ML090760776
| ML090760776 | |
| Person / Time | |
|---|---|
| Site: | University of Wisconsin |
| Issue date: | 08/25/2008 |
| From: | Agasie R Univ of Wisconsin - Madison |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RSC 976 | |
| Download: ML090760776 (1) | |
Text
{{#Wiki_filter:UNIVERSITY OF WISCONSIN NUCLEAR REACTOR LICENSE NO. R-74 DOCKET NO. 50-156 LEU CONVERSION REPORT REDACTED VERSION SECURITY-RELATED INFORMATION REMOVED REDACTED TEXT AND FIGURES BLACKED OUT OR DENOTED BY BRACKETS
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Nuclear Reactor Laboratory UWNR University of Wisconsin-Madison 1513 University Avenue, Room 1215 ME, Madison, WI 53706-1687, Tel: (608) 262-3392, FAX: (608) 262-8590 email: reactor~engr.wisc.edu, http://reactor.engr.wisc.edu August 25, 2008 RSC 976 United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555
Subject:
Docket 50-156, License R-74 Request for Amendment No. 17 to Facility License No. R-74
Dear Sirs:
The University of Wisconsin Nuclear Reactor (UWNR) requests to amend the license and technical specifications to facility license number R-74 to facilitate the conversion of the reactor from high enriched uranium (HEU) to low enriched uranium (LEU) in accordance with 10 CFR 50.64(b) (2) (ii). Enclosed is the Safety Analysis Report for the Conversion of the University of Wisconsin-Madison TRIGA Reactor from HEU to LEU Fuel which was prepared in accordance'with NUREG-1537 Part 1, Chapter 18, Guidelines for Preparing and Reviewing Applications for the 'Licensing of Non-Power Reactors; Highly Enriched to Low Enriched Uranium Conversions. I certify under penalty of perjury that the foregoing is true and correct. Sincerely, Executed on: Z-28 o '8 Robert J. Agaile Reactor Director Enclosure cc: Alexander Adams, with enclosure NtLK.
- 0 SAFETY ANALYSIS REPORT FOR THE CONVERSION OF THE UNIVERSITY OF WISCONSIN- MADISON TRIGA REACTOR FROM HEU TO LEU FUEL DOCKET NUMBER 50-156 FACILITY LICENSE NO. R-74 SUBMITTAL REPORT' Documentation of Analyses of Conversion of the University of Wisconsin- Madison TRIGA Reactor from HEU to LEU Fuel
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Submitted By: University of Wisconsin Nuclear Reactor University of Wisconsin- Madison Madison, Wisconsin August 2008 THE UNIVERSITY WISCONSIN 0M A D I S 0 N
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*nalysis August 2008 UWNR LEU Conversion A
V* TABLE OF CONTENTS TABLE OF CONTENTS ..................................................................................................................................... ii LIST OF TABLES ....................................................................................................................................................... v LIST OF FIGURES .................................................................................................................................................. vii I GENERAL DESCRIPTION OF THE FACILITY .................................................................................... I
- 1. 1 IN T RO DUCT ION............................................................................................................................................ 1 1.2
SUMMARY
AND CONCLUSIONS OF PRINCIPAL SAFETY CONSIDERATIONS ............................................. 2 1.3
SUMMARY
OF REACTOR FACILITY CHANGES ....................................................................................... 2 1.4
SUMMARY
OF OPERATING LICENSE, TECHNICAL SPECIFICATIONS, AND PROCEDURAL CHANGES .............. 3 1.5 COMPARISON WITH SIMILAR FACILITIES ALREADY CONVERTED ........................................................ 3 1.5.1 Comparisonto the Texas A&M University (TAMU) Reactor........................ ......................... 4 1.5.2, Comparisonto the Washington State University Reactor (WSU) ........................... 4 2 SITE CHARACTERISTICS ............................................................................................................................. 7 3 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS .......................................................... 9 3.1 INITIAL LEU STORAGE CONTAINER .................................................................................................... 9 4 REACTOR ANALYSIS ................................................................................................................................... 13 4.1 R EA C rO F A C IL IT Y ................................................................................................................................... 13 4 .2 R EAC T O R C O R E......................................................................................................................................... 13 4.2 .1 R eacto r F u e l ................................................................................................................ ... ................. 14 4.2 .2 Co n tro l B lad es .................................................................................................................................... 17 4.2.3 N eutron Moderatorand Refl ector ...................................................................................................... 19 4.2.4 Neutron Startup Source and Holder ............................................... 19 4.2.5 In-Core Exp erimental F acilities ......................................................................................................... 19 4 .2 .6 R eacto r Materials ......................................................................................................................... ..... 20 4.3 REACTOR TANK AND BIOLOGICAL SHIELD ......................................................................................... 22 4.4 C O RE SUPPO RT STRUCTURE ....................................... :... ............ ........................................................ 22 4.5 DYNAMIC DESIGN ..................................... ................... ................................................... 23 4.5.1 HE U F uel Neutronic A nalysis ........................................................................................................ 26 4.5.2 30/20 LEU Fuel Neutronic Analysis .............................................................................................. 37 4.5.3 Comparisons offlux magnitudes between HEU and 30/20 LEU core........................................... 48 4.5.4 Calculated Worth of Fuel Bundle Locations ................................................................................ 49 4.6 FUNCTIONAl, DESIGN OF THE REACTIVITY CONTROL SYSTEM ............................................................ 51 4.7 THERMAL-HYDRAULIC CHARACTERISTICS ........................ ................................................................. 5I 4.7.1 Analysis of Steady State Operation............................... I................................................................. 51 4.7.2 RELAP5 Code Analysis and Results.............................................................................................. 52 4.7.3 HE U Po w er S um mary ........................................................................................................................ 68 4.7.4 HEU Beginning of Life Steady State Core Analysis ...................................................................... 71 4.7.5 HEU Beginning of Life Core Pulse Analysis ................................................................................ 85 4.7.6 LEU Power Summary ........................................ . .... ........... I .. .............. ........ 99
- 4. 7.7. LE U Beginning of Life Core A nalysis............................................................................................... 110 4.7.8 LEU Middle of Life Core Analysis.............................................................. .................... 117 4.7.9 LEU End of Life Core Analysis .............................................. .... 122 4.7.10 LEU Core Pulse Analysis .............................................................................................. .. ............... 129 5 REACTOR COOLANT SYSTEMS ............................................................................................................. 147 0 UWNR LEU Conversion Analysis ii August 2008
6 ENGINEERED SAFETY FEATURES ....................................................................................................... 149 7 INSTRUM ENTATION AND CONTROL SYSTEM S ............................................................................... 151 8 ELECTRICAL POWER SYSTEMS ................................................................. 153 9 AUXILIARY SYSTEM S ............................................................................................................................... 155 10 EXPERIM ENTAL FACILITIES AND UTILIZATION ........................................................................... 157 10.1
SUMMARY
DESCRIPTION ......................................................................................................................... 157 10.2 EXPERIMENTAL FACILITIES .................................................................................................................... 157 II RADIATION PROTECTION PROGRAM AND WASTE MANAGEMENT ......................................... 159 12 CONDUCT OF OPERA TIONS .................................................................................................................... 161 12.1 ORGANIZATION AND STAFF QUALIFICATIONS......................................................................................... 161 12 .2 P RO C ED URES ............................................................................................................................................ 16 1 12.3 OPERATOR TRAINING AND REQUALIFICATION ...................................................................................... 161 12 .4 E M ERG EN C Y P LA N .................................................................................................................................. 16 2 12.5 PHYSICAL SECURITY PLAN ...................................................................................................................... 162 12.6 QUALITY ASSURANCE ............................................................................................................................. 162 12.7 REACTOR RELOAD AND STARTUP PLAN .................. .................................... 164 12.7.1 Accep tance Criteria.......................................................................................................................... 16 7 12 .7 .1. 1 In itia l C ritic a lity .................................................................................................................................. 16 7 12.7.1.2 Control Element.Worth by Rod Drop Method .................................................................................... 167 12.7.1.3 Control Element Worth by Rising Period Rod Bump Method ............................................................ 167 12.7.1.4 Shutdown Margin and Excess Reactivity .... ................................................................................ .. 168 12.7.1.5 Power and Temperature Coefficient .................................................................................................. 168 12.7.1.6 Pulse Mode Operational Testing ......................................................................................................... 168 13 ACCIDENT ANALYSIS ............................................................................................................................... 169 13.1 M AXIMUM HYPOTHETICAL ACCIDENT (M HA) ....................................................................................... 169 13, 1.1 MHA Fission Product Inventory in Fuel Element ............................................................................ 170 13.1.2 MHA Fission Product Release Fraction.............................. .......................................... /............. 171 13.1.3 MHA Activity in Pool Water ............................................................................................................. 174 13.1.4 MHA Fission Product Release to Air within the Reactor Laboratory.............................................. 174 13.1.5 MHA Release of Fission Products to UnrestrictedAreas................................................................. 180 13.1.5.1 M H A D ose to B uilding O ccupant ......................................................................................... ............ 180 13.1.5.2 M H A D ose to O utside B uilding ..................................................................................................... .. 182 13.1.6 Near MHA With Pool Intact and Ventilation System Inoperable..................................................... 183 13.1.6.1 Near MHA W ith Pool Intact Fission Product Release to Air W ithin the Reactor Laboratory ............ 185 13.1.6.2 Near MHA With Pool Intact Dose to Building Occupant ................................. 185 13.1.6.3 Near MHA With Pool Intact Dose to Outside Building ...................................................................... 186. 13.1.7 Near MHA With Pool Drainedand Ventilation System Operable.................................................... 186 13.1.7.1 Ventilation System Effective Stack Height ......................................................................................... 187 13.1.7.2 Maximum Ground Level Concentration ............................................................................................. 188 13.1.7.3 Near MHA With Ventilation Intact Dose to Outside Building .......................................................... 190 13.1.8 - Near MHA With Pool Intact and Ventilation System Operable........................................................ 192 13.1.8.1 Near MHA With Pool and Ventilation Intact: Dose to Outside Building ........................................... 192 13.1.9 Expected Dose to Public........ ............................. ...... ............ 193 13.2 RAPID ADDITION OF REACTIVITY ACCIDENT .......................................................................................... 196 13.2.1 Temperatureafter Pulsefor HEU BOL ................................................ I........................................ 196 13.2.2 Temperatureafter Pulsefor LEU BOL ............................................................................................ 201 13.2.3 Temperature after Pulsefor LEU MOL ........................................................................................... 205 13.2.4 Temperature after Pulsefor LEU EOL ................ : ............................................................................ 208 UWNR LEU Conversion Analysis ii i August 2008
13.3 REDUCTION-IN-COOLING ACCIDENTS ..................................................................................................... 212 13.3.1 Possible Means of Water Loss .................................................. 212 13.3.2 Radiation Levels Due to Unshielded Core ......................................... ............................................. 213 13.3.3 Fuel Temperature after Loss of Pool Water ..................................................................................... 215 13.. 3.4 L oss of Coo lant F lo w ........................................................................................................................ 2 30 13.4 OTHER ACCIDENTS ................................................................................................................................. 230 14 TECHNICAL SPECIFICATIONS ............................................................................................................... 231 15 OTHER LICENSE CONSIDERATIONS ................................................................................................... 241 15.1 PRIOR UTILIZATION OF REACTOR COMPONENTS ............................................ . . ......................... 241 15.2 LICENSE CONDITIONS .......................................... I.............................................................................. 241 15.3 DECOMMISSIONING ........... .......................................... ........................................ 242 REFERENCES ........................................................................................................................................................ 243 iv August 2008 UW-NR LEU UWNR Conversion Analysis LEU Conversion Analysis iv August 2008
LIST OF TABLES Table 1.5.1 Comparison of TAMU, WSU, and UWNR conversions ..................................................................... 5 Table 3.1. I[M ultiplication factors of the LEU initial storage container .................................................................. II Table 4.2.1 Core Com ponents for HEU and LEU cores ......................................................................................... 14 Table 4.2.2 Summary of HEU and LEU fuel characteristics ..................................... 17 Table 4.2.3 H EU and LEU fuel com positions ....................................................................................................... 21 Table 4.2.4 Non-fuel material compositions for UWNR core components ............................................................ 22 Table 4.5.1 Initial positions of transient rod at different HEU core lifetime ....................................................... 28 Table 4.5.2 The critical bank height for hot and cold operating condition of HEU fuel ...................................... 28 Table 4.5.3 Summary of fitting parameters and integral worth of each control element for HEU fuel .................. 32 Table 4.5.4 Effective delayed neutron fraction at different core lifetime .............................................................. 33 Table 4.5.5. Prom pt neutron lifetim e at different core lifetim es ............................................................................. 34 Table 4.5.6 Void and Coolant Temperature Coefficients for HEU at different core lifetimes ...................... 35 Table 4.5.7 Negative prompt temperature coefficients for HEU fuel ..................................................................... 35 Table 4.5.8' Initial positions of transient rod at different 30/20 LEU core lifetime .................................................... 41 Table 4.5.9 Thecritical bank height for hot and cold opcrating-condition of 30/20 LEU fuel .........................41 Table 4.5.10 Fitting parameters and integral worth of each control element for 30/20 LEU fuel ........................ 44 Table 4.5.11 Effective delayed neutron fraction at different core lifetime ............................................................ 45 Table 4.5.12 Prompt neutron lifetime at different core lifetimes .......................................................................... 45 Table 4.5.13 Void Coefficient for 30/20 LEU at different core lifetimes ............................................................. 45 Table 4.5.14 Negative prompt temperature coefficients for 30/20 LEU fuel ............................................................. 46 Table 4.7.1 Hydraulic input conditions for the hot channel (HEU).: ................................................................... 54 Table 4.7.2 Therm al hydraulic parameters used for the hot channel ..................................................................... 55 Table 4.7.3 Hydraulic parameters and CHF results at limiting locations in the core ..................... 57 Table 4.7.4 Axial nodilization of the single channel RELAP5/MOD3.3 model ..................................................... 60 Table 4.7.5 Radial nodilization of the single channel RELAP5/MOD3.3 model................................. 63 Table 4.7.6 Calculated vs. measured fuel temperatures at E4 SE with.0,01 mil gap for HEU BOL core ................ 76 Table 4.7.7 Steady state results of hot rod at core power of 1.5, 1.3, and 1.0 MW (HEU BOL) ........................... 78 0 Table 4.7.8 Predicted IFE measurements in D5 SW if IFE could be placed in D5 SW (HEU BOL) ................79 Table 4.7.9 Steady state CHF results of hot rod for core powers of 1.5, 1.3, and 1.0 MW (HEU BOL) .................. 83 Table 4.7.10 Thermal hydraulic parameters of the 2 channel model ...................................... 86 Table 4.7.11 Fuel tem perature coefficient for HEU BO L ............................. ......................................................... 8.8 Table 4.7.12 Steady state results of hot rod at 1.5, 1.3, and 1.0 MW (LEU BOL) .............................................. 11 I Table 4.7.13 Steady state CHF resulis of hot rod for core powers of 1.5, 1.3, and 1.0 MW (LEU BOL) ................. 114 Table 4.7.14 Expected IFE Temperature results at 1.0 M W (LEU BOL) ................................................................ 116 Table 4.7.15 Steady state results of hot rod at 1.5, 1.3, and l.0.MW (LEU MOL) ............................ 119 Table 4.7.16 Steady state CHF results of hot rod for core powers of 1.5, 1.3, and 1.0 MW (LEU MOL) ................ 121 Table 4.7.17 Steady state results of hot rod at 1.5, 1.3' and 1.0 MW-(LEU EOL) .................................................. 124 Table 4.7.18 Steady state CHF results of hot rod for core powers of 1.5, 1.3, and 1.0 MW (LEU EOL) ................. 127 Table 4.7.19 Fuel tem perature coeffi cient for LEU BOL ......................................................................................... 130 Table 4.7.20 Fuel tem perature coefficient for LEU M O L ........................................................................................ 130 Table 4.7.21 Fuel tem perature coeffi cient for LEU EOL ......................................................................................... 131 Table 4.7.22 Summary of pulse conditions as a function of pulse size (LEU BOL) ..................... 136 Table 4.7.23 Summary of pulse conditions as a function of pulse size (LEU MOL) ............................................... 141 Table 4.7.24 Summary of pulse conditions as a function of pulse size (LEU EOL) .......................... 146 Table 10.2.1 Experim ental Facility Reactivity Effects ................. .............................. ....................................... 157 T able 13.1.1 M H A Fission Product Inventory ........................................................................................................ 171 Table 13.1.2 MHA Released Fission Product Inventories ................................................................................. 173 Table 13.1.3 M H A O ccupational Lab Concentration ............................................................................................. 177 Table 13.1.4 M H A O ccupational External D ose by Isotope ................................................................................... 178 UWNR LEU Conversion Analysis V Augustv 2008
Table 13.1.5 MHA Occupational Thyroid Dose by Isotope ....................................
. 179 . Table 13.1.6 MHA Total Occupational Dose during 5 minute evacuation .............................................................. 179.
Table 13.1.7 Building V olum es used for Ground Release ........................................................................................ 181 Table 13.1.8 MHA Building Occupant Doses for Ground Release ........................................................................... 181 Table 13.1.9 M HA D ose to O utside Building ........................................................................................................ 183 Table 13.1.10 Near MHA with Pool Intact Released Fission Product Inventories .................................................. 184 Table 13.1.11 Near MHA with Pool Intact Occupational Dose during 5 minute evacuation .................................. 185 Table 13.1.12 Near MHA with Pool Intact Building Occupant Doses for Ground Release .................................... 185 Table 13,1.13 Near MHA With Pool Intact Dose to Outside Building .................................................................. 186 Table 13.1.14 Near MHA With Ventilation Intact: Public Concentration by Isotope .......................... 191 Table 13.1.15 Near MHA With Ventilation Intact: Dose to Outside Building ................................ ;....................... 192 Table 13.1.16 Near MHA With Pool and Ventilation Intact: Dose to Outside Building ...................... 193 Table 13.1.17 Expected Fission Product Inventories .............................................................................................. .194 Table 13.1.18 Expected Fission Product Release .................................................................................................... 195 Table 13.1.19 Expected D ose to Outside Building .................................................................................................. 196 Table 13.3.1 Calculated radiation dose rates after pool water is lost ......................................................................... 214 Table 13.3.2 Input conditions to determine the power profile for the LOCA transient ........................................... 225 Table 13.3.3 Summary of LOCA temperature results ........................................ 229 Table 13.4.1 Fuel and reflector handling accident reactivity effects ....................................................................... 230 J SUWNR LEU Conversion Analysis vi August 2008 v ,
LIST OF FIGURES Figure 3.1.1 Solid works model of the initial LEU storage container .................................................................... 10 Figure 3.1.2 Layout of the MCNP storage rack model. (Left) Top view (Right) Side view ............................... 10 Figure 4.2.1 Layout of U W NR core ............................................................................................................................ 14 Figure 4.2.2 Fuel elem ent dim ensions......................................................................................................................... 16 Figure 4.2.3 (Left) Safety control blade; (Right) Safety control blade shroud ............................................................ 18 Figure 4.2.4 Three-element assembly with guide tube for transient rod ................................................................. 19 Figure 4.5.1 Axial cross section of the MCNP5 model for HEU and LEU 30/20 core analysis ............................ 24 Figure 4.5.2 Horizontal cross section of the MCNP5 model for HEU and LEU core analysis ............................. 25 Figure 4.5.3 Excess reactivity of UWNR as a function of burnup from REBUS-MCNP ...................................... 26 Figure 4.5.4 Differential worth curve of Control Blade I for HEU fuel ................................................................. 30 Figure 4.5.5 Differential worth curve of Control Blade 2 for HEU fuel ................................................................. 30 Figure 4.5.6 Differential worth curve of Control Blade 3 for HEU fuel ................................................................. 31 Figure 4.5.7 Differential worth curve of the Regulating Blade for HEU fuel ...................................................... 31 Figure 4.5.8 Differential worth curve of the Transient Rod for HEU fuel ........................................................... 32 Figure 4.5.9 Negative prompt temperature coefficients as a, function of fuel temperature for HEU fuel at different co re lifetim es ............................................................................................................................................................... 37 Figure 4.5.10 Excess reactivity of UWNR LEU core as a function of bumup from REBUS-MCNP .................... 39 Figure'4.5.11 N ew LEU Core Configuration ........................................................................................................ 40 Figure 4.5.12 Differential worth curve of Control Blade 1 for 30/20 LEU fuel ..................................................... 42 Figure 4.5.13 Differential worth curve of Control Blade 2 for 30/20 LEU fuel ..................................................... 42 Figure 4.5.14 Differential worth curve of Control Blade 3 for 30/20 LEU fuel ..................................................... 43 Figure 4.5.15 Differential worth curve of the Regulating Blade for 30/20 LEU fuel ............................................ 43 Figure 4.5.16 Differential worth curve of the transient rod for 30/20 LEU fuel .................................................... 44 Figure 4.5.17 Negative prompt temperature coefficients as a function of fuel temperature for 30/20 LEU fuel at different core lifetim es ................................................................................................................................................ Figure 4.5.18 Ratios of flux measurements between LEU and HEU of 225 zones on the thermal column surface....49 48 08 Figure 4.5.19 LEU BOL Fuel Bundle Worth in % Ak/k, With Reflectors ........................................................... 50 Figure 4.5.20 LEUBOL Fuel Bundle Worth in % Ak/k, Without Reflectors .......................................................... 50 Figure 4.7.1 Diagram of subchannel used to model the UWNR hot rod 12................................................................ 55 Figure 4.7.2 A xial N odilization of Fuel Rod ........... ;.............................................................................................. 58 Figure 4.7.3 Radial nodilization in fuel element (not to scale)............................................................................... 61 Figure 4.7.4 Radial nodilization in fuel element (not to scale) ............................................................................... 62 Figure 4.7.5 Pin power [kW/rod] and power peaking factors (PPF) of the UWNR core at 1.0 MW (HEU BOL) ..... 69 Figure 4.7.6 Pin Power [kW/rod] and power peaking factors (PPF) of the UWNR core at 1.0 MW (HEU MOL) ..... 70 Figure 4.7.7 Pin Power [kW/rod] and power peaking factors (PPF) of the UWNR core at 1.0 MW (HEU EOL) ...... 70 Figure 4.7.8 Normalized axial power density distribution of hot rod (HEU BOL) ....................... 72 Figure 4.7.9 Normalized radial power density distribution of hot rod (HEU BOL) .......................................... 72 Figure 4.7.10 Radial temperature profile at E4 SE with 0.1 mil gap (12.1 kW/rod) and core power of 1.0 MW ...... 74 Figure 4.7.11 Radial temperature profile of hot rod at core power of 1.5 MW with varying gap widths (HEU BOL)
........I ....................................................................................................................... ................................................... 7 7 Figure 4.7.12 Axial temperature profile of hot rod at core power of 1.5 MW with 0. 1 mil gap (HEU BOL) ............ 77 Figure 4.7.13 Temperature profile of hot rod as a function of power (HEU BOL) .............................................. 80 Figure 4.7.14 Coolant flow rate of hot rod as a function of power (HEU BOL) ........................ 81 Figure 4.7.15 MDNBR of hot rod as a function of power (HEU BOL) ........ 7....... !................................................ 83 Figure 4.7.16 Power vs. flow rate of hot rod (Correlation power correlates to MDNBR = 1.00) (HEU BOL) .......... 84 Figure 4.7.17 The 2 channel RELAP5/MOD3.3 model schematic ........................................................................ 87 Figure 4.7.18 Prompt Negative Fuel Temperature Coefficient vs. AverageCore Temperature (HEU BOL) ............ 89 Figure 4.7.19 Power vs. Time of 1.34 %Ak/k ($1.78) Pulse for HEU BOL ............. ............................................ 91 UWNR LE1I Conversion Analvsis vii Aiuaust 92M1
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Figure 4.7.20 Integral energy profile of pulse vs. tim e at the BO L ......... ................................................................ 92 Figure 4.7.21 Maximum Pulse Temperature vs. Time for a 1.4%Ak/k reactivity insertion at D5SW ........... 93 Figure 4.7.22 Power Profile of 1.4%Ak/k pulse versus time for HEU BOL ......................................................... 94 Figure 4.7.23 Energy of 1.4% Akik pulse vs. time for HEU BO L .......... ;.................................................................... 94 Figure 4.7.24 Fuel temperature [°C] surface plot of D5 SW 0.06s after 1.4%Ak/k pulse (HEU BOL) ................. 95 Figure 4.7.25 Axial temperature distribution of hot pin 0.06s after pulse 1.67cm from centerline (HEU BOL) ....... 96 Figure 4.7.26 Radial temperature distribution of hot pin 0.06s after pulse 16.51 cm from bottom of fuel (HEU BOL)
.................... .. ............... .................... ....................................................................................................................... 96 Figure 4.7.27 Maximum pulse temperature in hot rod as a function of pulse size (HEU BOL) ............................ 97 Figure 4.7.28 Maximum pulse power as a function of reactivity insertion (HEU BOL) ....................................... 98 Figure 4.7.29 Total pulse energy as a function of reactivity insertion (HEU BOL) ............................................... 99 Figure 4.7.30 Pin Power [kW/rod] and power peaking factors (PPF) of the UWNR core at 1.0 MW (LEU BOL).. 100 Figure 4.7.31 Pin Power [kW/rod] and power peaking factors (PPF) of the UWNR core at 1.0 MW (LEU MOL). 101 Figure 4.7.32 Pin Power [kW/rod] and power peaking factors (PPF) of the UWNR core at 1.0 MW (LEU EOL).. 101 Figure 4.7.33 Normalized radial power density distribution of hot rod for LEU core ...................... 102 Figure 4.7.34 Normalized axial power distribution of hot rod for LEU core .......................................................... 102 Figure 4.7.35 Radial power distribution comparison (LEU BOL) .................................... 105 Figure 4.7.36 A xial pow er distribution com parison (LEU BO L) .............................................................................. 105 Figure 4.7.37 Thermocouple temperature as a function of pin power peaking factor ............................................. 106 Figure 4.7.38 Predicted maximum IFE temperatures across the core at a power of 1.816 MW (LEU BOL) ........... 108 Figure 4.7.39 Possible thermocouple locations of the UWNR (LEU BOL) .............................................................. 109 Figure 4.7.40 Radial temperature distribution of hot rod at 1.5 MW for varying gap sizes (LEU BOL) .................. 110 Figure 4.7.41 Axial temperature profile of hot rod at 1.5 M W (LEU BOL) ............................................................. 112 Figure 4.7.42 Temperature profile of hot rod at D5 SW vs. power of hot rod (LEU BOL) ...................................... 112 Figure 4.7.43 Coolant flow rate of hot rod at D5 SW vs. power of hot rod (LEU BOL) .......................................... 113 Figure 4.7.44 M DNBR as a function of hot rod power (LEU BOL) ......................................................................... 114 S Figure 4.7.45 Power to reach CHF as a function of flow rate (LEU BOL) ...............................................................
