 LEU-COMP-THERM-009-009 4.306 2.54 6.72 4.03E-03 3.882 1.14E-01 1.0000 0.0021 0.9990 0.00025 25 LEU-COMP-THERM-010-005 4.306 2.54 14.255 3.882 3.56E-01 1.0000 0.0021 0.9997 0.00025 26 LEU-COMP-THERM-010-016 4.306 1.892 15.393 1.597 2.85E-01 1.0000 0.0028 1.0031 0.00031
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# Name (%) (cm) (cm) (ppm) (at/barn) H/U (eV) keff exp keff calc 27 LEU-COMP-THERM-010-017 4.306 1.892 15.363 1.597 2.79E-01 1.0000 0.0028 1.0016 0.00029 28 LEU-COMP-THERM-010-018 4.306 1.892 14.973 1.597 2.75E-01 1.0000 0.0028 1.0012 0.00030 29 LEU-COMP-THERM-010-019 4.306 1.892 13.343 1.597 2.68E-01 1.0000 0.0028 1.0013 0.00029 30 LEU-COMP-THERM-012-002 2.35 1.684 3.86 2.68E-04 1.75E-01 1.0000 0.0034 0.9862 0.00088 31 LEU-COMP-THERM-012-003 2.35 1.684 3.46 4.21E-04 1.75E-01 1.0000 0.0034 0.9857 0.00093 32 LEU-COMP-THERM-012-004 2.35 1.684 1.68 1.82E-03 1.83E-01 1.0000 0.0034 0.9890 0.00093 33 LEU-COMP-THERM-012-005 2.35 1.684 1.93 1.58E-03 1.82E-01 1.0000 0.0034 0.9848 0.00091 34 LEU-COMP-THERM-013-001 4.306 1.892 13.273 1.597 2.86E-01 1.0000 0.0018 1.0003 0.00027 35 LEU-COMP-THERM-013-002 4.306 1.892 9.353 2.78E-04 1.597 2.94E-01 1.0000 0.0018 1.0000 0.00030 36 LEU-COMP-THERM-013-003 4.306 1.892 7.823 1.82E-03 1.597 2.98E-01 1.0000 0.0018 0.9995 0.00027 37 LEU-COMP-THERM-014-001 4.306 1.89 0 1.591 2.78E-01 1.0000 0.0019 0.9982 0.00027 38 LEU-COMP-THERM-014-002 4.306 1.89 490 1.591 3.33E-01 1.0000 0.0077 0.9861 0.00022 39 LEU-COMP-THERM-014-005 4.306 1.89 2550 1.591 5.84E-01 1.0000 0.0069 1.0007 0.00020 40 LEU-COMP-THERM-014-006 4.306 1.715 0 1.089 4.97E-01 1.0000 0.0033 1.0053 0.00023 41 LEU-COMP-THERM-014-007 4.306 1.715 1030 1.089 7.48E-01 1.0000 0.0051 1.0007 0.00022 42 LEU-COMP-THERM-016-012 2.35 2.032 6.33 4.03E-03 9.76E-02 1.0000 0.0031 0.9977 0.00110 43 LEU-COMP-THERM-016-013 2.35 2.032 9.03 4.03E-03 9.66E-02 1.0000 0.0031 0.9964 0.00140 44 LEU-COMP-THERM-016-014 2.35 2.032 5.05 4.03E-03 9.74E-02 1.0000 0.0031 0.9975 0.00084 45 LEU-COMP-THERM-039-001 4.7376 1.26 2.000 2.22E-01 1.0000 0.0014 0.9958 0.00030 46 LEU-COMP-THERM-039-002 4.7376 1.26 2.083 2.12E-01 1.0000 0.0014 0.9974 0.00034 47 LEU-COMP-THERM-039-003 4.7376 1.26 2.317 1.93E-01 1.0000 0.0014 0.9969 0.00074 48 LEU-COMP-THERM-039-004 4.7376 1.26 2.228 1.84E-01 1.0000 0.0014 0.9953 0.00085 49 LEU-COMP-THERM-039-005 4.7376 1.26 3.048 1.39E-01 1.0000 0.0009 0.9989 0.00083 50 LEU-COMP-THERM-039-006 4.7376 1.26 2.903 1.46E-01 1.0000 0.0009 0.9979 0.00095 51 LEU-COMP-THERM-039-007a 4.7376 1.26 2.000 2.13E-01 1.0000 0.0012 0.9963 0.00011 52 LEU-COMP-THERM-039-008 4.7376 1.26 2.083 2.03E-01 1.0000 0.0012 0.9965 0.00090 53 LEU-COMP-THERM-039-009 4.7376 1.26 2.083 1.98E-01 1.0000 0.0012 0.9982 0.00085 54 LEU-COMP-THERM-039-010 4.7376 1.26 2.317 1.74E-01 1.0000 0.0012 0.9970 0.00085 55 LEU-COMP-THERM-039-011 4.7376 1.26 2.000 2.22E-01 1.0000 0.0013 0.9953 0.00079 56 LEU-COMP-THERM-039-012a 4.7376 1.26 2.000 2.17E-01 1.0000 0.0013 0.9953 0.00016 57 LEU-COMP-THERM-039-013 4.7376 1.26 2.000 2.15E-01 1.0000 0.0013 0.9960 0.00083 58 LEU-COMP-THERM-039-014 4.7376 1.26 2.000 2.13E-01 1.0000 0.0013 0.9968 0.00093 59 LEU-COMP-THERM-039-015 4.7376 1.26 2.000 2.12E-01 1.0000 0.0013 0.9961 0.00085
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# Name (%) (cm) (cm) (ppm) (at/barn) H/U (eV) keff exp keff calc 60 LEU-COMP-THERM-039-016 4.7376 1.26 2.000 2.11E-01 1.0000 0.0013 0.9966 0.00085 61 LEU-COMP-THERM-039-017 4.7376 1.26 2.000 2.10E-01 1.0000 0.0013 0.9956 0.00089 62 LEU-COMP-THERM-042-003 2.35 1.684 2.69 1.82E-03 1.82E-01 1.0000 0.0016 0.9978 0.00054 63 LEU-COMP-THERM-048-001 3 1.32 0.996 6.76E-01 1.0000 0.0025 0.9985 0.00034 64 LEU-COMP-THERM-048-002 3 1.32 0.996 6.52E-01 1.0000 0.0025 0.9986 0.00028 65 LEU-COMP-THERM-048-003a 3 1.32 0.996 6.81E-01 1.0000 0.0025 0.9985 0.00010 66 LEU-COMP-THERM-048-004 3 1.32 0.996 6.83E-01 1.0000 0.0025 0.9982 0.00025 67 LEU-COMP-THERM-048-005 3 1.32 0.996 6.73E-01 1.0000 0.0025 0.9985 0.00026 68 LEU-COMP-THERM-050-001 4.738 1.3 0 2.203 2.00E-01 1.0000 0.0010 0.9981 0.00012 69 LEU-COMP-THERM-050-002 4.738 1.3 0 2.220 1.91E-01 1.0000 0.0010 0.9982 0.00012 70 LEU-COMP-THERM-050-003 4.738 1.3 822 2.189 2.07E-01 1.0000 0.0010 0.9986 0.00012 71 LEU-COMP-THERM-050-004 4.738 1.3 822 2.203 1.98E-01 1.0000 0.0010 0.9980 0.00011 72 LEU-COMP-THERM-050-005 4.738 1.3 5030 2.165 2.22E-01 1.0000 0.0010 0.9993 0.00012 73 LEU-COMP-THERM-050-006 4.738 1.3 5030 2.176 2.14E-01 1.0000 0.0010 0.9992 0.00012 74 LEU-COMP-THERM-050-007 4.738 1.3 5030 2.237 2.09E-01 1.0000 0.0010 0.9993 0.00012 75 LEU-COMP-THERM-051-010 2.459 1.636 1.636 15 3.12E-04 1.92E-01 1.0010 0.0019 0.9970 0.00011 76 LEU-COMP-THERM-051-011 2.459 1.636 1.636 28 3.13E-04 1.93E-01 1.0010 0.0019 0.9944 0.00013 77 LEU-COMP-THERM-051-012 2.459 1.636 1.636 92 2.43E-04 1.95E-01 1.0010 0.0019 0.9932 0.00011 78 LEU-COMP-THERM-051-013 2.459 1.636 1.636 395 7.74E-05 2.01E-01 1.0010 0.0022 0.9887 0.00011 79 LEU-COMP-THERM-051-014 