Figure 4.7.46 Radial temperature distribution of hot rod at 1.5 MW, for varying gap sizes (LEU MOL) ................. 118 Figure 4.7.47 Axial temperature profile of hot rod at 1.5 MW (LEU MOL) ................................. ........................... 118 115 Figure 4.7.48 Temperature profile of hot rod at D5 SW vs. power of hot rod (LEU MOL) ..................................... 120 Figure 4.7.49 Coolant flow rate of hot rod at D5 SW vs. power of hot rod (LEU MOL) ......................................... 120 Figure 4.7.50 M DNBR as a function of hot rod power (LEU M OL) ........................................................................ 121 Figure 4.7.51 Power to reach CHF as a function of flow rate (LEU MOL) .............................................................. 122 Figure 4.7.52 Radial temperature distribution of hot rod at 1.5 MW for varying gap sizes (LEU EOL).................. 123 Figure 4.7.53 Axial temperature profile of hot rod at 1.5 M W (LEU EOL) .............................................................. 125 Figure 4.7.54 Temperature profile of hot rod at D5 SW vs. power of hot rod (LEU EOL) .................. 125 Figure 4.7.55 Coolant flow rate of hot rod at D5 SW vs. power of hot rod (LEU EOL) ........................................... 126 Figure 4.7.56 M DNBR as a function of hot rod power (LEU EOL) ......................................................................... 127 Figure 4.7.57 Power to reach CHF as a function of flow rate (LEU EOL) ............................................................... 128 Figure 4.7.58 Maximum fuel temperature of hot rod after 1.4 %Ak/k pulse (LEU BOL) ......................................... 132 Figure 4'7.59 Power and total energy of 1.4 % Ak/k pulse (LEU BOL) ..................................................................... 133 Figure 4.7.60 Temperature surface plot of hot rod at 0.0655s after 1.4%Ak/k pulse (LEU BOL) ............................ 133 Figure 4.7.61 Axial temperature distribution of hot rod after 1.4%Ak/k pulse (LEU BOL) .................................... 134 Figure 4.7.62 Radial temperature distribution of hot rod after 1.4%Akik pulse (LEU BOL) ................................... 134 Figure 4.7.63 Maximum pulse temperature as a function of pulse size (LEU BOL) ........................ .................. 135 Figure 4.7.64 Maximum power and total energy as a function of pulse size (LEU BOL) ....................................... 135 Figure 4.7.65 Maximum fuel temperature of hot rod after 1.4 %Ak/k pulse (LEU MOL) ........................................ 137 Figure 4.7.66 Power and total energy of 1.4 % Ak/k pulse (LEU M OL) ................................. ........................... 138 Figure 4.7.67 Temperature surface plot of hot rod at 0.0650s after 1.4%A k/k pulse (LEU MOL) .......................... 138 Figure 4.7.68 Axial temperature distribution of hot rod after 1.4%k/k pulse (LEU MOL)..................................... 139 Figure 4.7.69 Radial temperature distribution of hot rod after 1.4%Ak/k pulse (LEU MOL) .................................. 139 Figure 4.7.70 Maximum pulse temperature as a function of pulse size (LEU MOL) .............................................. 140 UW RLUCovrin nlss iiAgut20 UWN-R LEU Conversion Analysis viii August 2008
Figure 4.7.71 Maximum power and total energy as a function of pulse size (LEU MOL) ..................................... 140 Figure 4.7.72 Maximum fuel temperature of hot rod after 1.4 %Ak/k pulse (LEU EOL) ......................................... 142 Figure 4.7.73 Power and total energy of 1.4 % Ak/k pulse (LEU EOL) .................................................................... 143 Figure 4.7.74 Temperature surface plot of hot rod at 0.0595s after 1.4%Ak/k pulse (LEU EOL) ..................... :...... 143 Figure 4.7.75 Axial temperature distribution of hot rod after 1.4%Ak/k pulse (LEU EOL) .................................... 144 Figure 4.7.76 Radial temperature distribution of hot rod after 1.4%Ak/k pulse (LEU EOL) ................................... 144 Figure 4.7.77 Maximum pulse temperature as a function of pulse size (LEU EOL) ................................................ 145 Figure 4.7.78 Maximum power and total energy as a function of pulse size (LEU EOL) ................. 145 Figure 13.1.1 Ground Level Concentration vs. Distance (Q= 1Ci/s) .......................................................................... 190 Figure 13.2.1 Core Power and Energy vs. Time after 1.4%Ak/k Pulse at 1.3 MW plotted for 2s (lIEU BOL) ...... 197 Figure 13.2.2 Temperature Profile vs. Time after 1.4%Ak/k pulse at 1.3 MW plotted for 2s (HEU BOL) ............. 198 Figure 13.2.3 Core Power vs. Time after 1.4%Ak/k pulse at 1.3 MW following SCRAM (HEU BOL) ................. 199 Figure 13.2.4 Temperature Profile vs. Time of 1.4% k/k pulse at 1.3 MW following SCRAM (HEU BOL) .......... 199 Figure 13.2.5 Fuel temperature plot after 1.4 %Ak/k pulse at 1.3 MW at 0.0410s (left) & 2.000s (right) (HEU BOL) ................................................................................................................................................................................... 2 01 Figure 13.2.6 Power and total energy after 1.4%Ak/k pulse at 1.3 MW (LEU BOL) .................... 202 Figure 13.2.7 Maximum hot rod temperature after 1.4%Ak/k pulse at 1.3 MW (LEU BOL) ................................... 202 Figure 13.2.8 Core power after blades SCRAM in at 2 seconds following 1.4%Ak/k pulse at 1.3 MW (LEU BOL) ...................................................................................................... I ............... 203 Figure 13.2.9 Max fuel temperature after blades SCRAM at 2s following a.l.4% Ak/k pulse at 1.3 MW (LEU BOL) .... ....................................................... ............... . ...................................................................................................... 203 Figure 13.2.10 Fuel temperature plot after 1.4 %Ak/k pulse at 1.3 MW at 0.0460s (left) & 2.252s (right) (LEU B O L ) .......................................................................................................................................................................... 2 04 Figure 13.2.11 Power and total energy after .1.4%Ak/k pulse at 1.3 MW (LEU MOL) ........................................... 205 Figure 13.2.12 Maximum hot rod temperature after 1.4%Ak/k pulse at 1.3 MW (LEU MOL) ................................ 206 Figure 13.2.13 Core power after blades SCRAM in at 2 seconds following 1.4%Ak/k pulse at 1.3 MW (LEU MOL) ......................... ........................................................................................................................................................ 20 6 Figure 13.2.14 Max fuel temperature after blades SCRAM at 2s following a.l.4%Ak/k pulse at 1.3 MW (LEU M O L ) ......................................................................................................................................................................... 2 07 Figure 13.2.15 Fuel temperature plot after 1.4 %Ak/k pulse at 1.3 MW at 0.0485s (left) & 2.252s (right) (LEU M O L ) ........................................................ ................................................................................................................ 208 Figure 13.2.16 Power and total energy after 1.4%Ak/k pulse at 1.3 MW (LEU EOL) ............................................ 209 Figure 13.2.17 Maximum hot rod temperature after 1.4%Ak/k pulse at 1.3 MW (LEU EOL) .................... 209 Figure 13.2.18 Core power after blades SCRAM in at 2 seconds following 1.4%Ak/k pulse at 1.3 MW (LEU EOL)
........................................... .. . . . . . . .................................................. .................. 2 10 Figure 13.2.19 Max fuel temperature after blades SCRAM at 2s following a.l.4%Ak/k pulse at 1.3 MW (LEU E O L) ........................................................................ . . ........................................................................................... 2 10 Figure 13.2.20 Fuel temperature plot after 1.4 %Ak/k pulse at 1.3 MW at 0.0460s (left) & 2.127s (right) (LEU E O L ) ............................................................. ....... I . ......... ...................................................... .......... 2 1 1 Figure 13.3.1 Strength and applied stress as a function of temperature for 1.7 and 1.6 H-Zr TRIGA fuel ............... 219 Figure 13.3.2 Total core power used in LOCA starting at 1.02 MW (HEU BOL) ................................................... 222 Figure 13.3.3 Total core power used for air cooled portion of LOCA transient (HEU BOL) ................................. 223 Figure 13.3.4 Temperature profile during LOCA transient (HEU BOL) ................................. 224 Figure 13.3.5 Power Profile used in LOCA starting at 1.02 MW (LEU BOL) ............................... .................. 226 Figure 1.3.3.6 Power Profile used in LOCA starting at 1.02 MW (LEU MOL) ............................. 226 Figure 13.3.7 Power Profile used in LOCA starting at 1.02 MW (LEU EOL) ......................................................... 227 Figure 13.3.8 Temperature profile during LOCA transient (LEU BOL) .................................................................. 227 Figure 13.3.9 Temperature profile during LOCA transient (LEU M OL) ................................................................. 228 Figure 13.3.10 Temperature profile during LOCA transient (LEU EOL) ................................................................ 228 UWNR LEU Conversion Analysis +
ix August 2008
1 GENERAL DESCRIPTION OF THE FACILITY 1.1 Introduction This report contains the results of the design, safety, and accident analyses performed for the conversion of the University of Wisconsin Nuclear Reactor (UWNR) from the use of highly enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel. The UWNR is a 1 megawatt (MW) TRIGA research reactor located in Madison, Wisconsin. The only facility changes required for this conversion are to replace the current FLIP/HEU fuel elements with TRIGA LEU fuel elements containing 30 weight percent uranium enriched to less than 20% inU-235 (30/20 LEU), and to add a new fresh fuel storage facility for temporary storage of LEU fuel until the core conversion can take place. The objective of this study was to design an LEU core with similar operational capabilities as the original HEU core and with .acceptable safety margins under both normal and accident conditions. In order.to provide comparisons between the proposed LEU core and the original-UWNR HEU FLIP core, complete analyses were performed for both cores. The proposed 30/20 LEU core consists of 8 1 non-instrumented fuel elements and 2 instrumented fuel elements (IFEs). The control elements of the core will remain unchanged: three shim-safety blades, one stainless steel regulating blade; and one transient control rod. In the following sections of this report, it is shown that throughout the lifetime of the proposed LEU core: UWNR LEU Conversion Analysis 1 August 2008 C-
" Shutdown margin meets Technical Specification limits " Reactivity coefficients meet required limits and are similar to the existing HEU core
- Fuel integrity is maintained under all operating conditions 0 Dose to public from Maximum Hypothetical Accident (MHA) is below maximum permissible limits.
1.2 Summary and Conclusions of Principal Safety Considerations in the following sections of this Conversion Analysis Safety Analysis Report (CA SAR), it is shown that the new UWNR LEU core will meet or exceed all safety requirements. Chapter 4, Reactor Analysis, presents the neutronics and thermal-hydraulics analyses and discusses the
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differences in reactor characteristics due to the fuel conversion. Chapter 12, Conduct of Operations, provides the reload and startup plan for the new UWNR LEU core. Chapter 13; Accident Analysis, discusses differences in hypothetical accident characteristics and consequences. Chapter 14, Technical Specifications, describes changes to the technical specifications. Other topics covered by this CA SAR-represent minor changes from the existing UWNR HEU Safety Analysis Report (HEU SAR).' 1.3 Summary of Reactor Facility Changes Two facility changes are required for this conversion: I. To replace the current FLIP/HEU fuel elements with TRIGA LEU fuel elements containing 30 weight percent uranium enriched to less than 20% in U-235 (30/20 LEU). The dimensions of the LEU fuel elements are identical to the FLIP fuel elements. In addition, the UWNR LEU Conversion Analysis 2 August 2008,
LEU 30/20 fuel has been approved by the Nuclear Regulatory Commission (NRC) for use in 2 non-power reactors. Section 4.2.1 provides a detailed description of the proposed LEU fuel.
- 2. To develop a new fuel storage facility, which will be analyzed within the LEU analysis of this document.
1.4 Summary of Operating License, Technical Specifications, and Procedural Changes The UWNR operating license will be amended to allow possession of both the HEU and LEU fuel inventories. during conversion as described in section 15.2. No significant procedure changes are anticipated other than creating new procedures for the Reload and Startup Testing Program as detailed in section 12.6. Changes to the Technical Specifications are detailed in section 14; significant changes include:
- Changing all references to fuel or fuel type from StandardiFLlP to LEU 30/20 fuel.
Removing all references to mixed cores. Removing the fuel temperature Safety Limit for Standard fuel (1000 'C). Adding a new Limiting Condition for Operation to specify valid locations for the Instrumented Fuel Element. 1.5 Comparison with Similar Facilities Already Converted Three TRIGA reactors have, or will soon be,. converted from HEU to LEU fuel. Texas A&M University (TAMU) and Washington State University (WSU) are both MTR conversion reactors like the UWNR. TAMU has completed their conversion and WSU is under review. More detail UWNR LEU Conversion Analysis 3 August 2008 r-
is given on those below. The Oregon State University reactor is a standard GA geometry and not considered a "similar facility" in this section. 1.5.1 Comparison to the Texas A&M University (TAMU) Reactor The TAMU 1 MW TRIGA reactor has already been converted to TRIGA LEU (30/20) fuel. The primary differences from the UWNR core are the following:
" For UWNR, no control elements will be replaced whereas for TAMU the entire control system was replaced. " The UWNR control consists of three safety blades, one regulating blade, and one transient control rod; whereas, the TAMU control consists of four fuel followed control rods, one water-followed regulating rod, and one air-followed transient rod.
- The UWNR fuel assemblies have uniform pitch whereas the TAMU fuel assembly pitch varies in two directions.
1.5.2 Comparison to the Washington State University Reactor (WSU) The WSU 1 MW TRIGA reactor has already been approved for conversion to TRIGA LEU (30/20) fuel. The primary difference' from the UWNR core are the following:
- WSU has a mixed core which contains both HEU TRIGA FLIP fuel elements and low enriched (<20% U-235) standard fuel elements. During the WSU conversion, the HEU TRIGA FLIP fuel elements will be replaced with 30/20 LEU fuel elements. The UfWNR core, however, contains only HEU TRIGA FLIP fuel elements, and will be completely replaced with 30/20 LEU fuel elements during the proposed conversion.
UWNR LEU Conversion Analysis 4 August 2008 7>
0 In the WSU conversion, the transient rod will be replaced, whereas, no control elements 0 will be replaced in the proposed UWNR conversion. A summary of WSU and TAMU conversions and the proposed UWNR conversion is provided in Table 1.5.1. Table 1.5.1 Comparison of TAMU, WSU, and UWNR conversions. Component TAMU WSU UWNR Conversion Core Full conversion Partial conversion 30/20 LEU Full conversion 30/20 30/20 LEU and burned 8.5/20 LEU LEU Fuel Element Same Same Same Instrumented Fuel Same Same Same Element Fuel Rod Same Same Same
,Assemblies Fuel Assembly Variable in two Uniform in all directions Uniform in all directions Pitch directions Control Rods 4 TRIGA FFCRs 3 Control (shim) blades 3 Control (shim) blades Regulating Rod TRIGA RR- water Stainless steel (servo) blade Stainless steel (servo) followed blade Transient Rod TRIGA TR- air TRIGA TR- water followed TRIGA TR- water followed followed
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55 August 2008 U"R Conversion Analysis UWNR LEU Conversion Analysis August 2008
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S 2 SITE CHARACTERISTICS The site characteristics of the UWNR are described in the UWNR HEU SAR. The HEU to LEU conversion does not impact the site characteristics. 9" August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 77 August 2008
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I O3 DESIGN OF STRUCTURES, SYSTEMS, AND COMPONENTS The design of structures, systems, and components of the UWNR are described in the UWNR HEU SAR. The HEU to LEU conversion does not require any changes to the existing structures, systems, and components of the UWNR, however, a new storage container will be acquired from the Oregon State TRIGA Reactor (OSTR) for the initial storage of the LEU fuel. This storage container is described in the following section. 3.1 Initial LEU Storage Container As the new LEU fuel is received at the UWNR facility, it will-be stored in a lockable steel storage container located on the beam port floor. This location is near an installed area radiation monitor which will serve the purpose of a criticality monitor. An illustration of this container is
- shown in Figure 3. .1. The criticality of this configuration was analyzed with MCNP using an infinite array of square unit cells, each with a 10 cm pitch and containing one fuel element. The area above the container was modeled as air, and the area below the container was modeled as concrete. These areas were, specified to be effectively infinite in height. Two aluminum support plates are used to maintain spacing between fuel elements. They are located six inches above the bottom of the fuel element and eight inches below the top of the fuel element. Both top-view and side-view layouts of the MCNP model are shown in Figure 3.1.2.
99 August 2008 UWNR LEU U"R Conversion Analysis LEU Conversion Analysis August 2008
0 Figure 3.1].1 Solid works model of the initial LEU storage container. Air Above 40 E Fresh LEU 30/20 fuel pin.... Water or air surroundings. Concrete helow ~ Figure 3. 1.2 Layout qf the MCNP storage rack model. (Left) Top view (Right) Side view. Criticality calculations of the storage container were performed for two scenarios: 1) fuel surrounded by air, and 2) fuel surrounded by water. 'The first scenario represents the normal configuration, and the second scenario represents a situation in which the fire alarm has been Conversion Analysis UWNR LEU Conversion Analysis 10 10 August 2008 August 2008 0
activated and the storage container is flooded with water from the sprinkler cooling system. The infinite multiplication factors of these two scenarios are shown in Table 3.1.1. The reason the k. with water surroundings is less than the ko. with air surroundings is because water is a stronger absorber than air, and within an infinite lattice, moderation and scattering become less important. In both scenarios the multiplication factors are well below the technical specification criticality limit of 0.8, and therefore this container is adequate for the storage of new LEU 30/20 fuel. Table,3. 1.I Multiplicationfactors of the LEU initial storage container. Scenario K. Icy I) Fuel surrounded by air 0.68938 0.00034
- 2) Fuel surrounded by water 0.65679 0.00031 0-11 > August 2008 UWNR LEU Conversion Analysis LW Conversion Analysis 11 "* August 2008
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O 4 REACTOR ANALYSIS 4.1 Reactor Facility The University of Wisconsin Nuclear Reactor has a heterogeneous pool-type design and is currently fuelled with TRIGA FLIP fuel. The reactor has a maximum licensed steady-state power level of I MW, and it is cooled by natural circulation of light water through the core. The fuel elements are bundled into 4-element assemblies which are located within a 7 by 9 grid plate. The control elements consist of three shim-safety blades, a stainless steel regulating blade, and a transient control rod. A complete description of the facility is provided in the current UWNR SAR. The only change that will be made to the reactor for the HEU to LEU conversion is the fuel type. The only other design change to be made to the UWNR facility is the addition of a . temporary fresh fuel storage facility for the LEU fuel. 4.2 Reactor Core The current UWNR HEU core consists of 23 fuel assemblies located within a 7 by 9 grid plate as shown in Figure 4.2. 1 With one exception, all of the fuel assemblies contain four fuel elements. One of the central assemblies of the core only has three fuel elements because it containsthe transient control rod in replacement of one fuel element. This results in a total of 91 fuel elements within the core. In addition, two of these fuel elements are manufactured with internal thermocouples, and are referred to as instrumented fuel elements (IFEs). A list of the major components of the UWNR HEU and LEU cores is provided in Table 4.2.1. Both the grid plate and the control elements are supported from the bridge structure at the top of the biological UWNR LEU Conversion Analysis ¢- 13 Auaust 2008
shield. The following sub-sections will discuss the reactor fuel, control elements, graphite reflectors, startup source, and in-core experimental facilities in greater detail. Figure 4.2.1 Layout of UWNR core. Table 4.2.1 Core Componentsfor HEU and LEU cores. Core Configuration HEU FLIP LEU 30/20 Non-Instrumented Fuel Elements 89 81 Instrumented Fuel Elements 2 2 Transient Rod (water followed) 1 I Aluminum Clad Reflector Elements 10 14 Shim-Safety Blades 3 3 Stainless Steel Regulating Blade I I 4.2.1 Reactor Fuel The HEU and LEU fuel elements have identical dimensions, and the only differences between the two fuels are due to material compositions. "lhe proposed LEU 30/20 fuel has been approved UWNR LEU Conversion Analysis 14 August 2008
2
- by the NRC for use in TRIGA reactors in NUREG-1282. As shown in Figure 4.2.2, each element contains a cylindrical segment of UZrH fuel bounded on the top and bottom by cylindrical graphite reflectors. The fuel elements are clad in stainless steel, and have stainless steel fittings on each end. In addition, the center of each element contains a zirconium rod in replacement of fuel meat, creating an annulus-shaped fuel meat cross-section. A summary of the fuel characteristics is provided in Table 4.2.2. Detailed material compositions are described in Section 4.2.6.
UWNR LEU. Conversion Analysis 15 August 2008
0 0 Figure 4.2.2 Fuel element dimensions. UWNR LEU Conversion Anal vsis j --- 16 August 2008
" --- 0 ........
0
Table 4.2.2 Summary of HEU and LEU fuel characteristics. Design Data TRIGA FLIP HEU* TRIGA 30/20 LEU* Number of fuel elements 91 83 U-235 Enrichment [atomic %] 70 19.75 Uranium weight percent 8.5 30 -. Fuel elements per assembly 4 4 Fuel alloy inner diameter [mm/(in)] 6.350 (0.25) 6.350 (0.25) Fuel alloy outer diameter [mm/(in)] 34.798 (1.37) 34.798 (1.37) Fuel alloy length [mm/(in)] 381.0 (15) 381.0 (15) Cladding thickness [mm/(in)] 0.508 (0.02) 0.508 (0.02) Cladding outer diameter [mm/(in)] 35.8394 (1.411) 35.8394 (1.411) Erbium weight percent 1.48 0.9 Cladding material SS 304 SS304
*English units are given in parentheses 4.2.2 Control Blades The UWNR reactivity control system consists of three Boral shim-safety blades, one stainless steel regulating blade, and one aluminum clad boron carbide transient rod. The poison section of the safety blades consists of a Boral sheet (35 wt% B4C, 65 wt% Al) 40.5 inches long, 10.5 inches wide, and 3/8 inches thick.' The safety blades are clad in aluminum and are located within an aluminum shroud. Drawings of the safety control blade and control blade shroud are shown in Figure 4.2.3.' The regulating blade is a stainless-steel sheet with curls along the vertical edges. The sheet is about 11 inches wide and 40 inches long, and it is supported and guided in a manner similar to the safety blades. The 1.25 inch diameter transient control rod is located in a special 3-element bundle, replacing one fuel element as shown in Figure 4.2.4. The poison section of the transient control rod consists of boron carbide and is approximately 15 inches long.' During the proposed HEU to LEU conversion, the transient rod guide tube will be replaced, but the actual transient rod will remain unchanged.
17 August 2008 Conversion Analysis UWNR LEU Conversion Analysis 17 August 2008
Of the five control elements, the three shim-safety blades and the transient control rod have S scram capability. The reactivity worths of all the control elements are discussed in Section 4.5. The fully-inserted position of all control elements is 1,5 inches below the bottom of the active fuel region.
" ____- 27.4-t I
'1~~*
I I I II 3/8" BORAL SHEET 1/8" ALUMINUM CLADDING" I I I Ii II 0 I i I I I I I I I II I Ii I II I Ii
- -~- -H--
I I Ii I Ii II II I II I I 0 I I
- I I 0
Figure 4.2.3 (Left) Safety control blade,- (Right) Safety control blade shroud. UWNR LEU Conversion Analysis 18 August 2008 0
Figure 4.2.4 Three-element assembly with guide tubefor transient rod. 4.2.3 Neutron Moderator and Reflector The UWNR core uses nuclear-grade AGOT graphite reflectors. The. reflector elements in the HEU core are shown in Figure 4.2.1. The existing reflector elements will remain unchanged during the proposed HEU to LEU conversion, however 4 additional reflectors of similar design and construction will be added in positions B5 and F5 in lieu of fuel bundles, and in positions D3 and D7 in lieu of currently installed irradiation baskets. 4.2.4 Neutron Startup Source and Holder The same neutron startup source that was used in the HEU core will be used in the proposed LEU core. The neutron startup source will remain unaffected by the proposed HEU to LEU conversion. The source location will also remain the same. 4.2.5 In-Core Experimental Facilities The proposed UWNR HEU to LEU conversion requires no changes to in-core experimental facilities. 19 August 2008 UNýNR Conversion Analysis LEU Conversion UWNR LEU Analysis 19 August 2008
4.2.6 Reactor Materials The FLIP HEU fuel differs from the 30/20 LEU fuel only in fuel element composition. The compositions of the HEU and LEU fuel are provided in Table 4.2.3.3 The compositions of the non-fuel core components are provided in Table 4.2.4.4 For the HEU fuel, the number densities of Er-166 and Er-167 were determined from the GA manufacturer specifications. 3 For the computational analyses, an additional amount of Er-166 (labeled "Er-16x") was added to the fuel to preserve the reaction rate of the remaining erbium isotopes. This was necessary because the only erbium cross sections available in MCNP5 are Erbium-166 and Erbium-167. Ca ( UWNR LEU Conversion Analysis 20 August 2008 d
Table 4.2.3 HEU and LEU fuel compositions. Physical Nuclide Density Material Density Nuclide Ig/cm 3I [atoms/barn-cml U234 6.59000E-06 U235 9.01660E&04 U236 4.23000E-06 U238 3.70650E-04 H-1 5.45931E-02 UZrH HEU FLIP Fuel Zr 3.52987E-02 Natural Hf 2.11792E-06 Natural C 1.49606E-03 Er- 166 1.06240E-04 Er- 167 7.29500E-05 Er- I6x 2.77999E-05 U234 7.15000E-06 U235 1.08821E-03 U236 6.27000E-06 U238 4.32194E-03 UZrH LEU 30/20 Fuel H-1 Zr Natural Hf Natural C 4.91576E-02 3.22796E-02 1.93677E-06 1.78701 E-03 I Er- 166 7.71700E-05 Er- 167 5.29900E-05 Er- 16x 2.01935E-05 21 August 2008 Conversion ~nalysis LEU Conversion UWNR LEU Analysis 21 August 2008
Table 4.2.4 Non-fuel material compositionsfor UWNR core components. Material Physical 1g/cmDensity 31 Nuclide Nuclide to/br-mDensity Ig/CM3,[atoms/barn-cm] Cr-50 8.067E-04 Cr-52 1.556E-02 SS 304 (clad) 7.98 Cr-53, 1.764E-03 Fe-56 5.882E-02 Ni-58 8.232E-03 Mn-55 1.760E-03 AI-27 5.869E-02 2.68 Ae-27 5.869E-02 Aluminum Fe-26 5.020E-04 B-10 2.095E-02 B 4C (transient rod) 2.42 B-I 1 8.431 E-02 C nat 2.632E-02 B-10 8.058E-03 B- I 3.223E-02 Boral (safety blades) 2.64 Cnat 1.007E-02 A1-27 3.831E-02 WH-1 6.688E-02 W0-16 3.344E-02 Graphite (reflector in fuel) 1.75 Natural graphite 8.78208E-02 Zirconium (rod) . 6.51 Natural Zr 4.29747E-02 0 4.3 Reactor Tank and Biological Shield The proposed conversion from HEU to LEU fuel for the UWNR will not require any changes in the reactor tank or biological shield. 4.4 Core Support Structure The proposed conversion from HEU to LEU fuel for the UWNR will not require any changes in the core"support structure. UWNR LEU Conversion Analysis 22 August 2008 0
- 4.5 Dynamic Design The Monte Carlo N-Particle version 5 (MCNP5)5 was extensively used to perform detailed neutronic analyses of UWNR operating with HEU fuel and 30/20 LEU fuel. MCNP5 was chosen because of its versatile capabilities to handle complex three-dimensional geometries and heterogeneous materials. In addition, a variety of nuclear data available in MCNP5 allows modeling of UWNR at both cold and hot critical conditions. The MCNP5 model of UWNR used in this report was proved to provide accurate neutroriics calculations as it was successfully validated.with operational data 6. This MCNP5 model is shown in Figure 4.5.1 and Figure 4.5.2.
Note that the FLIP fuel was loaded into the core in three occasions over a five-year span. First, nine fresh-FLiP bundles were loaded into the existing TRIGA Standard core. The mixed core with. nine FLIP bundles had an exposure of 49.5 MWD before the second reloading batch of six additional FLIP bundles. The mixed core with fifteen FLIP bundles had an exposure of 19.7 MWD before the final reloading batch. Therefore, the burnups in the oldest nine FLIP bundles are 69.2 MWD higher than the freshest FLIP bundles. It is assumed'that this total bumup was small enough that modeling the system with all fresh FLIP fuel is reasonable. Therefore, fission products Were not included in the model. 23 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 23 August 2008
-4>, -- -
Figure 4.5.1 Axial cross section of the MCNP5 model.fbr HEU and LEU30/20 core analysis. UWNR LEU Conversion Analysis 24 August 2008
Figure,4:5,2 Horizontal-crosssection of the MCNP5 modelJbr HEU and LEU core analysis. 0 UWNR LEU Conversion Analysis 25 August 2008
4.5.1 HEU Fuel Neutronic Analysis REBUS-MCNP 7 was used to approximate isotopic constituents of the fuel as the reactor operated. Since neutron spectra were not uniform along the axial and radial direction of the fuel pin, each fuel pin was divided into five equal-height axial zones and each zone was again divided \into three equal-volume radial zones. A total of fifteen zones per fuel pin were enough to provide a reasonable resolution while remaining within the computational limit of REBUS-MCNP. The burnup calculation was done under the assumption that the reactor was operated under a steady state power of 1 MWth. Note that the current burnup of UWNR is about 818 MWd. Based on Figure 4.5.3, middle-of-life (MOL) andend-of-life (EOL) were determined to be 1400 MWd and 2800 MWd, respectively. 4.00 3.50 -* - HEU Fuel - - -Recommended EOL (0.5% dk/k) /
- o3.50 2.50 1
2.00
- 0)1.50 ..- - - - - - - - - - - - - - - - - - - - -
0.00 ............................. 0 250 500. 750 1,000 1,250 1,500 1,750 2,000 2,250 2,500 2,750 3,000 MWd Figure 4.5.3,Excess reactivity of UW7VR aS a function ol burnupfrom REBUS,-MCNP. I JWNR ILE1J Conversion Analvsis
. ......... .................. j ....
26 August 2008 0
- The Technical Specification requirement for shutdown margin states that the reactor shall not be operated unless the shutdown margin provided by the control rods is greater than 0.2% Ak/k with (1) the highest worth non-secured experiment in its most reactive state, (2) the- highest worth control element and non-scrammable blade (the regulating blade) fully withdrawn, and (3) the reactor in the cold condi tion without xenon. The calculated shutdown margin was 0.903 %Ak/k
+/- 2.577E-4 %Ak/k with no experiments in the core. The maximum worth allowed for a non-secured experiment in the Technical Specification is 0.7 %Ak/k. Thus, the shutdown margin minimum of 0.2% Ak/k is satisfied, including all three conditions. The calculated shutdown margin with all control elements fully inserted and no experiments in the core was 6.07 %Ak/k +/- 1.09E-3 %Ak/k. A recorded shutdown margin on an initial core loading with no experiments on 08/03/1979 was 5.1 %Ak/k.
Material definitions within the MCNP5 model were defined so that the neutronic parameters during beginning-of-life (BOL) could be calculated. A fully-inserted position of the transient rod must first be defined as it was a basis for calculating differential worth and integral worth values for all control elements in UWNR. The fully-inserted position of the transient rod was iteratively searched for and found to be 8.99 inches, which is measured from 1.5 inches below the bottom of the active fuel region. The total worth of the transient rod in this position was 1.334 %Ak/k +/- 0.0453 %Ak/k, which complied with the technical specification of 1.40 %Ak/k. The calculated position was in good agreement with the historical record of 8.60 inches. The initial positions at different core lifetimes were also calculated using appropriate sets of isotopic definitions from REBUS-MCNP and summarized in Table 4.5. 1. SUWNR LEU Conversion Analysis 27 August 2008
'Table 4.5.1 Initial positions of transientrod at difJfeent HEU core lifetime.
Condition Initial Transient Rod Height (inches) BOL 8.99 MOL 9.23 EOL" 9.65 The critical bank heights for cold andhot operating conditions were calculated using a number of MCNP5 simulations with different bank heights. In each simulation, the transient rod either remained at 8.99 inches or moved to the testing bank height if it was higher than 8.99 inches. Coolant temperature was set at 27 TC. This coolant temperature is different than the coolant temperature used in section 4.7.1 at 54.44 °C, which is the highest pool temperature Technical Specification limit, due to unavailability of nuclear data. A difference of 27.44 TC in water is fiegligible neutronically. The .critical bank heights were used to establish base cases for subsequent neutronic calculations. Table 4.5.2 summarizes the critical bank height for each critical condition at BOL, MOL, and EOL. Table 4.5.2 The critical bank heightfor hot and cold operating condition ofHEU fuel. Condition Height inches) k:n Control Blades Transient Rod Cold (27 °C) 10.04 10.04 1.00004 +/-0.00021 Hot (327 °C) 11.50 11.50 0.99999 +/-0.00024 MOL Cold (27 'C) 10.03 10.03 1.00003 +/- 0.00015 Hot (327 -C) 11.38 11.38 1.00007 +/-0.00013 Cold (27 °C) 12.97 12.97 1.00002 +/-0.00013 Hot (327 °C) 15.26 15.26 . 0.99993 +/-0.00013 \The excess reactivity of the HEU core at BOL was calculated by having all of the control elements fully withdrawn from their initial critical positions. The excess reactivities at hot and cold operating conditions were found to be 2.745 %Ak/k +/- 7.550E-4 %Ak/k and 4.064 %Ak/k UWNR LEU Conversion Analysis 28 August 2008
O +1- 9.373E-4 %Ak/k, respectively. A recorded excess reactivity at initial core loading on 08/03/1979 was 3.16 %Ak/k at cold conditions. The differential worth curve and the integral worth of each control element were calculated by performing rod calibrations with the MCNP5 model. For each of the calibrating points along the control element, a critical configuration of the remaining control elements was searched for via a number of MCNP5 simulations. Once the critical configuration was obtained, a small reactivity insertion was introduced by slightly increasing the height of the control element. The differential worth at the calibrating point was given by the change in reactivity divided by the difference of the control element height. In this report, ten calibrating points were used for boral control blades (control blade 1, control blade 2 and control blade 3) and regulating blade (control blade . 4) and six calibrating points were used for the transient rod. The differential worths were then fit to the function dp =Ccos2rz zo)"L dz H where C, z0 , and H were varied to improve the fit. The integral worth was calculated by integrating the fitting function from 0 to 43.18 cm (17 inches). Calculated differential worth curves for all control elements are shown in Figure 4.5.4 through Figure 4.5.8. Error bars represent lc statistical errors of MCNP5 results. Fitting parameters and integral rod worths are summarized in Table 4.5.3. 29 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 29 August 2008
*-0.350 0.300 0.250 <:1 0.200 0
o0150 0.100 0.050 0.000 0 2 4 6 8 10 12 14 16 18 20 Blade Position [inches] Figure 4.5.4 Differential worth curve of Control Blade I for HEUfuel 0,350 0 0.300 0.250 0.200
.. )t 0.150 <1
- 0) 0.100 0.050 0.060
-0.050 Blade Position [inches]
Figure 4.5.5 Differential worth curve of Control Blade 2for HEUfuel UWNR LEU Conversion Analysis LEU Conversion Analysis 30 30 August 2008 August 2008 0
0.350
- MCNP
- FITTING FUNCTION 0 300 0.250 0.200 0
0 0* 0.150 0 0.100 0.050 0.000 0 2 4 6 B 10 12 14 16 18 ý20 Blade Position [inches] Figure 4.5.6 Differentialworth curve of Control Blade 3for lIEUfuel 0.070
- MCNP 0 0.060
-----------.- .---T_-------T_r T--
- -- ---- FrTTING FUNCTION.
0.050 0.040 R,, 0,030 0 0.020 ----------- 0 C 0.010 ---- 0 C 0.000 2 4 6 8 10 12 14 16 13 F I
-0.010 Blade'Position [inches]
Figure 4.5.7 Differential worth curve of the Regulating Blade for HEUfuel 31 August 2008 UWNR LEU UWNR Conversion Analysis LEU Conversion Analysis 31 August 2008
0.400 0 0.350 0.300 0.250 0.200 0 0.150 C 0 0.l0 0 0 0.050 0.000 8 10 12 14 16 18 20 Blade Position [inches] Figure 4.5.8 Differential worth curve of the Transient Rodfor HEU fuel Table 4.5.3 Summary offitting parameters and integral worth of each control elementfor HEUfuel. 0 Fitting Parameters Calculated Experimental C Integral Worth Integral Worth Element [%Ak/k in] Zo [in] H [in] [%Ak/k] [%Ak/ki Control Blade 1 0.250 9.148 18.862 2.349 1.948 Control Blade 2 0.261 8.829 17.291 2.256 1.845 Control Blade 3 0.283 9.370 19.911 2.788 2.528 Regulating Blade' 0.046 8.219 21.684 0.490 0.409 Transient Rod 0.367 5.045* 31.212 1.467 1.374 K Discrepancies between the experimental and calculated integral worths in Table 4.5.3 exist because the experimental values were recorded on 11/04/07 and do not reflect the actual worths at BOL which are not available due to the history of the phased conversion to HEU FLIP fuel. UWNR LEU Conversion Analysis Analysi's 32 32 August 2008 August 2008 0
The effective delayed neutron fraction, I3eff, was calculated using MCNP5 with aid from the following expression, kT where k is the multiplication factor of the system where only prompt neutrons are generated P/ after fission events and kT is taken from the system where both prompt and delayed neutrons are created appropriately after fission events. Table 4.5.4 Effective delayed neutronfraction at different core lifetime. Condition Effective Delayed Neutron Fraction BOL 0.0075 +/- 0.00017 MOL 0.0076 +/- 0.00019 EOL 0.0073 +/- 0.00018 . The calculated values were close to the value of 0.0070, which was derived from the Fuchs-Nordheim equation. 8 MCNP5 and the 1/v-absorber insertion method 9 were employed to calculate the prompt neutron lifetime (lj, where a small amount of the absorber was distributed in the reactor and 1P was calculated from
. 'p I. k,ý,f - kr S" - and P = limlI N(7 0 v. kp N->O p 33 August 2008 UWNR U"R LIEU Conversion Analysis LEU Conversion Analysis 33 August 2008
where: kre =-. the multiplication factor of the reference system, kp = the multiplication factor of the system containing the absorber, N = the number density of the absorber in atoms/barns - cm,
-,o= the absorber's absorption cross section in barns, v,= the reference speed.
In this report, 1°B was chosen as the I/v absorber. Its cro- and v, were 3837 b and 220,000 cm/s, respectively. Practically, Ipwas calculated by linearly extrapolating I1I from two different 10B concentrations to N=0. Therefore, lI is/ given by
'2 '1j
/p =1 P + N).