2.459 1.636 3.272 121 7.74E-05 1.69E-01 1.0010 0.0019 0.9891 0.00011 80 LEU-COMP-THERM-051-015 2.459 1.636 1.636 487 4.67E-05 2.01E-01 1.0010 0.0024 0.9920 0.00011 81 LEU-COMP-THERM-051-016 2.459 1.636 3.272 197 4.67E-05 1.69E-01 1.0010 0.0020 0.9919 0.00012 82 LEU-COMP-THERM-051-017 2.459 1.636 1.636 634 1.93E-05 2.02E-01 1.0010 0.0027 0.9933 0.00012 83 LEU-COMP-THERM-051-018 2.459 1.636 3.272 320 1.93E-05 1.70E-01 1.0010 0.0021 0.9929 0.00011 84 LEU-COMP-THERM-051-019 2.459 1.636 4.908 72 1.93E-05 1.51E-01 1.0010 0.0019 0.9930 0.00011 85 LEU-COMP-THERM-070-001 6.5 1.49E+00 1.0004 0.0016 1.0055 0.00028 86 LEU-COMP-THERM-070-002 6.5 1.46E+00 1.0004 0.0016 1.0050 0.00031 87 LEU-COMP-THERM-070-003 6.5 1.45E+00 1.0004 0.0016 1.0048 0.00027 88 LEU-COMP-THERM-070-004 6.5 1.43E+00 1.0004 0.0016 1.0047 0.00028 89 LEU-COMP-THERM-070-005 6.5 1.40E+00 1.0004 0.0016 1.0045 0.00025 90 LEU-COMP-THERM-070-006 6.5 1.40E+00 1.0004 0.0016 1.0040 0.00027 91 LEU-COMP-THERM-070-007 6.5 1.39E+00 1.0004 0.0016 1.0039 0.00025 92 LEU-COMP-THERM-070-008 6.5 1.37E+00 1.0004 0.0016 1.0044 0.00027
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# Name (%) (cm) (cm) (ppm) (at/barn) H/U (eV) keff exp keff calc 93 LEU-COMP-THERM-070-009 6.5 1.36E+00 1.0004 0.0016 1.0038 0.00025 94 LEU-COMP-THERM-070-010 6.5 1.34E+00 1.0004 0.0016 1.0039 0.00026 95 LEU-COMP-THERM-070-011 6.5 1.32E+00 1.0004 0.0016 1.0032 0.00027 96 LEU-COMP-THERM-070-012 6.5 1.30E+00 1.0004 0.0016 1.0033 0.00027 97 LEU-COMP-THERM-075-004 6.5 1.45E+00 1.0004 0.0017 1.0061 0.00019 98 LEU-COMP-THERM-075-005 6.5 1.45E+00 1.0004 0.0017 1.0059 0.00020 99 LEU-COMP-THERM-075-006 6.5 1.45E+00 1.0004 0.0017 1.0063 0.00021 100 PAT80SS1 4.74 1.6 2 2.77E-03 3.807 1.47E-01 1.0080 0.0140 1.0000 0.00190 101 PAT80SS2 4.74 1.6 2 2.77E-03 3.807 1.43E-01 1.0060 0.0140 0.9991 0.00170 Note 1: keff was adjusted for experiments LEU-COMP-THERM-001, -002, -051, -070, and -075 (Reference 23) as described in Section 3.3.7.1.1.
3.3.7.4 Trending Analysis A weighted linear regression is performed to evaluate any potential bias that may trend with an independent variable. The equations for the weighted linear regression coefficients are taken from NUREG/CR-6698, Section 2.4.2. The total uncertainty for each experiment, as shown in Table 3-24, is used as the weight. The total uncertainty combines the experimental uncertainty and the calculational uncertainty. The following physical or spectral parameters are investigated:
- U-235 enrichment
- fuel rod pitch
- FA separation
- boron concentration in moderator
- boron areal density in separator plates
- moderator to fuel area ratio
- neutron spectrum as quantified by EALF Each trend analysis uses a different set of experiments as described in Section 3.3.7.2.1 through Section 3.3.7.2.7. The result of the regression analysis is shown in Table 3-23.
The plots shown in Figure 345 through Figure 351 illustrate the data that the regression analysis used. The error bars plotted show the total uncertainty in the estimate, as shown in Table 3-24. Each of the figures also shows a tolerance band, which is calculated as described in NUREG/CR-6698, Section 2.4.4. The tolerance band is shown even if it is not used or is not valid. © Copyright 2018 by NuScale Power, LLC 105
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Some statistical rigor is used to determine if a trend is statistically significant. The factors considered are:
- F test. The F statistic is computed for the weighted least squares fit. If the F statistic is greater than the critical value, which is a function of the sample size and the number of degrees of freedom (n-2 for a linear fit), then the weighted least squares fit is likely to be meaningful.
- T test. The T statistic is computed for the slope from the weighted least squares fit.
The T statistic is the slope divided by the standard error of the slope (which is computed the same as it would be with an ordinary least squares fit, using the coefficients from the weighted least squares fit and unweighted data). If the T statistic is greater than the critical value, then the slope is likely to be meaningful. The critical value is the two-tailed inverse of the Students t-distribution, a function of probability (0.05 is used) and the number of degrees of freedom.
- R2, which ranges from 0 (no correlation) to 1 (perfect correlation).
The trending results are discussed for each parameter below.
- U-235 enrichment: The trend is considered statistically significant, as both the F and T tests are passed and R2 is fairly high. The curve fit residuals did not pass the normality tests so the tolerance band shown on Figure 345 is not considered valid.
As a result non-parametric methods must be applied (see Section 3.3.7.6).
- Fuel rod pitch: The trend is not considered statistically significant as both the F and T tests are failed, and R2 is very low. The tolerance band shown on Figure 346 is not required.
- Fuel assembly separation: The trend is considered statistically significant as both the F and T tests are passed and R2 is moderately high. While the curve fit residuals are not normal, the tests failed by a small margin. As the tolerance band in Figure 347 clearly bounds all of the data by a reasonable margin, the tolerance band is accepted as valid.
- Boron concentration in moderator: The trend is not considered statistically significant as both the F and T tests are failed, and R2 is low. The tolerance band shown on Figure 348 is not required.
- Boron areal density in separator plates. The trend is considered statistically significant as both the F and T tests are passed, even though the R2 value is fairly low. The curve fit residuals passed the normality test so the tolerance band shown on Figure 349 is considered valid.
- Moderator to fuel area ratio: The trend is not considered statistically significant as both the F and T tests are failed, and R2 is very low. The tolerance band shown on Figure 350 is not required.
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- Neutron spectrum: The trend is not considered statistically significant as both the F and T tests are failed, and R2 is very low. The tolerance band shown on Figure 351 is not required.
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Table 3-23 Regression analysis for possible bias trending variables Assembly Plate 10B Area Mod to Fuel Enrichment Pin Pitch Soluble Boron Neutron Parameter Separation Density Area Ratio (wt %) (cm) (ppm) Spectrum (eV) (cm) (at/barn) (unitless) Minimum 3 1.26 0 0 1.93E-05 1.00 0.110 Maximum 6.5 2.54 15.393 5030 4.03E-03 3.88 0.748 Design value or 5 1.25984 7.07136 0 0.00135 1.99 0.187 minimum Design 2000 0.512 maximum Count of 69 86 41 27 23 52 54 Experiments Weighted LSF Slope 2.60402E-03 5.03031E-04 7.53048E-04 7.22059E-07 1.737291 5.86527E-05 2.83303E-03 Standard Error 3.63881E-04 9.52430E-04 1.11599E-04 5.22163E-07 0.581087 3.87985E-04 2.06749E-03 of Slope Intercept 0.986317 0.996385 0.990906 0.995797 0.992048 0.997838 0.997347 Standard Error 0.001851 0.001538 0.000884 0.001191 0.001058 0.000970 0.000587 of Intercept R2 0.574 0.006 0.562 0.217 0.276 0.001 0.038 Normal Residuals No No No (Note 1) Yes Yes No No (Shapiro-Wilk) © Copyright 2018 by NuScale Power, LLC 108
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Normal Residuals No No NA NA NA No No (N15.15) F statistic 49.659 0.279 45.410 1.805 8.617 0.023 1.871 F critical value 3.984 3.955 4.091 4.242 4.325 4.034 4.027 F test result Pass Fail Pass Fail Pass Fail Fail t observed 7.156 0.528 6.748 1.383 2.990 0.151 1.370 t critical (0.05) 1.996 1.989 2.023 2.060 2.080 2.009 2.007 T test result Pass Fail Pass Fail Pass Fail Fail Trend is No. Failed F Yes. Pass F and Yes. Pass F and T No. Failed F and Yes. Pass F and No. Failed F and No. Failed F and statistically 2 and T tests; R2 T test. R high. test. R2 fairly high. T tests; R2 Low T test. T tests; R2 Low T tests; R2 Low significant low Minimum of 0.98513 0.98955 0.97654 0.98598 0.97703 0.99194 0.99211 tolerance band Minimum over 0.99162 0.98995 0.98342 0.98598 0.97971 0.99252 0.99245 NuScale Values Note 1: Normality is assumed; see discussion on fuel assembly separation trending results in Section 3.3.7.4. © Copyright 2018 by NuScale Power, LLC 109
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Figure 345 Regression analysis of U-235 enrichment © Copyright 2018 by NuScale Power, LLC 110
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Figure 346 Regression analysis of fuel rod pitch © Copyright 2018 by NuScale Power, LLC 111
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Figure 347 Regression analysis of fuel assembly separation © Copyright 2018 by NuScale Power, LLC 112
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Figure 348 Regression analysis of dissolved boron concentration © Copyright 2018 by NuScale Power, LLC 113
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Figure 349 Regression Analysis of 10B Area Density in Absorber Plates © Copyright 2018 by NuScale Power, LLC 114
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Figure 350 Regression analysis of moderator to fuel area ratio © Copyright 2018 by NuScale Power, LLC 115
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Figure 351 Regression analysis of neutron spectrum 3.3.7.5 Test for Normality Table 3-24 summarizes the data used to calculate the bias and the bias uncertainty. To ensure that a conservative value of the bias is chosen, all 101 experiments listed in Table 3-22 are utilized. knorm and t are calculated as described in Section 3.3.7.1.1 using the experimental and calculated results shown in Table 3-22. Using the D test as described in Section 3.3.7.1.2, the test statistic is found to be: 12.20434
= = 271.64 2.01856 x 10 The critical values for a sample size of 101 are 278.8 < D < 290.5. The critical values are linearly interpolated between sample sizes of 100 and 120 from ANSI N15.15-1974 (Reference 22). Since D is not within the critical values the distribution of knorm may not be assumed to be normal.