N2 -N 1(0-
/
Two 10B concentrations selected were 6E-8 and 1.2E-7 atoms/b-cm. Prompt neutron lifetimes are summarized in Table 4.5.5. The calculated value is very close to a conventional value of 22 ls that has been used for HEU FLIP fuel in UWNR. Table 4.5.5 Prompt neutron lifetime at different core lifetimes. Core condition Prompt Neutron Lifetime lsl BOL 26.95 +/- 0.8 MOL 20.27 +/- 1.8 EOL 23.00 +/- 1.7 The void coefficients were calculated using the MCNP5 model. The coolant density in the MCNP5 model was uniformly reduced by 2.5% to create void in the system. In addition, the coolant temperature coefficients were determined by increasing the coolant temperature by 100 K. Both void coefficients and coolant temperature coefficients are summarized in Table 4.5.6. 11"WNR ILFI C7onversion Analvsis 34 Aug ust 2008 0
The calculated void coefficients are favorably close to a measured value of -2E-3 as reported in 0 the HEU SAR. 1 Table 4.5.6 Void and Coolant Temperature Coefficientsfor HEU at different core lifetimes. Fuel Condition Void Coefficient Coolant Temperature I (Ak/k) /(%void) I Coefficient [(Ak/k)/(K)j Cold (27 OC) -1.126E-3 +/- 3.416E-5 9.313E-5 +1- 1.659E-7 BOL Hot (327°C) -1.059E-3 +/- 3.399E-5 9.350E-5 +/- 1.588E-7 Cold (27°C) -9.705E-4+/- 3.398E-5 1.005E-4+/- 2.11 E-6 Hot (327 -C) -9.618E-4 +/-3.397E-5 1.090E-4+/- 1.760E-6 Cold (27'C) -6.730E-4 +/-3.442E-5 1.207E-4+/- 1.827E-6 Hot (327 °C) -6.329E-4+/- 5.907E-5 1.391E-4+/- 1.826E-6 Negative prompt temperature coefficients at different fuel temperatures were calculated using appropriate temperature cross sections of fuel materials within the MCNP5 model. Available . temperature cross sections were 300 K, 400 K, 600 K, 800 K and 1200 K. The negative prompt temperature coefficient at 350 K was derived from a reactivity change between 300 K and 400 K temperature cross sections. Figure 4.5.9 and Table 4.5.7 summarize these results for BOL, MOL and EOL. Thefaverage negative prompt temperature coefficient was found to be 1.033E-2 +/- K 7.732E-5 %Ak/k K at BOL, which is in a good agreement with the historic HEU SAR value of 1.26E-2 %AkUk K. Table 4.5.7 Negative prompt temperature coefficientsfor HEUfuel. Negative Prompt Temperature Coefficient I %Ak/k K I Temperature (K) HEU BOL HEU MOL HEU EOL 350 4.573E-3 +/- 6.395E-5 3.745E-3 +/- 1.555E-4 3.639E-3 +/- 1.703E-4 500 7.167E-3 +/- 2.853E-5 6.OOOE-3 +/- 8.585E-5 5.636E-3 +/- 8.249E-5 700 1.038E-2 +-- 2.903E-5 7.633E-3 +/- 9.025E-5 6.410E-3 +/- 7.208E-5 1000 1.918E-2 +/- 1.521E-5 1.544E-2 +/- 4.302E-5 1.329E-2 +/- 3.722E-5 35 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 35 August 2008
BOL 0 0.025 0.020 0.015 y = 2.2376E-O5x -3.9401E-03 R2 = 9.7680E -01 Ea 0.0,10 0.005 . z 0.000 300 500 700 900 1100 1300 Fuel Temperature (K) MOL 0.025 0 0.020 EJ 0.015 0.010 y = 1.7609E-05x - 3.0222E-03 R2 = 9.4924E -01 0.005 (~U z 0.000 300 500 700 900 1100 1300 FuelTemperature (K) UWNR LEU Conversion Analysis U"R Analysis 36 36 August 2008 August 2008 0
E OL 0.014 0.012 " 0.010
- y = 1.4348E -05x - 1.9019E -03 ER 2=9.2214E -01 W 0.008 E 0.006 0.004 aw 0.002 z
0.000 300 500 700 900 1100 1300 Fuel Temperature (K) Figure 4.5.9 Negative prompt temperaturecoefficients as afunction offuel temperaturefor HEUfuel at different core lifetimes. 4.5.2 30/20 LEU Fuel Neutronic Analysis Neutronic analysis for 30/20 LEU Fuel was performed using the same methodology as the,HEU fuel analysis. The MCNP5 model from the HEU case was modified to accommodate a change in fuel compositions between the two cases. Fuel compositions of 30/20 LEU fuel are summarized in Table 4.2.3. First, it was necessary to determine whether this new core configuration could be operated safely under the approved Technical Specification shutdown margin. The fully-inserted position of the transient rod was iteratively searched for and found to be 9.40 inches. The total worth of the transient rod in this position was 1.355 %Ak/k +/- 0.0227 %Akfk, which complied with the technical. specification of 1.40 %Ak/k. The reactor was shown to' be supercritical (k 1.00419 +/- 0.00016) when the Technical Specification shutdown margin with no experiments 37 August 2008 UWNR LFU U"R Conversion Analysis LEU Conversion Analysis 37 August 2008
was calculated. Therefore, it was necessary to make a geometry modification in the reactor core. The proposed LEU core configuration has the two fuel bundles in locations B5 and F5 removed and replaced with graphite blocks. In addition, two reflector blocks are inserted in locations D3 and D7. A horizontal cross section of the MCNP5 model that represents the new core configuration is shown in Figure 4.5.11. This was an attempt to make the core less reactive by removing fuel bundles in important locations of the core. The Technical Specification shutdown margin with no experiments for the new core configuration was calculated to be 0.994 %Ak/k +/- 0.011 %Ak/k and the fully-inserted position of the transient rod was 9.82 inches with a total worth of 1.369 %Ak/k. Employing the same calculation techniques as the HEU case, REBUS-MCNP was used to calculate isotopic inventories as the reactor burns up. The excess r'eactivity as a function of burnup is shown in Figure 4.5.10. UWNR LEU Conversion Analysis 38 August 2008
4.00 350 - - - - - LEU 30/20 Fuel - -Recommended EOL (0.5% dk/k) .--
"2 3.00---
2.50 - 0 U~2.00 U w.U , 0.50 ------------- -- - - 0.00 . ..... I....I....I.. T...... 0 250 500 750 1,000 1,250 1,500 1,750 2,000 2,250 2,500 MWd Figure 4.5.10 Excess reactivity of UWNR LEU core as afunction of burnupfrom REBUS-MCNP. Middle-of-life (MOL) and end-of-life (EOL) were determined to be about 800 MWd and 1800 MWd, respectively. The fully-inserted position of the transient rod at different core lifetimes was also calculated using appropriate sets.of isotopic definitions from REBUS:MCNP and summarized in Table 4.5.8. The shutdown margin with no experiments and all control elements fully inserted was 5.677 %Ak/k +/- 0.0 12 %Akfk. 39 August 2008 UWNR LEU Conversion Analysis Analysis 39 August 2008
0 0 Fhgiire 4.5. / / New LEU Core C'onfiguration. UWNR. LEU Conversion Analysis 40 August 2008
Table 4.5.8 Initialpositions of transient rod at different 30/20 LEU core lifetime. Condition Initial Transient Rod Height (inches) BOL 9.82 MOL 10.21 EOL 9.63 Critical bank heights for hot and cold operating conditions and different core lifetimes were searched for and summarized in Table 4.5.9. Table 4.5.9 The criticalbank heightfor hot and cold operating condition of 30/20 LEU fuel Condition Height (inches) ker Control Blades Transient Rod Cold (27 'C) 10.13 10.13 1.00002 +/- 0.00009 Hot (327 °C) 11.73 11.73 1.00005 +/-0.00011 MOL Cold (27 'C) 10.526 10.526 1.00006 +/-0,00010 Hot (327 0C) 12.053 12.053 0.99998 +/- 0.00022 Cold (27 -C) 11.435 11.435 1.00011 +/-0.00011 Hot (327 0 C) 13.143 13.143 1.00004 +/-0.00019 The excess reactivity of the 30/20 LEU fuel at BOL was calculated by having all of the control elements fully withdrawn from their initial critical positions. The excess reactivities at hot and cold operating conditions were found to be 2.752 %Ak/k +/ 0.011 %Ak/k and 4.163 %Ak/k +/- 0.0 12 %Ak/k, respectively. The calculation strategies previously used in the HEU case were again employed to determine differential worth curves and total worths of all control elements. Calculated differential worth curves for all control elements are shown in Figure 4.5.12 through Figure 4.5.16. Error bars represent IG statistical errors of MCNP5 results. Fitting parameters and integral rod worths are summarized in Table 4.5.10. 41 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 41 August 2008
0.300 0 0.250 T 0,200 0.150 0.100 0.050 0.000 0 5 10 15 20 Blade Position [inches] Figure 4.5.12 Differential worth curve of Control Blade 1for 30/20 LEU fuel 0.300 0.250
-FITTING MCNP FUNCTION 0 0.200 ý-S 0.150 0
0.100 0.050 0.000 0 2 4 6 8 10 12 14 16 18 Blade Position [inches] Figure 4.5.13 Differential worth curve of Control Blade 2for 30/20 LEU fuel 42 August 2008 UWNR LEU Conversion Analysis LEU Conversion Ana)ysis 42 August 2008
0.350
- MCNP
*- FITTING FUNCTION 0,300 0.250 0.200 .C 0.150 0
0* 0.100 0.050 0.000 0 5 10 15 20 Blade Position [inches] Figure 4.5.14 Dijferentialworth curve of Control Blade 3for 30/20 LEU fuel 0.060 0.050 0.040 0.030 t 0 0.020 4~ 0.010 a 0.000
-0.010 Blade Position [inches]
Figure 4.5,15 Differential worth curve of the Regulating Bladefor 30/20 LEUfuel 43 August 2008 UWNR LEU U"R Conversion Analysis LEU Conversion Analysis 43 August 2008
0.300 0 MCNP 0 0 ,2 5 0 -- - - -- -- -- -- - -- -- -- -- - -- -- -- -- -- -- -FTTING F ION 0 C 0) a, 0 0,000 - 1 1 11 12 13 14 15 16 17 18
-0.050 Blade Position [inches]
Figure 4.5. 16 Differential worth curve of the transient rodfor 30/20 LEU fuel Table 4.5.10 Fittingparameters and integral worth of each control element for 30/20 LEU fuel 0 Fitting Parameters Calculated Experimental Control Element C Integral Worth Integral Worth [%Ak/kin]_z[ [inl H linl I%Ak/kI l%Ak/kl Control Blade 1 0.217 9.169 19.797 2.129 N/A Control Blade 2 0.215 8.641 19.568 2.091 N/A Control Blade 3 0.255 9.264 20.693 2.599 N/A Regulating Blade 0.034 7.262 1 23.648 0.382 N/A Transient Rod 0.453 4.280 32.284 1.369 N/A Effective delayed neutron fractions at various core lifetimes were calculated and summarized in Table 4.5.11. 44 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 44 August 2008
0 'Table 4.5.11 Effective delayed neutronfraction at different core lifetime. Condition Effective Delayed Neutron Effective Delayed Neutron Fraction - HEU Fraction - LEU BOL 0.0075 +/- 0.00017 0.00782 +1- 0.000134 MOL 0.0076 +/- 0.00019 0.00774 +/- 0.000148 EOL 0.0073 +/- 0.00018 0.00739 +/- 0.000170 Prompt neutron lifetimes for the 30/20 LEU core were calculated using the same technique as the HEU core. The results are summarized in Table 4.5.5. Table 4.5.12 Prompt neutron lifetime at different core lifetimes. Condition Prompt Neutron Lifetime Prompt Neutron Lifetime Ittsl - HEU !Itsl - LEU BOL 26.95 +/- 0.8 27.14 +/- 0.8 MOL 20.27 +/- 1.8 27.93 +/- 0.8 EOL 23.00 +/- 1.7 29.38 +/- 0.8 The void coefficients and the coolant temperature coefficients of reactivity for the 30/20 LEU core were calculated by reducing the coolant density throughout the core by 2.5 percent and increasing the coolant temperature by lOOK, respectively. The results at different core lifetimes are summarized in Table 4.5.13. Table 4.5.13 Void Coefficientfor 30/20 LEU at different core lifetimes. Fuel Condition Void Coefficient Coolant Temperature I (Ak/k) /(%void) I Coefficient I(Ak/k)/(K)j Cold (27 C) - 1.489E-3 +/- 5.101E-5 8.162E-5 +/- 1.268E-6 Hot (327 °C) -1.352E-3 +/-4.825E-5 9.046E-5 +/- 1.126E-6 Cold (27 °C) -1.353E-3 +/- 4.824E-5 8.477E-5 +/- 1.267E-6 Hot (327 'C) -1.278E-3 +/- 5.447E-5 9.228E-5 +/- 1.272E-6 Cold (27C) -1.364E-3 +/- 5.099E-5 9.103E-5 +/- 1.267E-6 Hot (327 (C) -1.250E-3 +/- 5.662E-5 1.006E-4 +/- 1.478E-6 45 August 2008 UWNR U"R LEU Conversion Analysis LEU Conversion Analysis 45 August 2008
Negative prompt temperature coefficients at different fuel temperatures were calculated using appropriate temperature cross sections of fuel materials within the MCNP5 model. Figure 4.5.17 and Table 4.5.14 summarize these results from BOL, MOL and EOL. As expected, the magnitudes of the negative prompt temperature coefficients in the 30/20 LEU core are less than those in the HEU core. Table 4.5.14 Negative prompt temperature coefficients for 30/20 LEU fuel. Negative Prompt Temperature Coefficient I %Ak/k K] Temperature (K) 30/20 LEU BOL 30/20 LEU MOL 30/20 LEU EOL 350 4.084E-3 +/- 4.178E-5 3.259E-3 +/- 8.369E-5 3.050E-3 +/- 9.104E-5 500 6.578E-3 +/- 1.786E-5 5.652E-3 +/- 3.876E-5 5.326E-3 +/- 4.920E-5 700 8.202E-3 +/- 2.265E-5 6.745E-3 +/- 4.336E-5 5.861E-3 +/- 4.630E-5 1000 1.335E-2 +1- l.148E-5 1.1OOE-2 +/- 2.210E-5 9.540E-3 +/- 2.186E-5 0 UWNR LEU UWNR Conversion Analysis LEU Conversion Analysis 46 46 August 2008 August 2008 0
BOL 0.016 - - 0.014 .. .. . S0.012 .. C- -O x-7W E <= 0.010 y 1.382E-05x -7.5862E -04 e 0R008= 9.8313E'-01 -. - 0.008 I E CLcap" , , 0.006 >- 0.004 U 0.002 0 z 0.000 300 500 700 900 1100 1300 Fuel Temperature (K) MOL 0.012 - S 0.010 C 0.008008 y.= 1.1380E_-05x - 5.90.38E -04 E R2 9.7565E -01 0.006 CL E T c,, *: 0.004 . o 0~
'-I 0.002 z
0.000 .... . 300 500 700 900 1100 1300 Fuel Temperature (K) 47 August 2008 UWNR LEU Conversion Analysis Conversion Analysis 47 August 2008
EOL - 0 .0 12 ... . . .. S 0.010 CLJ 0 .008 .. . . . . . . . 0.0062= Ea i 0.002 300 500 700 900 1100 1300 Fuel Temperature (K) Figure 4.5,17 Negative prompt temperature coefficients as a/unction offuel temperature.for30/20 LEU fuel at different core lifetimes. 4.5.3 Comparisons of flux magnitudes between 'HEU and 30/20 LEU core Neutron flux magnitudes at UWNR during hot critical condition were calculated for the HEU and 30/20 LEU MCNP5 models at eight experimental facilities: three whale tubes, four beam ports and the thermal column. Comparisons of the neutron flux between the HEU and 30/20 LEU cores were made by calculating the ratio of the neutron flux between the two cases. The ratio of neutron flux across the beam port surface for beam ports 1, 2, 3 and 4 between the 30/20 LEU and HEU cores were found to be 1.0239 +/- 0.002.6, 1.0620 +/- 0.0040, 1.1775 +/- 0.0285 and 1.1796 +/- 0.0285, respectively. The flux ratios of whale tube locations C2, C8 and E8, are 1.268 +/- 0.0018, 1.258 +/- 0.0018 and 1.252 +/- 0.0017, respectively. UIWNR LEU Conversion Analysis 48 August 2008 0
I. ~ ***, To measure and compare the neutron flux in the thermal column, a front surface of the thermal column facing the reactor core was divided into fifteen vertical zones and fifteen horizontal zones. The ratios of neutron flux between 30/20 LEU and HEU cores of these zones are shown in Figure 4.5.18. The average ratio is 1.0518 +/- 0.00042. t s0 40 30 to 1.04, I'Q-1:0Z 0 1M
;;I I
- --0 -20 -o1. 0 10 20 30 W1dt1 (cm) l'hqure 4.5.18 Ratios of/lux measurements between LEU and HEU of/225 zones on the thermal column surface, 4.5.4 Calculated Worth of Fuel Bundle Locations The reactivity worth of each LEU 30/20 fuel bundle was calculated by removing the bundle one at a time from the MCNP5 model: and comparing to the case with all fuel bundles installed. This calculation of fuel bundle worths was performed both with and without the 14 graphite reflectors installed, and the results are summarized in the following two figures.
49 August 2008 UWNR UWNR LEU Conversion Analysis LEU Conversion Analysis 49 August 2008
Figure 4.5.19 LEU BOL Fuel Bundle Worth in %AJk/k, With Reflectors 0 Figure 4.5.20 LEU BOL Fuel Bundle Worth in % Ak/k, Without Reflectors I IWNR I .l I Conversion Analvsi0 v ..................... j .... A50upust 2008 0
- 4.6 Functional Design of the Reactivity Control System There is no change in the reactivity control system due to the LEU conversion.
4.7 Thermal-Hydraulic Characteristics 4.7.1 Analysis of Steady State Operation RELAP5/MOD3.3 Patch 210 was used to perform a thermal hydraulic analysis on the UWNR HEU core using TRIGA fuel with natural convection water flow around 91 fuel elements. Analysis of the HEU core at BOL was performed with reactor power at 1.5 MW with an inlet coolant temperature of 54.44°C (130'F). According to Technical Specifications, the safety limit for reactor power level is 1.5 MW and the bulk coolant temperature shall not be greater than 130°F or 54.440C'. The reactor is nominally -operated with temperatures between 25 - 30'C (77-867F) and at 1.0 MW. Therefore, using a reactor power of 1.5 MW and an inlet water temperature of 54.44°C (130'F) provides a bounding analysis. The water temperatures used will lower the margin to critical heat flux (CHF) but will not influence the maximum fuel temperature significantly. In addition, the highest power fuel rod was used to analyze steady state and transient conditions. Using RELAP5/MOD 3.3, the natural convection, flow rate, maximum fuel centerline temperature, clad temperature profile, axial temperature profile, and radial temperature profile were determined, and flow rates as a function of pin power were calculated. The power in which the hottest rod reaches Critical Heat Flux (CHF) was calculated with the aid of RELAP5/MOD 51 August 2008 UWNR UWNR LEU Conversion Analysis LEU Conversion Analysis 51 August 2008
3.3. Departure from Nucleate Boiling Ratio (DNBR)'was determined with the Bernath correlation and the 2006 Groeneveld CHF tables. 4.7.2 RELAP5 Code Analysis and Results The thermal hydraulic analysis used the RELAP5/MOD 3.3 computer code to calculate the following steady state parameters:
- Channel flow rate
- Axial fuel centerline temperature distribution.
- Axial cladding temperature profile
- Axial bulk coolant temperature distribution In the RELAP5/MOD 3.3 steady state model, only I channel was used and no cross flow between adjacent channels was modeled. If cross flow between the hot rod and the surrounding rods were included in the model, the predicted coolant temperature in the hot channel would be reduced. Lowering the coolant temperature would decrease the buoyancy force and thus lower the coolant mass flow rate. A lower coolant temperature would increase the margin to CHF while lowering the coolant mass flow rate would decrease the margin to CHF. To substantiate this statement, the physical phenomenon governing cross flow between multiple channels was examined.
In order for cross flow to develop between two channels, a pressure difference between a hot and cooler channel axially along the channels must exist. However, the pressures at the core inlet and the core outlet are equal for these channels. Therefore, there is little difference in pressure UWNR LEU Conversion Analysis J 52 Aunust 2008
- .between the two channels as one traverses from the bottom of the core to the top of the core.