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Table 3-24 Data used for bias and bias uncertainty
# Name k norm total 1 LEU-COMP-THERM-001-001 0.99820 0.003162 2 LEU-COMP-THERM-001-002 0.99850 0.003195 3 LEU-COMP-THERM-001-003 0.99850 0.003195 4 LEU-COMP-THERM-001-004 0.99863 0.003147 5 LEU-COMP-THERM-001-005 0.99612 0.003129 6 LEU-COMP-THERM-001-006 0.99690 0.003162 7 LEU-COMP-THERM-001-007 0.99740 0.003362 8 LEU-COMP-THERM-001-008 0.99718 0.003245 9 LEU-COMP-THERM-002-001 0.99794 0.002077 10 LEU-COMP-THERM-002-002 0.99899 0.002021 11 LEU-COMP-THERM-002-003 0.99855 0.002017 12 LEU-COMP-THERM-002-004 0.99847 0.001822 13 LEU-COMP-THERM-002-005 0.99708 0.001918 14 LEU-COMP-THERM-008-001 0.99678 0.001235 15 LEU-COMP-THERM-008-002 0.99779 0.001239 16 LEU-COMP-THERM-008-003 0.99756 0.001232 17 LEU-COMP-THERM-008-004 0.99711 0.001237 18 LEU-COMP-THERM-008-016 0.99662 0.001239 19 LEU-COMP-THERM-008-017 0.99649 0.001256 20 LEU-COMP-THERM-009-001 0.99876 0.002117 21 LEU-COMP-THERM-009-002 0.99799 0.002120 22 LEU-COMP-THERM-009-003 0.99806 0.002119 23 LEU-COMP-THERM-009-004 0.99841 0.002119 24 LEU-COMP-THERM-009-009 0.99896 0.002115 25 LEU-COMP-THERM-010-005 0.99967 0.002115 26 LEU-COMP-THERM-010-016 1.00311 0.002817 27 LEU-COMP-THERM-010-017 1.00160 0.002815 28 LEU-COMP-THERM-010-018 1.00124 0.002816 29 LEU-COMP-THERM-010-019 1.00128 0.002815 30 LEU-COMP-THERM-012-002 0.98622 0.003512 31 LEU-COMP-THERM-012-003 0.98573 0.003525 32 LEU-COMP-THERM-012-004 0.98899 0.003525 33 LEU-COMP-THERM-012-005 0.98483 0.003520 34 LEU-COMP-THERM-013-001 1.00033 0.001820
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# Name k norm total 35 LEU-COMP-THERM-013-002 0.99998 0.001825 36 LEU-COMP-THERM-013-003 0.99951 0.001820 37 LEU-COMP-THERM-014-001 0.99822 0.001919 38 LEU-COMP-THERM-014-002 0.98608 0.007703 39 LEU-COMP-THERM-014-005 1.00067 0.006903 40 LEU-COMP-THERM-014-006 1.00534 0.003308 41 LEU-COMP-THERM-014-007 1.00066 0.005105 42 LEU-COMP-THERM-016-012 0.99770 0.003289 43 LEU-COMP-THERM-016-013 0.99640 0.003401 44 LEU-COMP-THERM-016-014 0.99746 0.003212 45 LEU-COMP-THERM-039-001 0.99583 0.001432 46 LEU-COMP-THERM-039-002 0.99744 0.001441 47 LEU-COMP-THERM-039-003 0.99690 0.001584 48 LEU-COMP-THERM-039-004 0.99527 0.001638 49 LEU-COMP-THERM-039-005 0.99890 0.001224 50 LEU-COMP-THERM-039-006 0.99786 0.001309 51 LEU-COMP-THERM-039-007a 0.99625 0.001205 52 LEU-COMP-THERM-039-008 0.99646 0.00150 53 LEU-COMP-THERM-039-009 0.99821 0.001471 54 LEU-COMP-THERM-039-010 0.99695 0.001471 55 LEU-COMP-THERM-039-011 0.99533 0.001521 56 LEU-COMP-THERM-039-012a 0.99534 0.001310 57 LEU-COMP-THERM-039-013 0.99597 0.001542 58 LEU-COMP-THERM-039-014 0.99682 0.001598 59 LEU-COMP-THERM-039-015 0.99608 0.001553 60 LEU-COMP-THERM-039-016 0.99663 0.001553 61 LEU-COMP-THERM-039-017 0.99558 0.001575 62 LEU-COMP-THERM-042-003 0.99782 0.001689 63 LEU-COMP-THERM-048-001 0.99848 0.002523 64 LEU-COMP-THERM-048-002 0.99856 0.002516 65 LEU-COMP-THERM-048-003a 0.99853 0.002502 66 LEU-COMP-THERM-048-004 0.99821 0.002512 67 LEU-COMP-THERM-048-005 0.99851 0.002513 68 LEU-COMP-THERM-050-001 0.99813 0.001007 69 LEU-COMP-THERM-050-002 0.99821 0.001007 70 LEU-COMP-THERM-050-003 0.99856 0.001007 71 LEU-COMP-THERM-050-004 0.99795 0.001006
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# Name k norm total 72 LEU-COMP-THERM-050-005 0.99926 0.001007 73 LEU-COMP-THERM-050-006 0.99923 0.001007 74 LEU-COMP-THERM-050-007 0.99926 0.001007 75 LEU-COMP-THERM-051-010 0.99597 0.001903 76 LEU-COMP-THERM-051-011 0.99341 0.001904 77 LEU-COMP-THERM-051-012 0.99221 0.001903 78 LEU-COMP-THERM-051-013 0.98767 0.002203 79 LEU-COMP-THERM-051-014 0.98807 0.001903 80 LEU-COMP-THERM-051-015 0.99102 0.002403 81 LEU-COMP-THERM-051-016 0.99091 0.002004 82 LEU-COMP-THERM-051-017 0.99226 0.002703 83 LEU-COMP-THERM-051-018 0.99193 0.002103 84 LEU-COMP-THERM-051-019 0.99203 0.001903 85 LEU-COMP-THERM-070-001 1.00511 0.001624 86 LEU-COMP-THERM-070-002 1.00455 0.001630 87 LEU-COMP-THERM-070-003 1.00437 0.001623 88 LEU-COMP-THERM-070-004 1.00430 0.001624 89 LEU-COMP-THERM-070-005 1.00413 0.001619 90 LEU-COMP-THERM-070-006 1.00359 0.001623 91 LEU-COMP-THERM-070-007 1.00346 0.001619 92 LEU-COMP-THERM-070-008 1.00395 0.001623 93 LEU-COMP-THERM-070-009 1.00341 0.001619 94 LEU-COMP-THERM-070-010 1.00353 0.001621 95 LEU-COMP-THERM-070-011 1.00284 0.001623 96 LEU-COMP-THERM-070-012 1.00291 0.001623 97 LEU-COMP-THERM-075-004 1.00569 0.001711 98 LEU-COMP-THERM-075-005 1.00550 0.001712 99 LEU-COMP-THERM-075-006 1.00587 0.001713 100 PAT80SS1 0.99206 0.014128 101 PAT80SS2 0.99314 0.014103
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 3.3.7.6 Bias and Bias Uncertainty The distribution of knorm for the 101 experiments used for the bias calculation is shown to not pass the normality test. The non-parametric method of determining the bias must be used, as discussed in NUREG/CR-6698 Section 2.4.4. Note that this also fulfills the requirement from the enrichment trend analysis in Section 3.3.7.4, where the trend was found to be statistically significant but the tolerance band was not valid because the curve fit residuals did not pass the normality tests. For 101 experiments and m=2, using Equation 31 of NUREG/CR-6698, =0.9645. Since this value is greater than 0.95, the second smallest value of knorm may be used to determine kL. Since is greater than 0.90, the Non-Parametric Margin of Table 2.2 of NUREG/CR-6698 is zero. Using equation 33 of NUREG/CR-6698 (with the second smallest value of knorm):
= 0.98573 0.003525 = 0.98221 Where the second smallest value of knorm is taken from experiment LEU-COMP-THERM-012-003 as shown in Table 3-24.
This determines the bias as 0.98221 - 1.0 = -0.01779. Since the kL is a 95/95 value, the bias is as well, so no additional uncertainty term is required. 3.3.7.7 Benchmark Summary The final bias must bound the term determined from the 101 experiments in Section 3.3.7.6 as well as the valid tolerance bands from the trend analysis in Section 3.3.7.4. From Table 3-23, the minimum value of any valid tolerance band occurs for assembly separation at a value of 0.97654. Setting the bias as 0.97654 - 1.0 = - 0.02346 conservatively bounds the statistically significant trends and the bias from the non-parametric analysis of the 101 experiments. Since the tolerance band is calculated as a 95/95 value, the bias of -0.02346 does not require an uncertainty term. 3.3.7.8 Implementation and Use The results presented in Section 3.3.7.7 are applicable to the fuel storage racks over the area of applicability given in Table 3-25. © Copyright 2018 by NuScale Power, LLC 120
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Table 3-25 Area of applicability for bias and bias uncertainty Parameter Range Estimated NuScale Value U-235 enrichment 3 - 6.5 wt% 5 wt% Fuel rod pitch 1.26 - 2.54 cm 1.26 cm FA separation 0 - 15.393 cm 7 cm Soluble boron concentration 0 - 5030 ppm 800 - 2000 ppm Boron areal density in separator 1.93E 4.03E-3 1.35E-3 at/barn plates Moderator to fuel ratio 0.996 - 3.882 1.99 Neutron spectrum (EALF) 0.11 - 0.748 eV 0.187 - 0.512 eV 3.3.8 Summary of Criticality Evaluations The results of the criticality analysis are summarized as follows:
- The normal and accident condition temperatures in the spent fuel pool range from 40 degrees F to 212 degrees F. The criticality analysis is performed at a temperature of 67.73 degrees F, and the k95/95 calculated including additional bias and uncertainty for a minimum temperature at 40 degrees F.
- The maximum k95/95 with full density moderation in unborated water remains below the applicable limit of 1.0 for fuel stored in the storage racks without credit for burnup.
- The maximum k95/95 with full density moderation in 800 ppm (natural boron) borated water remains below the applicable limit of 0.95 for fuel stored in the storage racks without credit for burnup.
- No accident condition is identified that would cause an inadvertent criticality.
- A range of gaps was considered to determine the maximum allowed deformation that maintains k95/95 less than the applicable limit of 0.95. This maximum allowed deformation shall be compared against seismic analysis to confirm that any permanent or transient gap resulting from a seismic event remains within the allowed range.
A summary of the criticality analysis results are presented in Table 3-26. © Copyright 2018 by NuScale Power, LLC 121
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 Table 3-26 Summary of criticality analysis results Conditions k95/95 Limit Flooded with unborated water at full moderator 0.96623 1.0 density Flooded with water at 800 ppm boron, full moderator 0.92191 0.95 density Flooded with water at 800 ppm boron, full moderator 0.92219 0.95 density, dropped FA Flooded with water at 800 ppm boron, rack 0.94954 0.95 deformation of 0.855 cm compression per cell The criticality analysis was completed without consideration for loading patterns or zoning consideration. Based on the conservative assumptions used in the criticality analysis, it is acceptable to place fuel in the racks in all cell locations. 3.4 Material Analysis 3.4.1 Material Evaluation for NuScale Fuel Racks 3.4.1.1 Methodology An important component in the design of the fuel storage racks is the selection of materials. The designated materials must be evaluated for suitability in the established environments and ability to meet the design requirements of the fuel storage racks. Of particular concern to the structural components of the racks is the chemical stability and compatibility of the designated materials with the designed environment and with the FAs themselves. 3.4.1.2 Initial Conditions, Boundary Conditions, and Limitations The structural materials are selected to be compatible with the spent fuel and SFP water environment for a minimum of 60 years. Explicit test data of these materials to extrapolate to a 60-year lifetime is not available. However, the materials specified to be used in the racks are not new. So the design takes credit for the vast successful operational experience of these materials in current light water reactor (LWR) cores and SFPs. These materials are used extensively in similar and more aggressive environments without experiencing any degradation due to pitting, intergranular, galvanic corrosion mechanisms. That experience has demonstrated the credited design strength is not affected. Also, chemical, galvanic, or other reactions among the materials, contents, and the expected SFP water environment that might occur during the 60-year life of the racks are not expected degradation mechanisms for these selected materials. © Copyright 2018 by NuScale Power, LLC 122
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 In these environments, components made of these materials typically fail from extremely high stresses that are not present in these racks. Other possible failure mechanisms are contamination events or manufacturing defects, which are outside the scope of this evaluation. A comparison of the pool chemistry between NuScale and a typical plant is provided in Table 3-27 to show that the expected chemistry conditions in the NuScale design are within the expected range of typical LWRs. Table 3-27 NuScale spent fuel pool water chemistry Criterion Limit Typical LWR pH Variable 4.0 - 5.5 Boric Acid 2000 ppm 2100 - 2700 ppm Conductivity Variable 15 µS/cm Chloride Ions 0.15 ppm maximum 0.15 ppm Fluoride Ions 0.15 ppm maximum 0.15 ppm Sulfate Ions 0.15 ppm maximum 0.15 ppm 3.4.1.3 Analysis, Evaluation, and Data The most significant risk of degradation to the metallic materials used for the structural components of the fuel storage racks is corrosion. The materials selected are strongly resistant to corrosion in the environmental conditions stated in Section 3.4.1.2 and Table 3-27. Low-carbon grade stainless steels demonstrate exceptional corrosion resistance in all but the most extreme environments. Because the proposed water chemistry aligns well with that used in LWRs, there is no reason to believe these corrosion mechanisms would pose a failure or degradation risk to the materials chosen. Preserving these conditions reduces the likelihood of general corrosion. There is no risk of galvanic corrosion between the stainless steel components or with the materials used in a FA (zirconium alloys, stainless steel alloys, and nickel-base alloys). These materials are regularly in contact without any observed galvanic corrosion. The final surface of all components of the racks must meet a 125-microinch surface finish or better. The majority of the structural metallic components are ((
}}2(a),(c) stainless steel alloy which meets the strength requirements of type
(( }} 2(a),(c) while maintaining the low carbon content required for (( }}2(a),(c) . The low-carbon designation is necessary to decrease the susceptibility to material sensitization during welding. Stainless steel gets its corrosion resistance from chromium in solution in the alloy matrix. For austenitic stainless steel, the heat of welding and typical slow cooldown drive the formation of chromium carbides at grain boundaries. This removes the chromium from being available to provide the alloys corrosion resistance in the local area © Copyright 2018 by NuScale Power, LLC 123
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 of the weld. When this occurs, the weld material is considered sensitized and is at a very high risk for stress-corrosion cracking (SCC). To avoid this phenomenon, low-carbon grades of stainless steel are specified. At these levels of carbon, weld sensitization would not occur. For the (( }}2(a),(c) components, the specifications designated are either (( }}2(a),(c) for bar stock. They are chosen to provide the raw material shape most suitable for component fabrication. When a component can be made using either plate or bar stock, both options are listed. A (( }}2(a),(c) material is specified for the rack design as it results in an alloy with superior corrosion resistance, which maintains the high strength of (( }}2(a),(c) components are used extensively in FAs, plant piping, and SFPs of LWRs. Alloy (( }}2(a),(c) has demonstrated acceptable performance in these environments with occasional SCC and chloride-induced SCC events. The ((
}}2(a),(c) has since been introduced to reduce general SCC susceptibility resulting from material sensitization. Alloy (( }}2(a),(c) is still susceptible to chloride-induced SCC, but this is not expected to be a concern given the stringent chloride controls established in Table 3-27. Alloy (( }}2(a),(c) also exhibits a reduced susceptibility to general corrosion. As long as the proposed water chemistry is maintained, with no contamination on the rack materials, there would be no SCC. ((
}}2(a),(c) chosen for the rack feet for the additional material strength. This alloy has a higher strength than standard austenitic stainless steels, beneficial for impact loading scenarios. (( }}2(a),(c) demonstrates similar corrosion resistance as (( }}2(a),(c). It has been used in nuclear plants for high-strength applications, such as valve stems, bolting, and control rod drive components. It is suitable for use in temperatures below 500 degrees F, which is easily satisfied provided the design requirement SFP temperatures are met. Martensitic stainless steel alloys
(( }}2(a),(c) are not subject to the sensitization, so there is no need to require a low-carbon material. Therefore, the material in this environment is not susceptible to corrosion mechanisms. The preferred heat-treatment condition for nuclear applications is H1100, which gives the necessary strength and performance characteristics. The corrosion resistance of (( }}2(a),(c) is similar, so galvanic corrosion between the two is not expected. The alloy (( }}2(a),(c) components do not contact FAs. The generic nomenclature of ((
}}2(a),(c) This allows the manufacturer to choose readily available cap screws which would be compatible with the other components. This material is only used for the cap screws that hold a locking clip in place on the storage rack feet. Because this material meets industry recommendations for bolting materials in a boric acid environment (References 24 and 25), the corrosion characteristics of this material is deemed acceptable.