Hence, a small, if any, cross flow would be expected. A look at the overall buoyancy/friction pressure changes in channels adjacent to the hot channel indicates the cross flow would be from the cold to the hot channel. The hot channel flow rate would increase due to the additional cross flow, however cross flow in this direction would decrease the hot channel buoyancy and diminish the effect of the increased hot channel flow rate. Thus, the overall effect would be a slightly higher margin to CHF due to cross flow. The steady state analysis was performed using the single hot channel divided into axial segments. This hot channel is located at D5 SW with a square flow channel. The hydraulic data used in the RELAP5/MOD3.3 model is given in Table 4.7.1. Since the Washington State University Reactor has identical fuel as the UWNR, the inlet and outlet pressure loss coefficients derived by General Atomics were used4' 1. According to technical specifications, there must be a minimum of 19 feet (5.79 m) of water above the core, which gives a pressure of 157.567 kPa (22.853 psia), with the assumption there is 101.3 kPa (14.69 psia) of normal atmospheric pressure at the top of the reactor pool. 53 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 53 August 2008
Table 4.7.1 Hydraulic input conditionsfor the hot channel (HEU) S Condition Si English Inlet Pressure Loss Coefficient4,11 2.02 Outlet Pressure Loss Coefficient 4"11 1.38 Absolute Pressure at Top of Core [kPa, psia] 157.567 22.853 Power of Hottest Rod [kW] 26.401 Reactor Power [MW] 1.5 Inlet Bulk Coolant [°C, F] 54.44 130 The RELAP5/MOD3.3 thermal hydraulic analysis was performed on the maximum powered channel. This channel's geometry is shown in Table 4.7.2. It was assumed in this analysis that all rods in the core had approximately the same axial heat generation shape and therefore the hot rod would produce the maximum local heat flux. Further, the hot rod was assumed to be bordered on all sides by a fuel rod having the same characteristics of the hot rod. Since the hot rod is located at D5 SW, the transient rod next to fuel element D5 SW is also assumed to be a fuel rod, adding additional conservatism tothe model. In the actual core, the hot channel will incorporate the hot rod and rods of lower power. Even if the maximum local heat flux is not in the hot rod, the conditions in the hot channel are still expected to bound the conditions at all other points in the core. UWNR LEU Conversion Analysis 54 August 2008
Table 4.7.2 Thermal hydraulicparameters usedfor the hot channel Parameter SI English Flow Area [cm 2, in 2] 4.7429 0.7352 Pitch [cm, in] 3.8862 1.530 Pitch / Diameter Ratio 1.0843 1.0843 Wetted Perimeter [cm, in] 11.4069 4.4909 Hydraulic Diameter [cm, in] 1.6632 0.65479 Diameter of the Heated Surface [cm, in] 3.58394 1.411 Fuel Element Heated Length [cm, in] 38.1 15.0 Fuel Element Surface Area [cm2, in 2] 428.9786 66.4918 Fuel Element Roughness [cm, in] .OE-4 3.9E-05 01.411 OD
-01.371 ID
~ *01.485 L]D I GUIDE TUBE 01,350 -ID 01,250 OD GUIDE TUBE TRASIENT ROD Figure 4.7.1 Diagramof subchannel used to model the UWNR hot rod 12 55 August 2008 UVrNR Conversion Analysis LEU Conversion UWNR LEU Analysis 55 August 2008
The pitch for the subchannel is 3.8862 cm (1.530 in). The fuel rod outer diametet for all fuel rods is 3.58394cm (1.411 in) and the guide tube diameter is 3.7719 cm (1.485 in)12. The flow 2 area for the subchannel is given by the following equation which is calculated to be 4.7429cm (0.7352 in 2): Flow Area = PitchC2 (Doaerdl3 )2 7r (DGuide Tube )2 4 4 4 4 As can be seen in Figure 4.7.1 the wetted perimeter is 3/4 of the diameter of a fuel rod plus 'A of the diameter of a guide tube and is calculated by the following equation to be 11.4069cm (4.4909in). Wetted Perimeter 7r4 DouirClcad + -4 D,,qe Tube The hydraulic diameter is calculated Using the following equation to be 1.6632cm (0.65479 in). Flow Area Hydraulic Diameter = 4 Wetted Perimeter The heated diameter used in the RELAP5/MOD3.3 model is the outer cladding diameter of 3.58394cm (1.411 in), and the fuel element heated length is 38.1 cm (15.0 in). Therefore, the total surface area of the fuel element is 428.9786cm 2 (66.4918 in2). In order to verify D5 SW is the hottest channel with the most limiting hydraulic diameter, E5 SW was also looked at, because-the flow area and wetted perimeter is different than in D5 SW due to the control blade. In addition, E5 NE wasalso analyzed to.look at an "average" core position in which there were no transient rods or control blades neighboring the channel. Table 4.7.3 shows how the hydraulic parameters change versus core position. Since E5 SW has a smaller hydraulic UWNR LEU Conversion Analysis 56 August v 2008
diameter, it is necessary to determine whether E5 SW or D5 SW has the more limiting CHF value. Table 4.7.3 Hydraulicparameters and CHF results at limiting locations in the core D5 SW E5SW E5 NE Dimension Adjacent to Adjacent to "Average" Transient Rod Control Blade Core Position Flow Area [cm 2 ] 4.7429 5.3105 5.0144 [in 2 ] 0.7352 0.8231 0.7772 Wetted Perimeter [cm] 11.407 15.1455 11.2593 [in] 4.4909 5.9628 4.4328 Hydraulic Diameter [cm] 1.6632 1.4025 1.7814 [in] 0.6548 0.5522 0.7014 Power per Rod [kW] 26.401 24.231 24.890 at 1.5 MW Flow Rate [kg/s] 0.1367 0.1305 -0.1333 [Ibm/s] 0.3010 0.2875 0.2936 MDNBR Groeneveld 2006 1.983 2.190 2.030 Bernath 1.287 1.293 1.374 Although the hydraulic diameter is more limiting for E5 SW, the higher power per rod is higher in D5 SW. The MDNBR for both the Groeneveld 2006 and Bemath correlations, to be described later, were lower in D5 SW than E5 SW or E5 NE as seen in Table 4.7.3. After performing this analysis, it was determined that all steady state analysis was to be performed in D5 SW adjacent to the transient rod for both the fuel temperature and Cl-IF results. Figure 4.7.2 compares a UWNR TRIGA FLIP fuel rod with the RELAP5/MOD3.3 nodilized subchannel volume. Node dimensions are given in Table 4.7.4. The length of the fuel is 38.1cm UWNR LEU Conversion Analysis 57 August 2008
(15 in), the length of both the upper and lower graphite reflectors is 8.763 cm (3.45 in), and the distance from the top of the bottom end fittingto the bottom of the top end fitting is 68.263 cm (26.875 in).' Upper End 20 Fitting
-A 19 Upper 18 Reflector 17 16 15 14 13 12 II 10 Fuel Rod 9
8 7 6 r3 4 0 Lower 2 Reflector Lower End Fitting Figure 4.7.2 Axial Nodilization of Fuel Rod The lengths of the lower and upper end fittings were set to be 9.1131cm (3.5878 in) and 11.4609cm (4.5 12 in) respectively. A brief sensitivity study looked at reducing the lower end fitting from the original 9.11 cm to 5 cm (3.59 in to 1.97 in) and increasing the upper end fitting from 11.46 cm to 15.57 cm (4.51 in to 6.13 in), and found the maximum fuel temperature was 58 August 2008 Conversion Analysis UWNR LEU Conversion Analysis 58 August 2008
unchanged and the flow rate changed by less than 1%. Therefore, the length of the lower and upper end fittings have a limited role in this analysis. Since the fuel nodal lengths must be equal length, Nodes 3 - 17 lengths are given by Equation 4.7.1. L3f1el Equation 4.7.1 lfuel L03_.7 refers to nodal length for Nodes 3 through 17, nful is the number of nodes defined for the fuel region, 15. This gives the length of each fuel node of 2.54 cm (1 inch). Table 4.7.4 is formulated from these lengths. 59 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 59 August 2008
Table 4.7:4 Axial nodilization of the single channel RELAP5/MOD3.3 model 0 Node Node Length Axial Location Description Icm] lini [cm] [in] 20 11.4609 4.512 76.2000 30 Upper End Fitting 19 4.3815 1.725 64.7391 25.4878 1.725 60.3576 23.7628 Upper Graphite 18 4.3815 17 2.54 1.0 55.9761 22.0378 16 2.54 1.0 53.4361 21.0378 15 2.54 1.0 50.8961 20.0378 14 2.54 1.0 48.3561 19.0378 13 2.54 1.0 45.8161 18.0378 12 2.54 1.0 43.2761 17.0378 11 2.54 1.0 40.7361 16.0378 10 2.54 1.0 38.1961 15.0378 Fuel 9 2.54 1.0 35.6561 14.0378 8 7 2.54 2.54 1.0 1.0 33.1161 30.5761 13.0378 12.0378 0 6 2.54 1.0 28.0361 11.0378 5 2.54 1.0 25.4961 10.0378 4 2.54 1.0 22.9561 9.0378 3 2.54 1.0 20.4161 8.0378 2 8.763 3.45 17.8761 7.0378 Lower Graphite 1 9.1131 3.5878 9.1131 3.5878 Lower End Fitting After constructing the axial nodilization for the RELAP5/MOD3.3 -model, the radial nodilization also needs to be determined. A cross sectional view of a fuel element is shown in Figure 4.7.3, and the radial nodilization input is shown in Figure 4.7.4. The fuel rod consists of a fuel pin with an annular U/ZrH/Er casting. The fuel slugs are hydrided and then fit tightly into stainless steel UWNR LEU Conversion Analysis UWWR Analysis 60 60 August 2008 August 2008 0
tubes. Then, the central void in the annulus is backfilled with a zirconium plug. A nominal gap exists between the fuel and the stainless steel clad. The gap initially is filled with air (primarily nitrogen) but as burnup of fuel evolves, hydrogen and fission gasses migrate into the gap. The gap will be detailed more thoroughly later on in the report. I, Stainless Steel 0 Zirconium Fuel Gap F'igure 4. 7.3 Radial nodilizatianinjuel element (not to scale) The mesh points within the, fuel region -used in the RELAP5/MOD3.3 model correspond to 2 nodes for the zirconium pin, 21 :nodes of equal radial volume for the fuel meat, I node for the gap, and 2 nodes for the stainless steel clad. The radial node length and location of each node is shown in Table 4.7.5. When performing the gap analysis, the inner clad node, would move outwards (Node 25). TRIGA fuel typically has a gap distance between the stainless steel and the fuel meat of UWNR LEU Conversion Analysis 61 August 2008
1.27E-4 to 1.016E-3 cm (0.05 to 0.4 mil)15. The conversion from mil to centimeters is given by the following equation: Lcn= 0.000254*Ln1ii I 1 2 3 4 5 6 7 8 910 11 12 13 14 15 16 17 18 19 20 21 22 23' 24 25 26 27 T Gap Zirconium Rod Fuel Stainless Steel Claddir Figure 4. 7 4 Radial nodilization in fuel element (not to scale) 62 August 2008 UWNR LEU Conversion Analysis Conversion Analysis 62 August 2008
Table 4.7.5 Radial nodilization of the single channel RELA P5/MOD3.3 model NodeDecito Node Length Radial Location Icmj [in] Icm] [in] 1 0.15875 0.06250 0.1588 0.06250 Zirconium Rod 2 0.15875 0.06250 0.3175 0.12500 3 0.172562047 0.06794 0.4901 0.19294 4 0.125986141 0.04960 0.616 0.24254 5 0.104278078 0.04105 0.7203 0.28359 6 0.090984088 0.03582 0.8113 0.31941 7 0.081762459 0.03219 0.8931 0.35160 8 0.074880497 0.02948 0.968 0.38108 9 0.069489782 0,02736 1.0374 0.40844 10 0.065118802 0.02564 1.1026 0.43408 11 0.061481615 0.02421 1.164 0.45828 12 0.05839334 0.02299 1.2224 0.48127 13 0.055728402 0.02194 1.2782 0.50321 Fuel 14 0.053398105 0.02102 1.3316 0.52424 15 0.051337772 0.02021 1.3829 0.54445 16 0.049498987 0.01949 1.4324 0.56394 17 0.047844679 0.01884 1.4802 0,58277 18 0.046345921 0.01825 1.5266 0.60102 19 0.044979764 0.01771 1.5716 0.61873 20 0.043727738 0.01722 1.6153 0.63594 21 0.042574785 0.01676 1.6579 0.65271 22 0.041508488 0.01634 1.6994 0.66905 23 0.04051851 0.01595 1.7399 0.68500 24 2.54E-4 0.00010 1.740154 0.68510 Gap 25 0.025908 0.01020 1.766062 0.69530 26 0.025908 0.01020 1.79197 0.70550 After implementing the axial and radial nodilization of the fuel, power peaking factors for the fuel are analyzed using MCNP5. Since it is essential to determine the highest rod power in the UWNR LEU Conversion Analysis 63 August 2008
core, the hot channel peaking factor, fuel axial peaking factor, and fuel radial peaking must be looked at. Each of these factors is found using MCNP5 using the following equations: Hot Channel Peak Factor = Maximum. Fuel Rod Power Core Average Fuel Rod Power Maximum Axial Power in the Hot Rod Hot Channel Fuel Axial Peak Factor
=
Average Axial Power in the Hot Rod Hot Channel Fuel Radial Peak Factor= Maximum Radial Power in the Hot Rod Average Radial Power in the Hot Rod Since the radial nodilization of the fuel defined with the RELAP5/MOD3.3 model has equal volume segments, no weighted averages need to be performed. This allows for easy comparison between the MCNP5 and RELAP5/MOD3.3 models. For steady state analysis, the most important parameters to look at for limiting pin location are: pin peaking factor, axial peaking factor, and reduced flow locations in the core. For pulsing analysis, pin peaking, axial, and radial peaking factors will be the most important when trying to find the peak temperature during a pulse. The methodologies set forth in the thermal hydraulic section are first implemented in the HEU core in order to establish confidence in the model. Applying these same methodologies to the LEU core will provide the basis for analyzing the LEU core. Since the HEU analysis is meant to establish confidence in the model, only BOL will be thoroughly analyzed in order to compare with measured data. The LEU core will be analyzed at all stages of core lifetime: BOL, MOL, and EOL to see how the thermal hydraulic analysis changes with burnup. Since the UWNR core is physically small in size with a high degree of neutron leakage and relatively large neutron UWNR LEU Conversion Analysis 64 August 2008
- mean free path, physical changes in the core such as control rod movement will have a relatively small effect on the flux distribution. With the rods full out the axial power profile tends to flatten out than with rods at the critical bank height. Therefore only the effect of bumup has been explicitly modeled for thermal hydraulic analysis.
Furthermore, to determine the Minimum Departure from Nucleate Boiling Ratio (MDNBR), it is necessary to use appropriate CHF correlations. The CHF correlations that will be used are the Groeneveld 2006 look up tables13 and Bernath correlation.1 4 The reason for choosing these correlations is because the Bemath correlation has been used as a supplement in research reactor SARs in the RERTR program including Washington State University Research Reactor 4 , Oregon State University Research Reactor' 5 , the University of Massachusetts Lowell Research Reactor 16, and the South African MNSR17 . The Bernath correlation produces the most limiting MDNBR values of correlations considered. The Groeneveld 2006 look-up tables are considered to be the most accurate method for calculating CHF values over other correlations such as McAdams and Groeneveld 1986. The Groeneveld 2006 and Bernath correlations are detailed below: Groeneveld 2006 13: The Groeneveld look-up tables are based on a tube 8mm in diameter uniformly heated on the inner surface. By interpolating from a mass flux, equilibrium quality, and pressure a predicted CHF value can be found. Since the geometry of most CHF experiments are not based on the 0 UWNR LELJ Conversion Analvsis 65 Aueust 2008
geometry used to create the original look-up tables, six correction factors have been added to this interpolatedCHF value. CHFBUNDLE = CHFTABLE X K, x K 2 x K3 x K 4 x K5 x K6 These correction factors are: hydraulic diameter (K1), rod bundle effects (K2) such as cross flux and additional surface friction, grid plate spacer dependency (K3), Axial heated position to include phase change dependence (K4), Axial flux distribution factor (K5 ) to account for non-uniform axial heat distributions, vertical/horizontal orientation factor (K6) to account for gravitational effects associated with flow regimes. Using the Groeneveld 2006 look-up tables given the mass flux, equilibrium quality, and absolute pressure calculated with the RELAP5/MOD3.3 model, an initial CHF (CHFTrAB.E) can be determined. Then the appropriate correction factors are then multiplied to the initial CHF to get the Groeneveld 2006 CHF value. Only the correction values K1 , K2, and K4 ate used forthe UWNR analysis. All other factors are assumed to be 1.0 because they are either not applicable or close to 1.0. K3 is the grid spacer factor and is not applicable since there are no grids. K5 is the axial flux distribution factor and is I when the quality is less than zero. K6 is the flow factor used to correct CHF for downward flows and some upward flow conditions when the flow rate is less than 100 kg/m 2-s. K, is defined as: K, =(8)for 3<D<25 Kl \-, 0.57 for D> 25 UWNR LEU Conversion Analysis J 66 August 2008
.Where, D is the hydraulic diameter measured in millimeters. K 2 is defined as: K, = min[0.8, 0.8 exp(- 0.5 X*)] Where, X is the equilibrium quality. K4 is defined by the following algorithm: if X<0, X=0 L L if- < 5,- = 5 D D CL= x X+ pg(l - X)/pr K4 =exp - exp(2a Where, L is the heated distance from the inlet to outlet of the channel and D is the heated diameter. The quantities pi and pg are liquid and gas density of water in kg/mi3 It is important to note that the factor a is strongly dependent upon quality. Negative quality corresponds to
- subcooled conditions and any changes between subcooled and saturated conditions will result in a significant change to K4 . Groeneveld 2006 CHF values for UWNR were calculated implementing these correction factors. For the hot channel, K, = 0.6935, and the remaining correction factors are dependent upon quality, which is a function of core position and core power.
Bernath Correlation 14 : The Bernath correlation defines CHF in units of lb, 1- 0 C / hr-ft2 (p.c.u./hr-ft2-°C) and is given by the following equations: 67 August 2008 UWNR LEU UWNR Conversion Analysis LEU Conversion Analysis 67 August 2008
CHF = h BO (TWo- Tb) hBo = 10890 D (D,D+ Di)
+ AvDi T° =57In(P)-54L {P 48 ifD , <0.l ft 6
A= De° if <Oi
-o+90 ifD >0.1 ft De where:
hBo Limiting film coefficient [p.c.u./hr-ft2 -°C] Tb Fluid bulk temperature [°C] TwBo Wall temperature at CHF [°C] v Fluid velocity [ftlsec] A "slope" Pabs Absolute pressure [psi] Di Diameter of heated surface [ft] De Hydraulic diameter [ft] For the hot rod, the slope (A) is 274.844, and the remaining terms are dependent upon quality and core position. 4.7.3 HEU Power Summary To determine where the hottest rod is in the core, it is important to look at pin power peaking factors of the 91 fuel pins in the core through burnup. The pin power peaking factors were determined with MCNP5 with hot conditions and the control blades at the critical bank height. These pin power peaking factors can be found in Figure 4.7.5. In conjunction with axialpeaking factors and hydraulic parameters, the hottest rod can be found in the corethrough core life. As UWNR LEU Conversion Analysis 68 vv August 2008 0
- seen previously, D5 SW has the most limiting hydraulic diameter due to the influence of the guide tube in reducing flow area and increasing the wetted perimeter, and it also has the highest power peaking factor of 1.60 at BOL. Therefore, the RELAP5/MOD3.3 I channel model will use D5 SW as the hot rod to find the most limiting conditions in the UWNR core. In addition, Figure 4.7.5 shows the power/rod of the core at 1.0 MW in kW/rod.
Figure4.7.5 Pin power [kW/rod] and power peakingfactors (PPF)of the UWNR core at 1.0 AW (lHEU BOL) Figure 4.7.6 and Figure 4.7.7 show the power distribution of the UWNR core at 1.0 MW for
-IEU MOL and HEU EOL respectively. The hottest rod is always at D5 SW, but the power peaking factor across the core flattens out with bumup. When coupled with the axial and radial power peaking factors, the most limiting fuel rod for steady state analysis is D5 SW at BOL for the HEU analysis.
69 August 2008 Conversion Analysis UWNR LEU Conversion Analysis 69 August 2008
0 Figure 4.7.6 Pin Power [fkW/rod] andpower peaking factors (PPF)of the UWNR core at 1.0 MW (HEU MOL) 0 Figure 4.7.7 Pin Power fkWirod/ and power peaking.factors (PP11F) of the UWNIR core at 1.0 MW (fEU EOL) 70 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 70 August 2008
4.7.4 HEU Beginning of Life Steady State Core Analysis After determining the peak pin power over bumup, the radial and axial heat generation profiles need to be inputted into the RELAP5/MOD3.3 model. These profiles are shown in Figure 4.7.8 and Figure 4.7.9. These profiles are determined using the MCNP5 model. It is important to note that RELAP5 requires a radial power density distribution entered, arid not a radial power distribution. This tends to flatten out the fuel rod radial temperature profile. In addition, as can be seen in Figure 4.7.8, the innermost region of the fuel is a zirconium rod, and thus has no heat generation in it. Each power profile has been normalized so that the average equals 1.0. The power profiles shown in this report are representative of the hottest rod in the core, at D5 SW, in a core of 91 FLIP fuel elements. The hot rod power peaking factor is 1.60, giving a hot rod power of 26.401 kW at a steady state power level of 1.5 MW. The peak radial power density factor is 1.455 at the outer edge of the fuel, and the peak axial power factor is 1.428 about 14 cm above the bottom of the fuel. 71 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 71 August 2008
Normalized Axial Power Distribution of Hot Rod Axial Distance from Bottom of Fuel [in] 0 0 2 4 6 8 10 12 14 1.5 - 1.3 C.
- o 0.7 Cu
'0.5 0.3 0 5 10 is 20 25 30 35 40 Axial Distance from Bottom of Fuel [cm]
Figure 4.7.8 Normalized axial power density distributionof hot rod (HEU B01) Normalized Radial Power Density Distribution of Hot Rod Radial Distance from Fuel Centerline [in) 0 0.1 0,2 0.3 0.4 0.5 0.6 c3.7 1.5 0 1.4 (U 1.3 1,2 1.1 9 3: 1 0 9 p. 0.9 0.8 . . V * *
- 0.7 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 Radial Distance from Fuel Centerline [cm]
Figure 4.7.9 Normalized radialpower density distributionof hot rod (HEU BOL) 72 August 2008 UWNR LEU Conversion Analysis Analysis 72 August 2008
The next portion of the model that needs to be analyzed is the gap between the fuel and the inner cladding since the radial temperature profile is very dependent upon the size of this gap. To determine what the proper gap distance should be in order to be sufficiently limiting to bound the BOL measured data, a thorough analysis was performed. The hot gap size is known to be between 0.00127 and 0.01016 mm (0.05 mil and 0.40 mil) 18. In order to increase the gap, but not, change the flow area, the gap width was taken at the expense of the inner cladding. By increasing the radial distance of node 24, or the inner cladding, the gap could be expanded. The gap conductance model implemented in the RELAP5/MOD3.3 model used the RELAP5-3D default noble gas mole fractions of 0.1066/0.1340/0.7594 He/Kr/Xe. The gas conductance was calculated to be 1.04E-02 W/m-K at 300 K to 4.85E-02 W/m-K at 1900 K using the RELAP5-3D1 9 code manual in the correct form of equations 4.5-3 through 4.5-6. To verify the accuracy of the model, it is necessary to compare the model to experimental data. Since no instrumented element can be placed in D5 due to the presence of the transient rod, another rod was' looked at for comparison. The HEU core has the IFEs located in D4 SW and E3 NE. However, in an effort to compare similar power peaking factors, pin E4 SE was chosen since it has a pin power peaking factor according to the UWNR SAR of 1.10', which is identical to the MCNP5 model. Using the RELAP5/MOD3.3 model, the pin was modeled with a core power of 1.0 MW or 12.1 kW in the E4 SE rod, which are the same conditions as the measured data given in the UWNR SAR on page 4-43 for the all FLIP core. In addition, the inlet water temperature of 30'C (86°F) and water height of 6.096 m (20 ft) above the core were used to simulate conditions when the measured data was taken. Although the fuel temperatures at the UWRLU ovrio nlsi 3Agut20 UWNR LEU Conversion Analysis 73 August 2008
thermocouple are insensitive to the coolant inlet temperature since there is nucleate boiling at the outer surface of the clad at the levels of the three thermocouples, using conditions that are similar to the measured data gives the model more accuracy. Using a constant radial gap of 0.00254 mm (0. 1 mil), the maximum radial temperature could be plotted as seen in Figure 4.7. 10. E4 SE Peak Radial Temperature Distribution with 0.10 mil Gap Radial Distance from Fuel Centerline [in] 0.0 0.1 0.2 0:3 0.4 0.5 0.6 0.7 400 712 1- l
~-612 3500.. -6
*CL E 200
.t I*
- 0.1 Gap Radial Temperature Distribution uT.3 j center l
S12 412 C E 150 A TO3 Bottom 312 X TO.3 Top 100 "r- - ,., -... . . . . .- - - - --- ,. . .212 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1:8 Radial Distance from Fuel Centerline [cm] Figure 4.7.10 Radial temperatureprqfile at E4 SE with 0.1 mil gap (12. / kWirod) and core power of 1.0 MW Figure 4.7.10 shows that the radial temperature profile of the REAP5/MOID3.3 model bounds the measured temperatures shown as T0.3 for a 0.1 rmil (0.00254 mm) gap. The instrumented elements, T0 .3, are measured 0.762cm (0.3 inch) from the centerline of the fuel in three axial locations. One is at the center of the fuel, and.the remaining two are 2.54cmn.(l inch) below the center of the fuel and 2.54cm (1 inch) above the center of the fuel. The difference in temperature between the RELAP5/MOD3.3 model is 26'C (78.8°F) for instrumented element 41 measured in UWNR LEU Conversion Analysis 74 August 2008 0
- the center of the fuel and 51C (123.8°F) for instrumented clement 41 measured 2.54cm (1 inch) below the center of the fuel and 2.54cm (1 inch) above the center of the fuel. Therefore, the steady state analysis is limiting with a 0. 1 mil (0.00254 mm) gap thickness.