3.4.1.3.1 Design Requirements for Neutron Absorber Material The neutron absorber used in the fuel storage racks is an MMC consisting only of aluminum alloy and boron carbide, with no polymer or organic components. These © Copyright 2018 by NuScale Power, LLC 124
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 materials are distinguished from earlier boron carbide-aluminum composites by their lack of interconnected porosity and their use of finer boron carbide particles. The MMCs produced by several different suppliers, using both 1000- and 6000-series aluminum matrix alloys, have been subjected to radiation and corrosion testing (Reference 29). The testing has demonstrated neither gamma nor neutron radiation has any effect on MMCs and, because of their lack of interconnected porosity, MMCs are not subject to blistering, swelling, or delamination as a result of corrosion. Testing also demonstrates that relative to pure water, boric acid solutions either have no effect on the corrosion of aluminum and MMCs, or actually inhibit the rate of corrosion. The following requirements are set for the absorber material:
- 1. The material must have sufficient strength and ductility for handling and fabrication, and to support its own weight in the rack.
The design calculations do not take credit for the mechanical properties of the neutron absorber. It is required that tensile testing at room temperature per ASTM B557-15 (Reference 27) shall demonstrate: a) ((
}}2(a),(c)
- 2. The dimensions of the poison plate are defined by the criticality analysis:
a) ((
}}2(a),(c) 3.4.1.3.2 Neutron Absorber Material Specification The MMC material specified by this design shall consist of boron carbide particles in a matrix of aluminum or an aluminum alloy. The material may be supplied either mill finish or anodized.
Criticality analyses have been performed for fuel storage rack design. The analyses demonstrate criticality control to meet the criticality-related requirements of 10 CFR 50.68 Based on those analyses, the MMC used for the fuel storage racks shall have the following characteristics: © Copyright 2018 by NuScale Power, LLC 125
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 ((
}}2(a),(c) 3.4.1.3.3 Neutron Absorber Material Evaluation The two potential environmental deterioration mechanisms are corrosion and radiation damage. Corrosion damage is related to the interconnected porosity of the neutron absorber material. By maintaining a high-density, low-interconnected porosity should minimize corrosion. The MMC is a matrix of full-density aluminum with a fine dispersion of boron carbide particles throughout. Because boron carbide is a ceramic material which is essentially inert from a corrosion or chemical reaction perspective, the corrosion behavior of MMC is essentially the same as that of aluminum. Like aluminum, a thin passive oxide film is formed when exposed to both air and water environments that is tightly bonded to the surface of the material. For the typical pH range of a SFP, the passivated surface of the aluminum (hydrated oxide of aluminum) affords protection against general corrosion because the coating is essentially insoluble, non-porous, and adherent to the surface of the aluminum. This protective surface formed on the aluminum is known to be stable (i.e., limited oxide dissolution) in water environments with a pH range of 4.0 to 10.0 and temperature up to 212 degrees F, thus bounding the typical SFP parameters.
Testing will be performed as part of a coupon monitoring program described in Section 3.4.1.3.4 to show that general corrosion is self-limiting, e.g., due to passivation, and the corrosion process does not result in significant release of boron carbide particles from the surface. From a radiation damage perspective, the neutron absorber material contains no polymers or organic material and, because it has negligible interconnected porosity to trap water, it is insensitive to radiolysis by ionizing (gamma) radiation. Testing of MMCs (References 28, 29, 30) has shown no effect from either gamma or neutron radiation. However, a monitoring program would be established, as described in Section 3.4.1.3.4, to ensure the absorber material would continue to meet the design requirements. © Copyright 2018 by NuScale Power, LLC 126
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 3.4.1.3.4 Neutron Absorber Coupon Monitoring Program In order to verify the continued performance of the neutron absorber over its lifetime in fuel storage racks, a surveillance program shall be established in accordance with ASTM C 1187-2007 (Reference 31). At least twelve coupons shall be immersed in storage racks in the SFP. The size of each coupon shall be large enough to obtain a tensile test specimen (approximately 1 x 8 inches) and a specimen for 10B areal density testing (approximately 2-inch square). All coupons shall be permanently marked to maintain traceability to production lots of MMC used in the racks. Prior to immersion in the racks, the coupon thickness and the 10B areal density shall be measured and recorded, and the coupons shall be photographed. At least one archive specimen shall be retained, and not immersed in the pool. ((
}}2(a),(c)
At approximately (( }}2(a),(c) years from when irradiated fuel is first inserted into the fuel storage racks, at least one coupon shall be removed. The coupon shall be measured and visually examined, and 10B areal density shall be measured, with the following acceptance criteria: a) ((
}}2(a),(c)
If any of the acceptance criteria are not satisfied, an engineering evaluation shall be performed to determine if there is any effect on the performance of the racks safety-related functions and if any corrective action is required. Coupons that are not destroyed may be returned to the pool for continued use in the surveillance program. 3.4.1.4 Implementation and Use The structural components of the fuel storage racks are at risk to degrade through corrosion mechanisms. To mitigate these risks, various stainless steels are chosen for the structural components of the racks. These materials have a proven positive performance in LWRs and are expected to perform well in the SFP environment for the design lifetime of 60 years. © Copyright 2018 by NuScale Power, LLC 127
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1 4.0 Summary and Conclusions Based on the evaluations summarized in this document, the fuel storage rack design as described is in compliance with the regulatory requirements for the design of fuel storage racks,with the exception of the structural analysis. The structural analysis within this report has been intentionally omitted. The COL applicant shall provide the NRC the structural evaluation of the fuel storage rack design and confirmation that the existing structural, thermal-hydraulic, criticality, and material analysis aspects of the design remain valid. This evaluation is dependent on a vendor specific design and the as-built configuration of spent fuel storage racks. The design of the spent fuel storage racks is considered acceptable when it meets the criteria of Appendix D to DSRS 3.8.4. The fuel storage racks allow placement of 5 percent U-235 enrichment and a maximum burn-up of ((
}}2(a),(c) without restrictions for zoning or loading patterns after 72 hours post shutdown of the reactor.
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5.0 REFERENCES
5.1 SOURCES OF INFORMATION RELIED UPON
- 1. U.S. Code of Federal Regulations, General Design Criteria for Nuclear Power Plants, Appendix A, Part 50, Chapter I, Title 10, Energy, (10 CFR 50 Appendix A).
- 2. U.S. Code of Federal Regulations, Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants, Appendix B, Part 50, Chapter I, Title 10, Energy, (10 CFR 50 Appendix B).
- 3. U.S. Code of Federal Regulations, Radiation Protection Programs, Section 20.1101(b),
Part 20, Subpart B, Chapter I, Title 10, Energy, (10 CFR 20.1101(b)).
- 4. U.S. Code of Federal Regulations, Criticality Accident Requirements, Section 50.68, Part 50, Chapter I, Title 10, Energy, (10 CFR 50.68).
- 5. U.S. Nuclear Regulatory Commission, Spent Fuel Storage Facility Design Basis, Regulatory Guide 1.13, Rev. 2, March 2007.
- 6. U.S. Nuclear Regulatory Commission, Damping Values for Seismic Design of Nuclear Power Plants, Regulatory Guide 1.61, Rev. 1, March 2007.
- 7. U.S. Nuclear Regulatory Commission, Combining Modal Responses and Spatial
- 8. Components in Seismic Response Analysis, Regulatory Guide 1.92, Rev. 3, September 2012.
- 9. U.S. Nuclear Regulatory Commission, Information Relevant to Ensuring that Occupational Radiation Exposures at Nuclear Power Stations will be as Low as is Reasonably Achievable, Regulatory Guide 8.8, Rev. 3, June 1978.
- 10. U.S. Nuclear Regulatory Commission, Storage of Components Associated with Fuel Assemblies, Spent Fuel Storage and Transportation Interim Staff Guidance - 9, Rev. 1, (SFST-ISG-9, Rev. 1).
- 11. U.S. Nuclear Regulatory Commission, Standard Review Plan, NUREG-0800.
- 12. U.S. Nuclear Regulatory Commission, Control of Heavy Loads at Nuclear Power Plants, NUREG/CR-0612, July 1980.
- 13. U.S. Nuclear Regulatory Commission, Recommendations for Addressing Axial Burnup in PWR Burnup Credit Analyses, NUREG/CR-6801 (ORNL/TM-2011/273), March 2003.
- 14. U.S. Nuclear Regulatory Commission, Review and Prioritization of Technical Issues Related to Burnup Credit for LWR Fuel, NUREG/CR-6665 (ORNL/TM-1999/303),
February 2000. © Copyright 2018 by NuScale Power, LLC 129
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- 15. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2007 edition and addenda, Section V, Nondestructive Examination, New York, NY.
- 16. American National Standards Institute/American Nuclear Society, Criticality Safety Criteria for the Handling, Storage, and Transportation of LWR Fuel Outside Reactors, ANSI/ANS-8.17-2004 (R2014), reaffirmed July 28, 2014, LaGrange Park, IL.
- 17. American National Standards Institute/American Nuclear Society, Design Requirements for Light Water Reactor Fuel Handling Systems, ANSI/ANS-57.1-1992 (R2015),
reaffirmed June 16, 2015, LaGrange Park, IL.
5.2 REFERENCES
CITED
- 1. U.S. Nuclear Regulatory Commission, Seismic Design Classification for Nuclear Power Plants, Regulatory Guide 1.29, Rev. 5, July 2016.
- 2. U.S. Nuclear Regulatory Commission, Design Specific Review Standard for NuScale SMR Design, New and Spent Fuel Storage, Section 9.1.2, Rev. 0, June 2016.
- 3. U.S. Code of Federal Regulations, Domestic Licensing of Production and Utilization Facilities, Part 50, Chapter I, Title 10, Energy, (10 CFR 50).
- 4. American Society of Mechanical Engineers, Quality Assurance Program for Nuclear Facilities, ASME NQA-1a-2009, 2008 revision with 2009 addenda, New York, NY.
- 5. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2007 edition, Section II, Materials, New York, NY.
- 6. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2007 edition, Section IX, Subsection NCA, Welding and Brazing Qualifications, 2007 addenda, New York, NY.
- 7. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2007 edition, Section II, Part D, Materials - Properties (Customary), 2008 addenda, New York, NY.
- 8. U.S. Nuclear Regulatory Commission, Design Specific Review Standard for NuScale SMR Design, Seismic Design Parameters, Section 3.7.1, Rev. 0, June 2016.
- 9. American Society of Civil Engineers/Structural Engineering Institute, Seismic Design Criteria for Structures, Systems, and Components in Nuclear Facilities, ASCE/SEI 43-05, Reston, VA, 2005.
- 10. U.S. Nuclear Regulatory Commission, Technical Basis for Revision of Regulatory Guidance on Design Ground Motions: Hazard- and Risk-consistent Ground Motion Spectra Guidelines, NUREG/CR-6728, May 2001.
© Copyright 2018 by NuScale Power, LLC 130
Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1
- 11. U.S. Nuclear Regulatory Commission, Design Specific Review Standard for NuScale SMR Design, Guidance on Spent Fuel Pool Racks Section 3.8.4, Rev. 0, June 2016.
- 12. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2007 edition, Section III, Division I, Subsection NF, Supports, New York, NY.
- 13. American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, 2007 edition, Section III, Division 1, Appendices, New York, NY.
- 14. Young, W.C., Roarks Formulas for Stress and Strain, 6th edition, McGraw-Hill, New York, NY, October 1, 1988.
- 15. American National Standards Institute, American National Standard for Radioactive Materials - Special Lifting Devices for Shipping Containers Weighing 10,000 pounds (450 kg) or More, ANSI N14.6-1993, New York, NY.
- 16. American Institute of Steel Construction, Manual of Steel Construction, 9th edition, Chicago, IL, July 1989.
- 17. Blodgett, Omer W., Design of Welded Structures, James F. Lincoln Arc Welding Foundation, July 2016.
- 18. American National Standards Institute/American Nuclear Society, American National Standard for Decay Heat Power in Light Water Reactors, ANSI/ANS-5.1-1979, LaGrange Park, IL.
- 19. Idelchik, I.E., Handbook of Hydraulic Resistance, 3rd edition, Begell House, Danbury, CT 1994.
- 20. U.S. Nuclear Regulatory Commission, Guide for Validation of Nuclear Criticality Safety Calculational Methodology, NUREG/CR-6698, January 2001.
- 21. U.S. Nuclear Regulatory Commission, Criticality Benchmark Guide for Light-Water-Reactor Fuel in Transportation and Storage Packages, NUREG/CR-6361 (ORNL/TM-13211), March 1997.
- 22. American National Standards Institute, Assessment of the Assumption of Normality (Employing Individual Observed Values), ANSI N15.15-1974, reaffirmed 1981, New York, NY.
- 23. Organization for Economic Co-operation and Development, Nuclear Energy Agency, International Handbook of Evaluated Criticality Safety Benchmark Experiments, NEA No. 7231, 2014.
- 24. Owen, D.B., Factors for One-Sided Tolerance Limits and for Variables Sampling Plans, Sandia Corporation Monograph SCR-607, Albuquerque, NM, March 1, 1963.