A gap of 0. 1 mil (0.60254 mm) was chosen for the analysis and supported by temperature measurements. Typically, gaps near values of 0.1 mil occur for the hot rod at EOL after the fuel has swelled. Near the BOL, the hot rod gaps are generally greater, usually between 0.25 mil to 1.0 mil (0.00635 mm to 0.0254 mm). The use of the small gap for the BOL analysis is due to the fact that the use of the gas mixture He/Kr/Xe would generally occur for MOL and EOL, while air would be present in the gap at BOL. Air has a higher thermal conductivity than the He/Kr/Xe gas mixture. Thus, the use of a small, 0.1 mil (0.00254 mm), gap and the He/Kr/Xe gas mixture . for BOL analysis is assumed to be equivalent to using a larger gap and air as the gap gas. In addition, performing the steady state analysis at 1.5 MW would also indicate smaller gaps. Analysis for both the HEU and the LEU core will use a 0. 1 mil gap with the preceding He/Kr/Xe gas mixture for all stages of core lifetime: BOL, MOL, and EOL. To tabulate these results, a summary of the measured versus calculated temperatures as a function of power can be seen in Table 4.7.6. When performing pulse calculations, the maximum fuel temperatures occurring during the first half of a second of a power pulse are less sensitive to gap thickness than are steady-state temperatures, since a half of a second is too short for substantial heat transfer. In addition, high fuel temperatures experienced during a pulse will cause differential radial expansion to reduce the gap thickness.
\UWNR LEU Conversion Analysis 75 August 2008
Table 4.7.6 Calculatedvs. measuredfuel temperaturesat E4 SE with 0.1 mil gapfor HEU BOL core Measured 'C Calculated 'C Power Hot Channel (OF) (OF) kW] Element 41 TMax To. 3 TouterClad Tcoolant Bottom: 305 (581) 12.1 Center: 325 368.26 351.55 131.65 88.87 (617) (694.87) (664.79) (268.97) (191.97) Top: 305 (581) After entering the radial power density profile, axial profile, and pin power into the RELAP5/MOD3.3 model, the steady state results can be determined. Since it is difficult to determine what the actual gap of the hot rod is, the gap was plotted with +/-50% uncertainty (0.05 mil, 0.00127 mm). Then the temperature profiles of the hot rod at 1.5 MW (26.401 kW in D5 SW) were determined. A radial temperature profile of the hot rod was performed with gap sizes ranging from 0.05 mil to 0.15 mil (0.00127 to 0.00381 mm) as seen in Figure 4.7.11. The maximum fuel temperature, outside cladding temperature and bulk coolant temperature at each node are plotted in Figure 4.7.12. UWNR LEU Conversion Analysis ,/ 76 Auuust 2008
r- - Radial Temperature Distribution of Hot Rod at 1.5 MW Radial Distance from Fuel Centerline [in] 0.1 0.2 0.3 0.4 0.5 0.6 0.7 800 1412 700 m m - - A 1212 A
-600 U A A.
- a m A 1012
*.500 AA , 72 W 400 E 0.05 mil gap 62 V' 300 10.10 mil gap 412 200 0.15 mil gap 100 212 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 Radial Distance from Fuel Centerline [cm]
Figure 4.7,11 Radial temperature profile of hot rod at core power of 1.5 MW with varying gap widths (lIEU BOL) 0 Axial Temperature Distribution at 1.5 MW Axial Height from Bottom of Fuel [in] 0 2 4 6 8 10 12 14 700 1232
-+/- Centerline 600
.. outer Cladding 1032 S500
*400 1 A Bulk Water- 832 300
- 632 U U U a a a a . *
- U 100
- 23 200 432 100 *232 A A A A A A £ A
A A
£% *32 0 5 10 15 20 25 30 35 Axial Height from Bottom of Fuel [cm]
F'igure 4.7. 12 Axial temnperatureprofile of hot rod at core power of I. 5 MW with 0. 1 rail gap (lIEU ,OId UWNR LEU Conversion Analysis 77 August 2008
Figure 4.7.11 shows the radial temperature distribution of the hot pin at 26.4 kW, giving a peak centerline fuel temperature of 642.03'C (I 187.65°F) with a gap size of 0.1 mil. Figure 4.7.12 0 shows the axial temperature profile for the fuel centerline, outer cladding, and bulk coolant. The peak temperature is located on the 6 th node, or 13.97 cm (5.5 in) from the bottom of the fuel where the axial power profile is also the highest. In addition to this analysis performed at a core power of 1.5 MW, the same parameters can be calculated at a core power of 1.3 MW (trip set point + uncertainty) and 1.0 MW (normal operating conditions) as shown in Table 4.7.7. Table 4.7.7 Steady state results of hot rod at core power of 1.5, 1.3, and 1.0 MW (lEU BOL) Parameter Core[MW] Power S1I Enls English 1.5 26.401 Rod Power [kW] 1.3 22.886 1.0 17.604 1.5 0.13665 0.3010 Flow Rate for Hottest Channel [kg/s], [Ibm/s] 1.3 0.12084 0.2662 1.0 0.09738 0.2145 1.5 0.31626 1.0376 Maximum Flow Velocity [m/sj, [ft/sJ 1,3 0.27322 0.8964 1..0 0.21397 0.7020 1.5 878&885 Maximum Wall Heat Flux [kW/rn2 ] 1.3 760.906 1.0 585.302 1.5 642.03 1187.65 Maximum Fuel Centerline Temperature ['C], ['F] 1.3 575.65 1068.17 1.0 475.38 887.684 1.5 140.90 285.62 Maximum Outer Clad Temperature [°C], [°FF] 1.3 138.95 282.11 1.0 135.75 276.35 1.5 127.14 260.85 Exit Outer Clad Temperature [°C], [°F] 1.3 125.97 258.75 1.0 124.05 255.29 1.5 100.44 212.79 Exit Bulk Coolant Temperature [°C], ['F] 13 99.49 211.08 1.0 97.48 207.46 UWNR LEU Conversion Analysis 78 August 2008 0
Since the hot rod located at D5 SW is next to the transient rod, it is impossible to place the IFE in this rod. However, if one were able to place the IFE in D5 SW the three thermocouples in the IFE would measure the following as shown in Table 4.7.8. As stated previously, the thermocouples are located 0.3 inches (0.762 cm) from the fuel centerline. Table 4.7.8 PredictedIFE measurements in D5 SW if lFE could be placed in D5 SW (lIEU BOL) IF IELocation Core[MW] Power $S I°Cl English [OF] 11.3.5 593.46 533.47 1100.23 992.25 Bottom - 16.51fue - inches) cm (6.5 ..... from bottom of 1.3: 531,47 992.25 1.0 442.73 828.91 1.5 579.53 1075.15 970.29 Middle- 19.05 cm (7.5 inches) from bottom of 1.3 521.27 fuel 1.3 521L27 970.29 1.0 433.12 811.62 1.5 548.43 1019.17 Top - 21.59 cm (8.5 inches) from bottom of fuel 1.3 494.02 921.24 1.0 411-.68 773.02 After performing the steady state analysis on the hot rod, a study was performed on how the temperature profile and coolant flow rate would change as a function of hot rod power as shown in Figure 4.7.13 and Figure 4.7.14. Figure 4.7.13 shows that after 3 kW, the maximum fuel temperature increases linearly with power, the outer cladding temperature does not change very quickly with changing rod power, and the coolant. temperature begins to level, off around 100 0 C (212'F) at high power. The maximum centerline temperature is shown at the axial location where the maximum fuel centerline temperature is hit. For HEU BOL, this is at the 6th axial node or 13.97 cm (5.5 in) from the bottom ofthe fuel'. The maximum outer cladding is generally 79 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 79 August 2008
also at the 6'ý axial node, because there is no axial conduction within the fuel rod. The, maximum bulk coolant temperature is found at the top of the fuel. Temperature Profile of Hot Rod vs. Power (HEU BOL) 1.0 MW 1.3MW 1,5MW 700 1232
- Max Centerline Temp 600 -- - .....
- Max Outer Cladding Temp
- 1032 500 A Max Bulk Coolant Temp
- 4* 3 i., 832 400
- 632 300 432.. 06 432 200 urn SE 0 . *. 32 0 5 10 15 20 25 30 Power [kW/rod]
Figure 4, 7.13 Temperature profile of hot rod as a function of'power (lIEU BOL) Next, Figure 4.7.14 shows the coolant flow rate as a function of power. As expected the coolant flow rate increases with power, but not at a constant linear rate until rod powers reach about 19 kW. For rod powers greater than 19 kW, coolant flow rate increases approximately linearly with power. 80 August 2008 UWNR~ LEU U"R Conversion Analysis LEU Conversion Analysis 80 August 2008
Coolant Flow Rate vs. Power of Hot Rod (HEU BOL) 1.0 MW 1.3 MW 1.5 MW 0.16 0.35 0.14 0.30 0.12 0 "0.25E 0.1 w M 0.20 3: 0.08 L 0.15 Le 0.06 0 o
- 0.10 0 U 0.04 0 0.02 0.05 0 , 0.00 0 5 10 15. 20 25 30 Power [kW/rod]
Figure 4.7.14 Coolantflow rate of hot rod as afunction of power (lIEU BOL) After determining the fuel temperature at various core power levels and examining how changes in power affect fuel temperature and mass flow rate, it is important to analyze what the DNBR of the hot rod will be. Typically the method used to determine the DNBR is to run a code at a specific power level and then calculate the DNBR based.on the thermal hydraulic conditions obtained for that power level. The RELAP5/MOD 3.3 default CHF correlation is the Groeneveld 1986 correlation where the code calculates the local pressure, mass flux, and quality to compute the DNBR. The problem with using this method is that low power levels calculate considerably higher CHF values than power levels close to a MDNBR of 1.0. Low rod powers have very low quality and thus the correlations, such as Groeneveld 1986, will calculate large values of CHF. UWNR LEU Conversion Analysis 81 August 2008
In lieu of this line of reasoning, the typical DNBR ratio is ahltered from the typical form of local critical heat flux divided by the local heat flux to: DNBR = Rod Power Level whereCorrelationPredicts CHF Actual Rod Power Level Inputted into RELAP The DNBR is calculated by entering in a particular rod power level into R-ELAP5/MOD 3.3 and using the RELAP5 calculated mass flow rate from that rod power level to calculate the numerator. Then the numerator can be solved by using this calculated mass flow rate to determine the power at which CHF is reached for that particular flow rate. The numerator is solved by using an excel spreadsheet which calculates the local heat flux and local critical heat flux as a function of flow rate and rod power. By keeping the flow rate constant, it can be found what rod power level would be necessary to achieve a MDNBR of 1.0 at that flow rate. This calculated number is the numerator used in all further CHF calculations. Since the power to reach CHF is a function of the flow rate, the numerator is also a function of flow rate. It is important to understand the context in which the Groeneveld 2006 and Bernath correlations come from. These correlations, were not developed for use in TRIGA analysis and thus, actual MDNBR numbers coming from the correlations are only as accurate as the correlations themselves in this extrapolated region of usability. At the UWNR, the flow of coolant water has three forces acting upon it. These forces are: buoyancy supplied by the heated water next to the fuel pins increasing water flow, expansion and contraction losses at the end fittings causing drag on the water, and frictional losses with water and fuel element interactions reducing water flow. UWNR LEU ConversionAnalysis 82 August 2008
For the hot rod, the steady state CHF results for core power levels of 1.0, 1.3, and 1.5 MW are shown in Table 4.7.9 for both the Groeneveld and the Bernath correlations. In addition, Figure 4.7.15 shows how MDNBR changes as a function of power. Table 4.7.9 Steady state CHF results of hot irodfor core powers of] 5. 1.3, and 1.0 MW (HEU BOL) Core Power Groeneveld CHF Parameter Bernath
..... .IMWl . . 2006.
Power/rod to reach DNB of 1.00 [kW] 51.74 34.90 M.DNBR 1.51.98 1.29 _ _ _ _ _ _ _1.3 2.17 1.39 1.0 2.71 -1.60 MDNBR vs. Power of Hot Rod (HEU BOL) 1.0OMW 1.3 MW 1.5 MW 12 -- - -- -.--. 0Groeneveld 20N6 a Bernath 10 . . . .. .. .. . . ... -
...----- M ree d 0 86 .. . .. . .. . . . . .. . .
0 ,0* 41 4 1,4 i lB **,*~ .. 0 5 10 15 20 25 30 35 40 45 Power [kW/rod I t'gure 4.7. /5 MI)NBR (?f hot rod as a Junction o'power OfHf U BOL,) In addition to showing the MDNBR as a function of power, the same concept can be shown in a slightly different manner, so the power can be plotted as a function of the flow rate. The RELAP line in Figure 4.7.16 corresponds to the flow rate calculated by RELAP5/MOD3.3 from a given 83 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 83 August 2008
rod power level. Using this calculated flow rate, the poiver at which the correlations predict CHF can be found. The distance between the power where the correlation predicts CHF and the 0 actual rod power shows the margin to CHF at a specific mass flow rate. Thus, Figure 4.7.16 shows the RELAP5/MOD3.3 model's power as a function of flow rate in comparison to what power the Groeneveld 2006 and Bernath correlations would achieve CHF for a specific flow rate. Due to the code beginning to oscillate around 28 kW/rod, a projection was performed, shown by the dotted line, to see when the RELAP5/MOD3.3 model intersects the Bernath correlation. Flow Rate vs. Power of Hot Rod for HEU BOL 1.0 MW 1.3MW 1,5MW 60 50 S40 30 0 20 Groenevled 2006
-- Bernath 10 - RELAP5/MOD3 3 6--&--RELAP5 Projection 0.05 0.07 0.09 0.11 0.13 0.15 0.17 0.19 0.21 Flow Rate [kg/s]
Figure 4.7.16 Power vs..flow rate of hot rod (Correlationpower correlatesto MDNBR 1.00) (lIEU BOL) Since the RELAP5/MOD3.3 single channel model begins to see flow oscillations ýaround 28kW/rod or 0.14387 kg/s (0.3 172 lbm/s) flow rate, it is assumed the flow rate would remain UWNR LEU Conversion Analysis 84 August 2008
constant for rod powers above 28 kW/rod. Thus, the power necessary to reach CHF would be 51.74 kW/rod and 34.90 kW/rod for the Groeneveld 2006 and Bemath correlations respectively for all rod powers above 28 kW/rod. Figure 4.7.16 shows that the Groeneveld 2006 correlation has 3 local maxima due to the effects of the switching of the coolant from negative quality to positive quality, making substantial changes in the K2 and K4 terms. The three local minimums are not physical effects, but rather the Groeneveld 2006 correlation is trying to converge on the minimum power to reach CHF between 15 different axial nodes. If the number of axial nodes were increased it is suspected that the Groeneveld 2006 correlation would smooth out to mitigate the influence of changing axial location of negative quality. The margin to DNB in the UWNR is comparable to that of other TRIGA reactors that have been operating for decades. 4.7.5 HEU Beginning of Life Core Pulse Analysis Pulse analysis was performed using a two-channel RELAP5/MOD3.3 model that implemented the point reactor kinetics model in RELAP5. The first channel is identical to the hot channel described in 4.7.2, and the second channel is based on the geometry, radial heat generation, and axial heat generation of the remaining core. Since the first channel has a different flow area than the fuel elements not next to the transient rod, the hydraulic diameter is not identical to the single channel. The flow area and the wetted perimeter of the core channel were calculated to take the total flow area and subtract out the hot channel. The hydraulic diameter was calculated as 4 times the flow area divided by the wetted perimeter. 85 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 85 August 2008
The total flow area was calculated by the following equation to be 456.039 cm 2 (70.686 in 2): Flow Area 90[Pitch' - -r(Rod Diameter)2 +F/ow Area of D5 SW The total wetted perimeter was calculated by the following equation to be 1024.742 cm (403.44 in): Wetted Perimeter = 907r(Rod Diameter)+ Wetted Perimeter of D5 SW These parameters are shown in Table 4.7.10 and implemented into a two channel model. The core channel has the same axial nodilization as the hot channel. The radial and axial heat generation of the core channel was based off the hot channel heat generation shape. While there are differences in the heat generation between the core and the hot channel, these differences are not expected to be substantial. The purpose of the core channel in this 2 channel model is not to provide limiting thermal hydraulic considerations, but rather a lumped channel that simulates the core. Table 4.7.10 Thermal hydraulic parameters of the 2 channel model Parameter Hot Channel Core Channel Total Flow Area cm 2 4.7429 451.296 456.039 (in 2). (0.7352) (69.951) (70.686) Wetted Perimeter cm 11.4069 1013.335 1024.742 (in) (4.4909) (398.951) (403.44) Hydraulic Diameter cm 1.6632 1.7814 1.7801 (in) (0.65479) (0.70135) (0.7008) In addition to the thermal hydraulic parameters, the-heated perimeter of the core channel must also be changed to reflect the number of fuel elements in the subchannel. The heated perimeter is found by multiplying the number of fuel elements in the. subchannel (90 for the HEU core, 82 UWNR LEU Conversion Analysis 86 Auizust 2008
- for the LEU core) by the heated perimeter of a single fuel element. By incorporating all of these parameters, the entire core's power can be determined during a pulse to determine the maximum fuel temperature in the hot channel. The schematic of the 2 channel model implemented in RELAP5/MOD3.3 can be shown in Figure 4.7.17.
Cold Leg Hot Core Channel Channel Horizontal Connector Figure 4.7.17 The 2 channel RELAP5/MOD3.3 model schematic To be certain that D5SW was the hottest fuel pin, additional fuel rod locations and their respective radial and axial nodilization were analyzed to see if they produced a higher fuel temperature, but D5SW produced the highest fuel temperatures. In order to run the point reactor kinetics model in RELAPS/MOD3.3, the fuel temperature coefficient must be entered in terms of 87 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 87 August 2008
dollars. Table 4.7.11 shows a summary of the negative fuel temperature coefficient from MCNP5 and the associated change in p as fuel temperature increases. P3was calculated from 0 MCNP5 to be 0.00753 + 0.00017. In addition, Figure 4.7.18 shows the prompt negative fuel temperature coefficient as a function of the average core temperature, which is an identical figure as shown in section 4.5.1. Table 4. 7.11 Fuel temperature coefficient for HEU BOL Temperature Fuel Temperature Coefficient Endpoint Midpoint Total kerr Aer A $ Worth I0CI, (10F1) [OCI, (10FI) eff Akff Ap IWr 26.85 (80.33) 1.00311. 0 76.85 (170.3) 0.00458 -0.004573 -0.6072 126.85 (260.3) 0.99853 -0.6072
..... 226.85 (440.3) 0.01409 -0.0143134 -4119036 0
326.85 (620.3) 0.98444 -2.5108
' J>' 426.85 (800.3)'*
4 2 . 0.0 972: .002;Y764.*i-2.7575 526.85 (980.3) 0.96472 . -5.2683 726.85 (1340) 0.89K5 0.06647 -0.076706 -10.1867 926.85 (1700) 0.89825 -15.455 UWNR LEU Conversion Analysis Analysis 88 88 August2008 August 2008 0
Prompt Negative Fuel Temperature Coeffiaent vs. Average Core Temperature (HEU BOL)
*- 20 C
.C 18 U
16 0 k2 14
-12
- 0) 10 CL 8
-j 4 4
2 M 0 i iT 0 100 200 300 400 500 600 700 800 900 1000 z Average Fuel Temperature [°C1 Figure 4. 7. 18 Prompt Negative Fuel Temperature Coefficient vs. Average Core Temperature (HEU BOL) . Since the MCNP5 fuel temperature libraries give identical kfn results over a discrete temperature range it is necessary to calculate the endpoint temperatures in which the MCNP5 libraries calculate a change in kfn- for a particular fuel temperature. The midpoint temperatures are the average temperatures of the flat lines shown in Figure 4.7.18, an identical representation of the MCNP5 results presented in Figure 4.5.9. Thus, Table 4.7.11 is broken up into two parts. The temperature in which the MCNP5 model calculated kITr is color coded white. The calculated values at the midpoint between two adjacent MCNP5 ktT data points is color coded blue. The third column labeled 'ken-' is the MCNP5 calculated krr data point at a particular temperature. The fourth column labeled 'Ake"' is calculated as the upper kT" data point minus the lower kITdata point. The fifth column labeled 'Ap' is calculated as the reactivity of the upper Analysis 89 August 2008 UWNR LEU Conversion Analysis 89 August 2008
kerr data point minus the reactivity of the lower kff data point. The sixth column labeled '$' is the fifth column divided by the delayed neutron fraction, P3.The last column labeled 'total worth [$]' is directly entered into the RELAP5/MOD3.3 input decks for the negative fuel temperature coefficient. At the first data point there is no negative fuel temperature coefficient,, thus the first entry is zero. Each subsequent data point is the addition of the 6th column to the previous RELAP negative fuel temperature coefficient. Thus, the total negative fuel temperature worth is the sum 'of the 6 1h column or -0.116377Ak/k (-$15.455). The negative fuel temperature coefficient shown in Figure 4.7.18 is calculated as Ap divided by the change in end point temperatures. The negative fuel temperature coefficient at 726.85°C (1340.337F) is -1.9176 x 10-' Ap/K (-0.02547 $/K). Moderator reactivity feedback was not included in the model. In order to justify this, two cases were run with and without moderator feedback, and there was no difference between the pulse powers or the peak fuel temperatures. Since pulse transients occur on such short time scales (less than 0.1 seconds) and the thermal resistance.in the fuel, gap, and clad are significant, the heat flux on the outer surface does not change appreciably until several seconds after the pulse has initiated. Erbium reactivity feedback was included in the MCNP model which was used to calculate the prompt negative fuel temperature coefficients as reported in Table 4.7.11. In addition, all pulsing analyses assumed a 0.1 mil gap. Since the time scale of a pulse is very short, the gap size does not affect the maximum fuel temperature after a pulse appreciably. UWNR LEU Conversion Analysis
.......... d-90 August 2008
Technical specifications require that a SCRAM must occur within 15 seconds of the initiation of a pulse. This analysis assumes that the transient rod remains out of the core for 15.0 seconds and then all scrammable control elements are frilly inserted into the core. To determine if the two-channell model predicts measured. pulses, a comparison is shown in Figure 4.7.19 of the pulse power versus time for a 1.34%Ak/k ($1.78) pulse at BOL compared with -the measured pulse. This is plottedout for a period of 0.10 seconds past the initiation of the pulse;which is well after the peak pulse power and fuel temperature have been achieved. Initial powver of" the core is specified to be I kW prior to pulse per technical specifications. All pulses are modeled with instant reactivity insertion to give the most limiting pulse power and maximum fuel temperature, Power Profile of 1.34% Ak/k Pulse 10000 1 Measured Data 0-RELAP- 1000 calculated
~-100 0
0. 10 0 0.0a d.0i t.te 0.04 &05 0.06 0.07 0.oB 0.09 0.1 L Time [s] Figure 4.7.19 Power vs. Time ofl134 %klk ($1.78) Pulse/for lIEU BOL In addition to the power profile, the total energy of the plots can be compared as seen in Figure 4.7.20. Comparing the RELAP5/MOD3.3 model with the measured pulse trace, it can be seen 91 August 2008 UWNR LEU Conversion Analysis LEU. Conversion Analysis 91 August 2008
that the total peak power and integrated power of the pulse of the RELAP5 model is 1963.69 MW and 24.6 MJ after 0.088 seconds respectively and the measured pulse trace is 858.9 MW and 12.7 MJ after 0.088 seconds respectively. This demonstrates that the RELAP5/MOD3.3 model is bounding. In addition to being limiting, the RELAP5/MOD3.3 model has a similar power profile and period to that of the measured data. Total Energy vs. Time of 1.34% Ak/k Pulse 30
. Measured UWNR Pulse 25 -- RELAP calculated Pulse 20 10 0
0 0.01 0.02 0.03 0.04 0.05 0.06 0.07 0.08" 0.09 0.1 Time [s] Figure 4, 7.20 Integral energy prqjile of pulse vs. time at the ROL For a maximum pulse of 1.4%Ak/k ($1.86) according to technical specifications, the maximum fuel temperature, peak pulse power, and total energy released during a pulse can be found. During the pulse, the maximum fuel temperature of the fuel rod is found when the total energy released during a pulse has reached a plateau. The total energy released is the summation of the change in core power multiplied by the time step. For a 1.4%Ak/k ($1.86) insertion in D5SW, the maximum fuel temperature is 661.05'C (1221.897F) and occurs at 0.060 seconds from the UWNR LEU Conversion Analysis 92 August 2008 0
ejection of the transient rod. This is a 604.93 0 C (I 120.87°F) change in temperature from the initial starting power of'] kW as shown in Figure 4.7.21. The thermocouple temperature location in the hot rod is also shown for comparison as the temperature builds in. While this thermocouple does not exist in D5SW, it is useful to see how the temperature builds in .to the center of the fuel after the pulse. The maximum pulse temperature does not exceed the. fuel temperature safety limit of 1150'C (2100'F) or the operating fuel temperature limit of 830'C (I 526'F) 20 at any time during the pulse. Maximum Temperature vs. Time of 1.4%Ak/k Pulse 1200 --- 2032
-L. 800 . ----------- -1532 ."