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Fuel Storage Rack Analysis TR-0816-49833-NP Rev. 1
- 25. Electric Power Research Institute, Guidelines for Threaded-Fastener Application in Nuclear Power Plants, EPRI NP-6316, Palo Alto, CA, August 13, 1989.
- 26. Electric Power Research Institute, Boric Acid Corrosion Evaluation (BACE) Program Phase 1 - Task 1 Report: Industry Experience Reference and Available Test Data Summary, EPRI TR-101108, Palo Alto, CA, December 31, 1993.
- 27. American Society of Test Materials, Standard Test Methods of Tension Testing Wrought and Cast Aluminum- and Magnesium-Alloy Products, ATSM B557-15, ASTM International, West Conshohocken, PA.
- 28. Radiation Damage Testing and Neutron Transmission Testing of Aluminum/Boron Carbide Metal Matrix Composite, Univ. Mass. Lowell RR-1 Rev. 1, July 30, 1998 (submitted to NRC as part of 72-1027 application).
- 29. Electric Power Research Institute, Qualification of METAMIC for Spent Fuel Storage Application, EPRI 1003137, Palo Alto, CA, October 15, 2001.
- 30. Northeast Technology Corp., BORTECTM Qualification Program for Nuclear Fuel Storage Applications, NET 205-02, September 2003.
- 31. American Society of Test Materials, Standard Guide for Establishing Surveillance Test Program for Boron-Based Neutron Absorbing Material Systems for Use in Nuclear Spent Fuel Storage Racks, ASTM C 1187-2007, West Conshohocken, PA.
- 32. American Society of Test Materials, Standard Guide for Examination and Evaluation of Pitting Corrosion, ASTM G46-94 (2013), West Conshohocken, PA.
- 33. American National Standards Institute/American Nuclear Society, Design Requirements for Light water Reactor Spent Fuel Storage facilities at Nuclear Power Plants, ANSI/ANS-57.2-1983 (W1993), withdrawn 1993 - reaffirmation underway, LaGrange Park, IL.
- 34. American National Standards Institute/American Nuclear Society, Design Requirements for New Fuel Storage Facilities at Light Water Reactor Plants, ANSI/ANS-57.3-1983 (W1993), withdrawn 1993 - reaffirmation underway, LaGrange Park, IL.
- 35. Radill, K. C., Palazzolo, A. B., Influence of Temperature and Impact Velocity on the Coefficient of Restitution, NASA Technical Memorandum 106485, Army Research Laboratory Memorandum Report ARL-MR-135, July 1994.
- 36. Marks Standard handbook for Mechanical Engineers, Mc-Grawhill Book Co, Eighth Edition.
- 37. Atlas of Stress-Strain Curves, ASM International, Materials Park, OH, 2nd Edition, 2002.
- 38. Royston, Patrick Approximating the Shapiro-Wilk W-test for non-normality, Statistics and Computing (1992).
© Copyright 2018 by NuScale Power, LLC 132
LO-1018-61991 : Affidavit of Thomas A. Bergman, AF-1018-61992 NuScale Power, LLC 1100 NE Circle Blvd., Suite 200 Corvallis, Oregon 97330 Office 541.360-0500 Fax 541.207.3928 www.nuscalepower.com
NuScale Power, LLC AFFIDAVIT of Thomas A. Bergman I, Thomas A. Bergman, state as follows: (1) I am the Vice President of Regulatory Affairs of NuScale Power, LLC (NuScale), and as such, I have been specifically delegated the function of reviewing the information described in this Affidavit that NuScale seeks to have withheld from public disclosure, and am authorized to apply for its withholding on behalf of NuScale. (2) I am knowledgeable of the criteria and procedures used by NuScale in designating information as a trade secret, privileged, or as confidential commercial or financial information. This request to withhold information from public disclosure is driven by one or more of the following: (a) The information requested to be withheld reveals distinguishing aspects of a process (or component, structure, tool, method, etc.) whose use by NuScale competitors, without a license from NuScale, would constitute a competitive economic disadvantage to NuScale. (b) The information requested to be withheld consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), and the application of the data secures a competitive economic advantage, as described more fully in paragraph 3 of this Affidavit. (c) Use by a competitor of the information requested to be withheld would reduce the competitors expenditure of resources, or improve its competitive position, in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product. (d) The information requested to be withheld reveals cost or price information, production capabilities, budget levels, or commercial strategies of NuScale. (e) The information requested to be withheld consists of patentable ideas. (3) Public disclosure of the information sought to be withheld is likely to cause substantial harm to NuScales competitive position and foreclose or reduce the availability of profit-making opportunities. The accompanying technical report reveals distinguishing aspects about the method by which NuScale develops its fuel storage rack design. NuScale has performed significant research and evaluation to develop a basis for this method and has invested significant resources, including the expenditure of a considerable sum of money. The precise financial value of the information is difficult to quantify, but it is a key element of the design basis for a NuScale plant and, therefore, has substantial value to NuScale. If the information were disclosed to the public, NuScale's competitors would have access to the information without purchasing the right to use it or having been required to undertake a similar expenditure of resources. Such disclosure would constitute a misappropriation of NuScale's intellectual property, and would deprive NuScale of the opportunity to exercise its competitive advantage to seek an adequate return on its investment. (4) The information sought to be withheld is in the enclosed technical report titled Fuel Storage Rack Analysis, TR-0416-49833, Revision 1. The enclosure contains the designation Proprietary" at the top of each page containing proprietary information. The information considered by NuScale to be proprietary is identified within double braces, "(( }}" in the document. (5) The basis for proposing that the information be withheld is that NuScale treats the information as a trade secret, privileged, or as confidential commercial or financial information. NuScale relies upon AF-1018-61992 Page 1 of 2
the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC § 552(b)(4), as well as exemptions applicable to the NRC under 10 CFR § 2.390(a)(4) and 9.17(a)(4). (6) Pursuant to the provisions set forth in 10 CFR § 2.390(b)(4), the following is provided for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld: (a) The information sought to be withheld is owned and has been held in confidence by NuScale . (b) The information is of a sort customarily held in confidence by NuScale and, to the best of my knowledge and belief, consistently has been held in confidence by NuScale. The procedure for approval of external release of such information typically requires review by the staff manager, project manager, chief technology officer or other equivalent authority, or the manager of the cognizant marketing function (or his delegate), for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside NuScale are limited to regulatory bodies, customers and potential customers and their agents, suppliers, licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or contractual agreements to maintain confidentiality. (c) The information is being transmitted to and received by the NRC in confidence. (d) No public disclosure of the information has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or contractual agreements that provide for maintenance of the information in confidence. (e) Public disclosure of the information is likely to cause substantial harm to the competitive position of NuScale, taking into account the value of the information to Nu Scale, the amount of effort and money expended by NuScale in developing the information, and the difficulty others would have in acquiring or duplicating the information. The information sought to be withheld is part of NuScale's technology that provides NuScale with a competitive advantage over other firms in the industry. NuScale has invested significant human and financial capital in developing this technology and Nu Scale believes it would be difficult for others to duplicate the technology without access to the information sought to be withheld. I declare under penalty of perjury that the foregoing is true and correct. Executed on November 5, 2018. AF-1018-61992 Page 2 of 2}}