1 000E 000f .. .1032 . E 400
- E
-*',Max Temp [CI i-
'200- Safety Analysis Limit 532
- -- -- Operating Limit
-Thermocouple' Location inD5SW 32 0 0.05 0.1 0.15 0.2 0.25 Time [s]
Figure 4.7.21 Maximum Pulse Temperature vs. Tiviefiar a 1,4%?dk'k reactivitv insertion at i)5SW The peak pulse power for a 1.4%Ak/k insertion is 2.29 GW and the total energy is 28.05 MWafter 0.25 seconds. The pulse power and the total energy curves can be seen in Figure 4.7.22 and Figure 4.7.23 respectively. The peak pulse occurs 0.045s after the transient rod has been ejected. UWNR LEU Conversion Analysis '93. August 2008
Core Power vs. Time of 1.4% Ak/k Pulse 0 1.E+10 11E+09 1.E+08 11E+07 a. 1.E+06 11E+04 1.E+03 ____ 0 0.05 0.1 0.15 0.2 0.2s Time [s] Figure 4.7.22 Power Profile of 1.4%'k/k pule versus time/obr lIEU BOL Integral Core Energy of 1.4% Ak/k Pulse vs. Time 1.E+08 1.E+07 1.E+06 1.E+05 b9 1.E+04 a,I Wa 11+.03 11E+02 . L.E+01 0 0.65 0.1 0.15 0.2 0,25 Time [s] Figure 4.7.23 Energ.'v of 1.4?6z1 k/k pulse vs. time for HEU BOL 94 August 2008 UWNR LEU Conversion Analysis UWNR Analysis 94 August 2008
It is important to note that the maximum fuel temperature occurs, at the outside edge of the fuel where the radial power distribution is the highest. The temperature profile, when the fuel reaches its highest temperature can be seen in Figure 4.7.24. This happens at a point 1.6994 cm (0.669 1 in) from the fuel centerline and 16.5 1cm (6.5 in) from the bottom of the fuel. In addition to this surface plot, the peak axial and radial pulse temperature distributions are shown separately in Figure 4.7.25 and Figure 4.7.26 respectively. 35 U- 40 E 25ý
*6 -400
.Ea20
&i 200 17o oT*
0 0.5 1 1.5 2 Radial Distance from Fuel Centerline 1cm] Figure 4, 7.24 Fuel temperaturel
/Cj surface plot oqfD5 SW O,06s after 1.4%MJlk pulse (HEU BOL) 95 August 2008 UWNR LEU UWNR Conversion Analysis LEU Conversion Analysis 95 August 2008
0 Hot Rod Axial Temperature Distribulon 0.06s After 1.4% Ak/k 2 4 Axial Distance from Fuel Centerline [in] 6 8 10 12 14
- 0 700 .-.... ......... . . . . .. . --..--.--......
1262
-650 1
** - 1162 S600 o a
- 1062
-550
* "962 M*
500- -- 96
- 0. *0 .
E 8 E S45086 400 - 762 7 350 3 ................... ......... ............ 662 0 5 10 20 25 30 3S 40 Axial Distance from Bottom of Fuel [cm) Figure 4.7.25 Axial temperature distribution of hot pin 0. 06s a*t er pulse 1. 67cm from centerline (lIEU 80L) Hot Rod Radial Temperature Distribution O.06s After 1.4% Ak/k Pulse 0 0.1 0.2 Radial Distance from Fuel Centerline [in] 0.3 0.4 05 0.6 0.7 0 700 . 600-*1232 600 ÷ 500 *1032 *
* * * * * *....832
,- 400 . 3 0 ______ 632 2 G 300 - CL CU CL E "E 43220 200I 232 100 - 32 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 Radial Distance from Fuel Centerline [cm] Figure 4.7.26 Radial temperature distribution ofhot pin O.06s qfterlulse 16.5/cm from bottom offuel (HEU 1301j UWNR LEU. LEU Conversion Conversion Analysis Analysis 96 96 August 2008 August 2008 0
It is important to also determine what size pulse would be required to exceed the fuel temperature safety limit of 1150'C (21027F) or the operating fuel temperature limit of 830'C (1526'F). To determine these pulse sizes, various pulses were run from I.0%Ak/k ($1.33) to 2.3% Ak/k ($3.05) in increments of 0.1%Ak/k. The maximum fuel temperature, peak pulse power, and total pulse energy were plotted in Figure 4.7.27 through Figure 4.7.29. Maximum Pulse Temperature vs. Pulse Size 1200 7- 2172 1 1972 1000 41772
' ' 900 ------
800 1 - - -............ 4 1572 o. 1372 0 700 a-J . E I 1172 E ______ Calculated 972 500 S-
-Operating Limit 772 400-
- Safety Analysis Limit 300 , 572 S
1.2 1,4 1.6 1.8 2 2.2 2.4 Pulse Reactivity Insertion [%Ak/k] Figure 4.7.27 Maximum pulse temperaturein hot rodas a./imction qfpulse size (lIEU BOL) Figure 4.7.27 shows the maximum reactivity insertion before the operating fuel temperature limit and safety analysis fuel temperature is exceeded is 1.6%Ak/k ($2.12) and 2.2%Ak/k ($2.92) respectively. A pulse of 2.3%Ak/k ($3.05) achieves a maximum fuel temperature of 1155.65'C (2112.17°F), which is 5.65'C (I0.17 0 F) greater than the fuel temperature safety limit. The fuel Analysis 97 August 2008 UWNR LEU Conversion Analysis LW Conversion 97 August 2008
temperature after a pulse is comparable to that of other TRIGA reactors that have been operating for decades. 0 Maximum Pulse Power vs. Reactivity Insertion 12 10 iA 6 E X0 M 2 00 0 rr-- -- 1 1.2 1.4 1.6 1.8 2 2.2 2.4 Pulse Reactivity Insertion [%Ak/k] Figure 4.7.28 Maximum pulse power as afinction of reactivity insertion (HIEU BOL) S UWNR LEU LW Conversion Analysis Analýsis 98 98 August 2008 August 2008 0
Total Pulse Energy vs. Reactivity Insertion 60 50' 40
- w 30 _ _ _ _ _ _ ____
1 20 10 0 " 1 1.2 1.4 1.6 1.8 2 2.2 2.4 Pulse Reactivity Insertion [%Ak/k] Fýiure 4. 7.29 Total pulse energy as a/unction oj reactivity insertion (HtEU BOL) 4.7.6 LEU Power Summary The LEU core pin power peaking factors and the hot rod radial and axial power peaking factors are shown in Figure 4.7.30 through Figure 4.7.34. The pin power peaking *factors were determined with MCNP5 with hot conditions and the control blades at the critical bank height. The pin, radial and axial power peaking factors are used to compute the expected fuel temperature and margin to CHF for the LEU fuel. Due to the removal of 2 bundles, or 8 fuel rods, there are 83 fuel pins in the LELJ core and thus pin powers are higher for the LEU fuel than the HEU fuel. The hot rod power is 19.36kW at 1.0 MW, with a 1.61 pin power peaking factor. This compares with the HEU hot rod power of 17.60kW at 1.0 MW. Given that the hot rod 99 August 2008 UWNR LEU U"R Conversion Analysis LW Conversion Analysis
- 99 August 2008
power is higher, it can be expected that the maximum fuel temperature will be higher and margin to CHF will be lower compared to HEU. 0 Figure 4.7.30 Pin Power [kW/rod/ andpower peakingJctors (PPF) of the UWNR core at 1.0 AlW (LEU BOL) 0 100 August 2008 0 UWNR LEt) Conversion Analysis LELJ Conversion Analysis 100 August 2008
Figure4.7.31 Pin Power [kW/rod] andpower peakingfactors (PPF) of the UWNR core at 1. 0 MW (LEU MOL) aithe U*INR core at 1.0 MW (LEU EOL) Figure 4.732 Pin Power [kW/'rod] andpower peaking.factors (.PPF) UWNR LEU Conversion Analysis 101 August 2008
Radial Power Distribution for LEU Core 1.S
*BOL a 1.4 C I A 0 o MOL 1,3 "A---
I
"- 1.2 A EOL A - 1.1 3
0.
- 0.9
- A.: *A 0.8 0.7 0 0,2 0.4 0,6 0.8 1 1.2 1.4 1.6 1.8 Radial Distance from Fuel Centerline [cm]
F'igre4. 7.33 Normalized radialpower densav disLribution ofhot rod./br LEU core Axial Power Distribution'for LEU Core 1.6 1.4
- BOL 0 C A A MOL 0
1.2
- 4. SA A. EOL A
1 U
.0 0.8 a X 0.6 0.6 U 0.4 0.2 r -1 I I - --
0 0 S 10 15 20 25 30 35 40 Axial Distance from Bottom of Fuel [cm]
!:igure 4.7.34 Nortmalized axial power distribution of hot rod for .At"U core 102 August 2008 UWNR LEU Conversion Analysis Analysis 102 August 2008
If'the reactor staff needed to replace the hot rod located at D5 SW with a fresh fuel pin at either MOL or EOL the pin power peaking factor would increase to 1.68 and 1.74 respectively. However, at no time during the replacement of the hot rod with a fresh fuel pin would CHF, fuel temperature limits in air and water or the operational temperature limit be exceeded during steady state and accident analysis scenarios 21 . While no fuel failure is expected, the margin to these limits would decrease. To relieve these concerns, the reactor staff would substitute a burned fuel pin in D5 SW and put fresh fuel in the periphery of the core to decrease the magnitude of the pin power peaking factor change. Furthermore, the fuel temperature limits for the 30/20 LEU fuel are identical to the HEU FLIP fuel. 2 The water cooled fuel temperature safety limit is I150'C (2100 0 F) and the air cooled fuel temperature safety limit is 950'C (1740'F) as described in section 13.3.3. Currently the IFEs in the HEU core are located at D4 SW and E3 NE. The intent is to maintain the IFEs at these locations in the proposed LEU core. Therefore, in order to determine if these core positions are still valid for the LEU core, an analysis was conducted to ascertain appropriate core channels for the IFE locations. The LSSS for. the IFE is 400°C in order to prevent the hot rod from reaching the fuel temperature safety limit of I150'C. However, if the IFE is placed in a very cold location, such as B3 NE, the reactor power would have to go well above the trip set point of 1.25 MW before reaching the fuel temperature LSSS of 400'C at which point the CHF of the hot rod would be possible. This analysis determined which channels were appropriate to ensure that with an LSSS of 400'C an IFE would protect the fuel temperature safety limit. 103 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 103 .August 2008
The power when CHF would occur in the hot rod is calculated to be 35.164 kW/rod with the Bernath correlation, or a total core power of 1.816 MW. This number makes the limiting assumption that the mass flow rate of water will not increase as core power increases once the
- RELAP5 model predicts the flow will oscillate. However, further calculations have shown that if one assumes the flow rate continues on the same linear trend as shown by the dashed line in Figure 4.7.43, the power to reach CHF would more realistically be approximately 41 kW/rod with the Bernath correlation or a total core power of 2.113 MW as seen in Figure 4.7.45. This provides additional margin to the power required to reach CHF by 16%. All further analysis used 1.816 MW as the core power to compute the maximum thermocouple temperature.
Furthermore, instead of using the hot rod axial and radial power profiles, the cold rod axial and radial power profiles were used to calculate the temperatures of the thermocouples. The thermocouples are located 0.3 inches (0.762 cm) from the fuel centerlinej 6.5, 7.5, and 8.5 inches (16.51, 19.05, 21.59 cm) above the bottom of the active fuel. A comparison between the hot (D5 SW) and cold (B3 NE) rod power peaking shapes is shown in Figure 4.7.35 and Figure 4.7.36. The axial peaking shapes of the cold rod are lower in the cold rod than the hot rod in the region of the thermocouple. Thus, it is expected that by using the cold rod axial peaking factors for all rods, the maximum predicted thermocouple will be bounding. UWNR LEU Conversion Analysis 104 August v 2008
,Radial Power Distribution (LEU BOL) 18 B3 NE 1 -,6 Li.. OD5SW
. 1.4
- 0. 1.2 0
Cu 0.8 , 0 6---- . ......... 0 02 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 Radial Distance from Fuel Centerline [cm] Figure 4. 7.35 Radial power distribution comparisonl (LEU BOL) Axial Power Peaking Factor (LEU BOL) 14 L. 1.3 U
-C-1.2 --------
. . .* . . . . . .° . .
U: .. . . . . 11
- 3.
- Cu 1 CL 0.9
,. 0.8 03 Cu0.6
- 0.4 0 5 10 15 20 25 30 35 40 Axial Distance from Bottom of Fuel [cm]
Figure 4. 7.36 Axial power distribution comtpurison (LEU 130L) UWNR LEU Conversion Analysis 105 August 2008
All analyses conducted used a hot gap size of 0.1 mils as previously assumed for the hot rod. Then, using the pin power peaking factors of the core, the. maximum thermocouple reading for each rod in the core was calculated. It was noticed during the course of the analysis that the predicted maximum thermocouple reading had a very linear shape as a function of the pin power peaking factor. Thus a least squares regression line was found to see what the maximum thermocouple temperature would be as a function of the pin power peaking factor. This equation is: T = 339.794(PPF) + 130.714 where PPF is the pin power peaking factor and T is the maximum thermocouple temperature in °C. This is shown in Figure 4.7.37. Thermocouple Temperature as a function of pin power peaking factor (LEU BOL) 500 y = 3.39794E+02x + 1.30714E+02 R'= 9.99921E-01
= 450 E
2 400 0 E 350 --------- _ _ _ 205 0.4 0.5 0.6 0:7 0.8 0.9 1 Pin Power Peaking Factor Figure 4.7.37 Thermocouple temperature as a function of pin power peakingfactor Conversion Analysis 106 August 2008 UWNR LFU LJWNR LW Conversion Analysis 106 August 2008
With this equation, a map of the maximum predicted thermocouple temperatures across the LEU BOL core is shown in Figure 4.7.38. Note that these are maximum predicted thermocouple readings, not maximum predicted fuel centerline temperattires. While 400'C is the LSSS for the thermocouple, acceptable, thermocouples should not read exactly 400'C at the predicted lower bound of power to reach CHF. That does not leave adequate margin, thus it was decided to give 25°C of further margin. Thus the acceptable positions in the core would need to have a pin power peaking factor of at least 0.866 to protect the hot rod from reaching CHF. The acceptable positions are shown in Figure 4.7.39 where the red (X) signifies the predicted temperature is less than 400, a yellow (!) signifies the temperature is between 400 and 425 and a green check mark (<) signifies the temperature is greater than 425 at 1.816 MW and thus is a valid location for the IFE. 107 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 107 August 2008
Figure 4.7.38 Predictedmaximum IFE temperaturesacross the core at a power of 1.816 MW (LEU BOL) 108 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 108 August 21008
Figure 4.7.39 Possible thermocouple locations of the UWNR (LEU BOL) The analysis was only conducted for LEU BOL, however the pin power peaking factors in D4 SW or E3 NE do not change substantially with bumrup. As shown in Figure 4.7.30 through Figure 4.7.32, these peaking factors change by less than 1.5%. UWNR LEU Conversion Analysis 109 August 2008
4.7.7 LEU Beginning of Life Core Analysis All LEU fuel results use the exact geometry set forth in the HEU I channel model used in section 4.7.4. The changes made in the single channel model are the radial and axial peaking factors and the rod power given in section 4.7.6. The radial temperature is plotted in Figure 4.7.40 given with uncertainty of 50% for the gap size. For all analysis afterwards it is assumed the gap size between the fuel and the cladding is 0. 10 mil (0.00254mm). The axial temperature profile is shown in Figure 4.7.41 and a summary table of steady state results for 1.0, 1.3, and 1.5 MW are shown in Table 4.7.12. In addition, the temperature profile and the coolant flow rate are shown as a function of power in Figure 4.7.42 and Figure 4.7.43 respectively. Radial Temperature Distribution of Hot Rod at 1.5 MW (LEU BOL) Radial Distance from Fuel Centerline [in] 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 800 - 1412 700 A A A 1212 S. 600 A AA A A A
** 1012 500 --
AL **812 400 -- " E 300 612 E* 200 412
. 0.05 mils a 0.10 mils A 0.15 mils 100 1 212 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 Radial Distance from Fuel Centerline [cm]
Figure 4.7.40 Radial temperaturedistrihutiona/hot rod at 1.5. MJ'for varying gap sizes (LEU BOL) UWNR LEU Conversion Analysis J 110 Auust 2008 0
Table 4.7.12 Steady state results of hot rod at 1.5. 1.3, and 1.0 MW (LEU BOL) Core Power Parameter CMWe SI English, 1.5 29.041 Rod Power [kW] 1.3 25.169 1.0 19.361 1.5 0.14878 0.32799 Flow Rate for Hottest Channel [kg/s], [lbm/s] 1.3 0. 13 143 0.28975
.1.0 0.10535 .0.23226 1.5 0.34546 1.13340 Maximum Flow Velocity [mi/s], [ft/s] 1.3 0.30327 0.99498 1.0 0.23297 0.76434 1.5 925.57 Maximum Wall Heat Flux [kW/m 2] 1.3 802.68 1.0 617.44 1.5 662.83 1225.09 Maximum Fuel Centerline Temperature [°C], ['F] 1.3 594.4 1101.92.
1.0 490.1,5 914.27 1.5 141.6 286.88 Maximum Outer Clad Temperature [°C], [°F] 1.3 139.6 283.28 1.0 136.3 277.34 1.5 127.47 261.45 Exit Outer Clad Temperature [°C], [°F] 1.3 127.14 260.85
-- 1.0 125.09 257.16 1.5 101.32 214.38 Exit Bulk Coolant Temperature [°C], [°F] 1.3 100.04 212.07 1.0 98.23 208.81 11] August 2008 UWNR LEU Conversion Analysis Conversion Analysis III August 2008
Axial Temperature Profile of Hot Rod at 1.5 MW for LEU BOL Axial Distance from Bottom of Fuel [in] 0 2 4 6 8 10 12 14 50 700 - ____________- 600 t .... ....... 0
- Centerline
- Outer Cladding 1232 1032 Up, A Bulk Water 832 400 632 300 E
WJ 432 Ew 2UU U U a . LA~A * * -~.U A--- U .A 100 23.2 o 5 A5 oA 0 32 0 5 10 15 20 25 30 35 40 Axial Distance from Bottom of Fuel [cm] Figure 4.7.41 Axial temperature profile ofhot rod at 1.5 MW (LEU BOL) Temperature Profile of Hot Rod vs. Power (LEU BOL) 1.0MW 1.3 MW 1.5 MW 800 1432 Max Centerline Temp + 700 m Max Outer Cladding Temp + 1232 600 A max Bulk Loolant lemp 4P 1032 Z' UU S500 832 " M 400
- 0) 632 C.
C-E 300 E 200 432 I 100I a__a._ _ _ N n a a .a 2,32 0AA A A5A A A .AAA------.
--- 32 0
0 5 10 15 20 25 30 35 Power in Hot Rod [kW] Figure 4. 7.42 Temperature profile ofhot rod at.D5 SW ws. power of hot rod (LEU BOL) August 2008 UWNR LEU Conversion Analysis UWNR. LEU 112 Analysis 112 Conversion August 2008
0 0.25 Coolant Flow Rate of Hot Rod vs. Power (LEU BOL) 1.0MW 1.3MW 1.5MW
- RELAP Calculated Flow 0.5 Rate , -
77 0.2 -
- -Extrapolated Region 0.4 E
.0 M 0.15 ,
0.3 LA.* 0 1
- 0. -- U-.
** ***0' 0.2 0 M o 0 V.. 0.05 -- . Q..
0 -,----0 0 5 10 15 20 25 30 35 40 45 Power in Hot'Rod [kW] Figure 4.7.43 Coolathflow rate of'hot rod at D5 SW vs, power f hot rod (LEU BO1) In Figure 4.7.43; the dotted line is a projection of the flow rate since RELAP5/MOD3.3 predicted an oscillating flow rate at 29 kW/rod. This projection was determined using the previous trend starting from 19.4 kW/rod. After determining the steady state temperatures and flow rates, the steady state results were found for core powers of 1.0, 1.3, and 1.5 MW shown in Table 4.7.13. The MDNBR as a function of rod power was plotted in Figure 4.7.44 and the 0 power to which CHF would be achieved as a function of flow rate is shown in Figure 4.7.45. Analysis 113 August 2008 UWNR LEU LEIJ Conversion Analysis 113 August 2008
Table 4.7.13 Steady state Cl-IF results of hot rodfor core powers of 1.5, 1.3, and 1.0 MW(LEUBOL) CHF Parameter Core Power Groneveld 0 IMWI] 2006 Power/rod to reach DNB of 1.00 [kW] 52.786 35.164 1.5 1.818 1.211 MDNBR 1.3 2.095 1.331 1.0 2.680 .1.520 MDNBR vs. Power of Hot Rod (LEU BOL) 12 10MW 1.3MW 15MW 10*-
- Groeneveld 2006 a Bernath Z 6 -
-- DNBR= 1.0 t------ - -___ km A- a 0
0 10 15 20 25 30 35 40 45 Power in Hot Rod [kW] Figure 4.7.44 MDNBR as afinction of hot rodpower (LEU BOL) Conversion Analysis LEU Conversion UWNR LEU Analysis 114 114 August 2008 August 2008 0
Flow Rate vs. Power of Hot Rod (LEU BOL) 1,0 MW 1.3MW 1.5 MW 70 60 50 0 40 0 30 a. 0 20 0. 10 0 0,04 0.06 0.08 0.1 0.12 0,14 0.16 0.18 0.2 0.22 Flow Rate [kg/s] Figure 4.7.45 Power to reach CtIFas a/function of/low rate (LEU 1301.), With higher power levels in the hot rod in the LEU BOL core, the margin to CHF is lower than the hot rod in the HEU BOL core. Since the flow rate began to oscillate around 29 kW/rod the rod power at which CHF would result was calculated at the last known, stable flow rate of ,0.14417 kg/s (0.3 1784 lbrn/s). Thus, the power necessary to reach CHF at LEU BOL would be 52.786 kW/rod and 35.164 kW/rod for the Groeneveld 2006 and Bemath correlations respectively for all rod powers above 28 kW/rod. This is comparable to the power necessary to reach CHF calculated in the HEU BOL analyzed in section 4.7.4. The IFE positions for the LEU core are intended to be placed in the identical location as the current HEU core located at D4 SW and E3 NE. The pin power peaking factors of these locations at BOL are 1.47 and 0.89 respectively. Each IFE location has 3 axial thermocouple UWNR LEU Convers ion -Arialysi s 1.15 August 2008
locations located at the axial fuel centerline, 1 inch (2.54 c6mn) below the axial fuel centerline and 1 inch (2.54 cm) above the axial fuel centerline. The thermocouples are 0.3 inches (0.762 cm) from the radial fuel centerline. In order to provide reasonable predicted thermocouple temperatures, nominal operating conditions were used. These nominal operating conditions are: 1.0 MW reactor power, water 20 feet (6.096m) above the core, and an inlet. water temperature of 30'C (86°F). In addition, tolerance bands have been added to account for the.uncertainty in the gap. Using previous methodology in the HEU analysis, the gap size will be nominally set to 0.1 mils (0.00254 mm) with a tolerance of +/- 50%. Thus the gap size will range from 0.05 mils to
- 0. 15 mils (0.00 127 to 0.00381 mm) to give a band of uncertainty. The results of the IFE analysis for LEU BOL are shown in Table 4.7.14.
Table 4.7. 14 Expected IFE Temperature results at 1.0 MW (LEU BOL) Expected Lower Bound Upper Bound IFE Location Temperature Temperature Temperature oC (OF), °C (OF) 0C (OF) Bottom 445.46 (833.83) 398.15 (748.66) 489.391(91.2.89) D4 SW Center 437.32 (.819.18) 391.03 (735.85) 480.35 (896.63) Top 423.42 (794.15) 378.92,(7'14.05) 464.88 (868.78) Bottom 299.27 (570.69) 270.90 (519.62) 326.27 (619.29) E3 NE Center, 291.37 (556.47) 264.14 (507.44) 3117.34 (603.20) Top 279.87 (535.77) 254.30 (489.74) 304.30,(579.74) Since the pin power peaking factor for D4 SW and E3 NE does not change significantly during core lifetime as seen in section 4.7.6, the expected IF.E temperatures are not expected to'change significantly. The pin power peaking factors for D4 SW increase from 1.47 at BOL to 1.49 at MOL and decrease to 1.47 at EOL. The pin power peaking factor for E3 NE varies from 0.88 at BOL to 0.87 at EOL. Since the IFE locations have very similar pin power peaking factors, very UWNR LEU Conversion Analysis 116 August 2008
little change in the IFE temperature results are expected. Thus only the LEU BOL numbers have been calculated for the IFE locations. 4.7.8 LEU Middle of Life Core Analysis The differences between the BOL and MOL stages of core lifetime for the LEU core are the core power peaking factors in addition to the radial and axial power profiles shown in section 4.7.6. The radial temperature is plotted in Figure 4.7.46 given with uncertainty of 50% for the gap size. For all analysis afterwards it is assumed the gap size between the fuel and the cladding is 0.10 mril (0.00254mm). The: axial temperature profile is shown in Figure 4.7.47 and a summary table of steady state results for 1.0, 1.3, and 1.5 MW are shown in Table 4.7.15. In addition, the temperature profile and the coolant flow rate are shown as a function of power in Figure 4.7.48 and Figure 4.7.49 respectively. m UWNR LEU Conversion Analysis 117 August 2008
Radial Temperature Distribution of Hot Rod at 1.5 MW (LEU MOL) Radial Distance from Fuel Centerline[in) 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 8 00 .. . . . . . . .. . . . . . . .. . . . . . ., 1412 700 A A A A-* . A AL 1212 600 . A LA. 1012 500 - - CL um AA 812 400 . %*Aa CL E a, 612 E IT 300 412, 200 - 4...
- 0.05 mils u 0.10 mils A 0.15 mils 100 + . . . ._.. .. . .212 212 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 Radial Distance from Fuel Centerline [cm]
Figure 4.7.46 Radial temperature distribution af/hot rod at 1.5 MW.for varying gap sizes (LEU MOL) Axial Temperature Profile of Hot Rod at 1.5 MW for LEU MOL E Axial Distance from Bottom of Fuel [in] 0 2 4 6 8 10 12 14 700 600 -"- ....... " ' .. .
- Centerline 1232 600
- Outer Cladding[ 1032 I-
- Bulk Water -32o 632 300 -- -. ................ --...............
E 432 200 V I-- A £ £ A A A A A A - . A .. . 232 32 0 5 10 15 20 25 30 35 40 Axial.Distance from Bottom of Fuel (cm] Figure 4.7.47 Axial temperature profile o/'hot rod at 1.5 MW (LEU MOL) UWNR LEOLEU Conversion Analysis Analysis 118 118 August 2008 August.2008 0
Table 4.7.15 Steady state results ofjhot rod at 1.5, 1.3, and 1.0 A4W (LEU MfOL) Parameter Core Power SI English ________________________________ MWI 1.5 28.888 Rod Power [kW] 1.3 25.037 1.0 19.259 1.5 0.14781 0.32587 Flow Rate for Hottest Channel [kg/s], [Ibm/s] 1.3 0.13049 0.28768 1.0 0.10457 0.23054 1.5 0.34932 1.14606 Maximum Flow Velocity [m/s], [ft/s] 1.3 0.30029 0.98520 1.0 0,23098 0.75781 1.5 914.851 Maximum Wall Heat Flux [kW/m 2 ] 1.3 793.381 1.0 610.293 1.5 665.06 1229.11 Maximum Fuel Centerline Temperature []CJ, f[F]. .1.3 596.27 1105.219 1.0 491,51 91.6.72 1.5 141.43 286.57 Maximum Outer Clad Temperature [°C], [°F] 1.3 139.45 283.01 1.0 .136.16 277.09 1.5 .127.25 261.05 Exit Outer Clad Temperature [°C], [1F] I.3 126.93 260.47 1.0 124.91 256.84 1.5 101.16 214.09 Exit Bulk Coolant Temperature [°C], [OF] 1.3 100.13 212.33 1.0 98.33 208.99 119 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 119 August 2008
Temperature Profile of Hot Rod vs. Power (LEU MOL) 1.0MW .3MW 1.5MW 800 1432
- Max Centerline'Temp
- 700 W
- Max Outer Cladding Temp 1232 600
- Max Bulk Coolant Temp 1032 Z7 500 832 400 632 300 *_.* * -----.l E E a, W~~..uuu~umm~m ulm, mu UEU a, A AAA i~ik~ Ar*k** *~~ *A 432 200 100 232 0 32 0 5 10 15 20 25 30 35 Power in Hot Rod [kW]
Figure 4.7.48 lernperaturL profile of hot rod at D5 SW4vs. power o hot rod (LEU MOL) Coolant Flow Rate of Hot Rod vs. Power (LEU MOL) 1.0MW 1.3 MW 1.5MW 0 0.25
- RELAP Calculated Flow 0.5 Rate
*;. 0.2
~0
. . Extrapolated
. .. Region 0.4 E
-a 4.1 M
0.3 ci 0* 0 0.1 C 0.2 E 0 0 U~ 0.05 0.1 U o -.---
- . ................. i.. 0.0 0 59 10 15 20 25 330 35 40 45 Po wer in Hot Rod [kW]
Figure 4. 7.449 Coolant flow rote of'hot rod at D5 SW is. power of/hot rod (LEbU UO.) UWNR LEU Conversion Analysis 120 August 2008 S
As can be seen from Figure 4.7.49, the flow rate at-29 kW/rod oscillated according to RELAP5/MOD 3.3, so a projection was formed as seen by the dashed line. After determining the steady state temperatures and flow rates, the steady state results were found for core powers of 1.0, 1.3, and 1.5 MW shown in Table 4.7.16. The MDNBR as a function of rod power was plotted in Figure 4.7.50. and the power at which CHF would be achieved as a function of flow rate is shown in Figure 4.7.51. Table 4.7.16 Stead, state CHF results of hot rodfor core powers of 1.3. and 1.0 MW (IEU MOL) CHF Parameter Core Power Groneveld Bernath IMWI 2006 Power/rod to reach DNB of 1.00 [kW] 52.261 35.345 1.5 1.809 1.224 MDNBR 1.3 1.982 1.339 1.0 2.678 1.527 MDNBR vs. Power of Hot Rod (LEU MOL) 1.0 MW 1.3 MW 1.5 MW 12 10
- Groeneveld 2006 8
gernath B z 6 U V 4 - - DNBR = 1.0 U 4 4 9~ @@@
£ 2 -
= wmo a i8umIi
- O*': # I
-------- -- -- ---- ---Fm&m U-U a Ire 0
5 10 15 20 25 30 35 40 45 Power in HotRod [kW] Figure 4.7.50 MDNBR as aUfmction a /hot rodpower (LEU MOL) 0 UWNR LEU Analysis Conversion Analysis LEU Conversion. 121 121 August2008 August 2008
Flow Rate vs. Power of Hot Rod (LEU MOL) 1.0MW 1,3MW 1.5MW 70 60 50 0 40 0 30
-- Groeneveld 2006 0
M, 20 - Bernath 1in ----- R_- RFI APR
- -- Extrapolated Region 0
0.04 0.06 0.08 0.1 0F12 0.14 0.16 0.18 0.2 0.22 Flow Rate [kg/s]
-igure 4.7.51 Power to reach (C1Fas a/fimrwtion qf/low rate ('LEU MOL)
With similar power levels in the hot rod in the LEU MOL core, the margin to CHF is similar to the CHF margin in the hot rod in the LEU BOL core.. Since the flow rate began to oscillate around 29 kW/rod,, the rod power at which CHF would result was calculated at the last known stable flow rate of 0.14394 kg/s (0.31733 lbm/s). Thus, 'the power necessary to reach CHF at LEU MOL would be 52.261 kW/rod and 35.345 kW/rod for the Groeneveld 2006 and Bernath correlations respectively for all rod powers above 28 kW/rod. This is comparable to the power necessary to reach CHF calculated in the LEU BOL analyzed in section 4.7.7. 4.7.9 LEU End of Life Core Analysis The differences between the MOL and EOL stages of core lifetime for the LEU core are the core power peaking factors in addition to the radial and axial power profiles shown in section 4.7.6. UWNR LEU Conversion Analysis 122 August 2008
The radial temperature is plotted in Figure 4.7.52 given with uncertainty of 50% for the gap size. For all analysis afterwards it is assumed the gap size between the fuel and the cladding is 0. 10 mil (0.00254mm). The axial temperature profile is shown in Figure 4.7.53 and a summary table of steady state results for 1.0, 1.3, and 1.5 MW are shown in Table 4.7.17. In addition, the temperature profile and the coolant flow rate are shown as a function of power in Figure 4.7.54 and Figure 4.7.55 respectively. Radial Temperature Distribution of Hot Rod at 1.5 MW (LEU EOL) Radial Distance from Fuel Centerline [in] 0.0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 800 -- -- -_--_-- ___ _ 1412 700 ,,. ,, , 700 - A m n n A AA 1212 U600 *
- A, -L v.A.. n I AA 1012 500 . ......
. A- A
- A UnoA 812 0.
~,4001e .'-
A. E 300 612 -- 200 412
*0.05 mils w 0.10mils A 0.15 mils 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 Radial Distance from Fuel Centerline [cm]
Figure4.7,52 Radial temperature distributionof hot rod at 1.5 M.V for varying gap sizes (LEU EOL) 123 August 2008 (:ollversion Analysis UWNR LEU Conversion Analysis 123 August 2008
Table 4.7. 17 Steadv state results,of hot rod at 1-5, 123. and 1.0 MW (LEU EOL) Parameter Core Power SI English _________________________________ IMWI 1.5 28.328 Rod Power [kW] 1.3 24.5,51 1.0 18.885 1.5 0.14603 0.32195 Flow Rate for Hottest Channel [kg/s], [Ibm/s] 1.3 0.12876 0.28387 1.0 0.10278 0.22659 1.5 0.36307 1.19117 Maximum Flow Velocity [m/s], [ft/s] 1.3 0.29651 0.97280 1.0 0.22702 0.74482 1.5 860.07 Maximum Wall Heat Flux [kW/m 2] .1.3 746.13 1.0 573.95 1.5 641.91 1187.44 Maximum Fuel Centerline Temperature [°C], ['F] 1.3 575.92 1068.66 1.0 475.43 887.87 1.5 140.48 284.86 Maximum Outer Clad Temperature [°C], ['F] 1.3 138.61 281.50 1.0 135.42 275.76 1.5 127.69 261.84 Exit. Outer Clad Temperature [°.C], ['F] 1.3 127.38 261.28 1.0 125.33 257.59 1.5 100.67 213.21 Exit Bulk Coolant Temperature [°C], ['F] 1.3 99.84 211.71 1.0 98.22 208.80 UWNR LEU Conversion Analysis 124 August 2008
Axial Temperature Profile of Hot Rod at 1.5 MW for LEU EOL Axial Distance from Bottom of Fuel [in] 0 2 4 6 8 10 12 14 700 .
- Centerline - 1232 600
- Outer Cladding - 1032 500 ----
- Bulk Water -832 a, 400 .
E' 4.*
.......... * .. -632 a-300 . . . .
432 200 . . A, . 232 100 A -AA A-A A A A AA A 0 32 0 5 10 is 20 2S 30 3S 40 Axial Distance from Bottom of Fuel [cm] Figure 4. 7,53 Axial temperatureprofile oi'hot rod at 1.5 MW (LEU EOL) Temperature Profile of Hot Rod vs. Power (LEU EOL)
.OMW 1.3MW 1.5MW Soo 1432
- Max Centerline Temp M* nOuter Cladding Tempn Max
- 1232 2
600 e A Max Bulk Coolant Temp
. 1032 a.
03 500 832 CL 400 632 CL E 30 250 E 0.W 200 432 200 &-A 0 .. I....L. AAAA
.. . ,,I,., , *.- * . 232 32 0 5 10 15 20 25 30 35 Power in Hot Rod [kW]
Figure 4. 7.54 Temperatureprojile ol hot rod at L)5 SW v,. power 0/hot rod (LOEU EOL) 125 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 125 August 2008
0 Coolant Flow Rate of Hot Rod vs. Power (LEU EOL) 1.0MW 1.3 MW 1.5 MW 0.25
- RELAP Calculated Flow 0.5 Rate 0.2 i. i
- -- Extrapolated Region S. ___ ___ 0.4 E (U 0.15 0.3
- 0 .4 cc 0 0.1 0.2 0 0 0.05 *,* * . . ... ........... 01 0 U
0 0 0 5 10 15 20 25 30 35 40 45 Power in Hot Rod [kW] Figure4.7.55 Coolantflow rate.of hot rod at.D5 SW vs.power of hoi rod (LEUIFOL) As can be seen from Figure 4.7.55, the flow rate at 28 kW/rod oscillated according to RELAP5/MOD 3.3, so a projection was formed as seen by the dashed line. After determining the steady state temperatures and flow rates, the steady state results were found for core powers of 1.0, 1.3, and 1.5 MW shown in Table 4.7.18. The MDNBR as a function of rod power was plotted in-Figure 4.7.56 and the power to which CHF would be achieved as a function of flow rate is shown in Figure 4.7.57. 126 August 2008 UWNR LW Conversion Analysis LEU Conversion Analysis 126 August 2008
Table 4.7.18 Steady state CtHFresults of hot rod for core powers of !.5. 1.3, and 1.0 MW (LEU EOL) CHF Parameter Core Power Groneveld Bernath IMWI 2006 Power/rod to reach DNB of 1.00 [kW] 53.565 34.694 1.5 1.891 1.225 MDNBR 1.3 2.226 1.350 1.0 2.717 1.542 MDNBR vs. Power of Hot Rod (LEU EOL) 12 1.0MW 1.3 MW 1.5 MW 10
- Groeneveld 2006 C Bernath Z 6 .. ....... . . . . . . . . . .
Sa-- -- DNBR=1.0 4U
. .. .. . .. __ U 4 - .. .. . .. .... .. . . . . . . .. ... . ..... .. .. . .......
20 0 r---... l-. . r. ............. ------ 0 5 10 15 20 25 30 35 40 45 Power in Hot Rod [kW] Figure 4.7.56 MDNBR as a finction of hot rodpower (LEU EOL) Analysis 127 August 2008 UWNR LEU Conversion Analysis 127 August 2008
Flow Rate vs. Power of Hot Rod:(LEU EOL) 1.0MW 1.3 MW 1.5 MW 70 ~ T -_____________ bO S50 0 (Z 40 0 4-30 0
- Groeneveld 2006 0 -Bernath 10o . ' ____ -- . - RLP.£*Falculated.fIo 0~
0 ..................................................... ,.........~....... . -- -- Extrapolated Region 0,04 0.09 0.14 0.19 0.24 Flow Rate [kg/si Figure 4.7.57 Power to reach C-IF as aflnction oJflow rate (IYU EOL) With slightly smaller power levels in the hot rod in the LEU EOL core, the margin to CHF is slightly higher than CHF margin in the hot rod in the LEU MOL core. Since the flow rate began to oscillate around 28 kW/rod, the rod power at which CHF would result was calculated at the last known, stable flow rate of 0. 13994 kg/s (0.30851 lbm/s). Thus, the power necessary to reach CHF at LEU EOL would be 53.565 kW/rod and 34.694 kW/rod for the Groeneveld 2006 and Bernath correlations respectively for all rod powers above 27 kW/rod. The power necessary to reach CHF calculated for LEU EOL is slightly higher than the power necessary to reach CHF in the LEU MOL analyzed in section 4.7.8. UWNR LEU Conversion Analysis
. . ... . . . . . ............... j ....
128 August 2008 0
4.7.10 LEU Core Pulse Analysis Using the same pulsing methodology as stated in section 4.7,5 the predicted pulse power and maximum fuel temperature can be calculated for e'ach stage of core life. Since there are now a total of 83 rods in the LEU core as opposed to 91 rods in the HEU core, the core channel of the two channel model was changed to reflect this difference. The negative fuel temperature coefficient for LEU BOL, MOL and EOL are shown in Table 4.7.19 through Table 4.7.21 as calculated by MCNP5. These tables can also be seen in section 4.5.2 where the negative fuel temperature coefficients are shown in Figure 4.5.1.7. The delayed neutron fraction P was calculated from MCNP5 to be 0.00782 +/- 0.000134 for BOL, 0.00774 +/- 0.000148 for MOL, and 0.007389 +/- 0.00017 for EOL. These results were used to calculate the following plots and tables for each stage of core life:
- 1. Maximum fuel temperature and the maximum fuel temperature in the thermocouple location in the hot rod for a 1.4%Akik pulse as a function of time
- 2. The pulse power and total energy of a 1.4%Ak/k pulse plotted as a function of time
- 3. Fuel temperature surface plot at the time of maximum temperature for a 1.4%Ak/k pulse
- 4. Radial and axial fuel temperature plots of hot rod at the time of maximum fuel temperature for a 1.4%Ak/k pulse
- 5. Maximum fuel temperature plot as a function of pulse size in the hot rod
- 6. Maximum pulse power and total energy after 0.25 seconds from initiation of pulse as a function of pulse size
- 7. Summary table of pulse conditions for each pulse analyzed 129 August 2008 UVVNR Conversion Analysis UWNR LEU Conversion Analysis 129 August 2008
Table 4.7.19 Fuel temperaturecoefficient for LEU BOL 0 Temperature Fuel Temperature Coefficient Endpoint Midpoint Total 10C [°Cl,___([°F_ J,(1FJ) 10C1, (1F1) keff __)_ Akeff _I_______(___F1) Ap I$1 Is' W orth 26.85 (80.33) 1.00176 0 76.85 (170.3) 0.00408 .0.00421
-0.52230 126.85 (260.3) 1.01340 -0.52230 226.85 (440.3) 0.01333 0.01316 -1.68233 326.85 (620.3) 1.00007 -2.20463 426.85 (800.3) 0.01614 0.01640 -2.09765 526.85 (980.3) 0.98393 -4.30228 726.85 (1340) 0.04913 0.05342 -6.83064 926.85 (1700) 0.93480 -1I.1329 Table 4.7.20 Fuel temperaturecoefficient for LEU MOL 0
Temperature Fuel Temperature Coefficient Total Endpoint Midpoint kerr Akefr Ap I$1 Worth 1°Cl, (I°FI) °CI' (1°F1) .$1 26.85 (80.33) 1.00156 0
--. ' 76.85 (170 3) ,!_- ____ 0.00335 1 0 00326,, -0 42.1404, 126.85 (260.3) 1.01221 -0.42104
,_._-_ _..." " 226.85 (440.3) 0.01145 0.01130 -1.46037 326.85 (620.3) 1.00076 -1.88142 426.85 (800.3) .. 0.01333 0:0,1349 J-.74282.
526.85 (980.3) 0.98743 -3.62424
, 726,.85(1340), .__-__ 0..041.2 Q4401' -5:68555 926.85 (1700) . 0.94631 -9.30979 IJWNR LEU Conversion Analysis UWNR Analysis 130 1.30 August 2008 August 2008
-O4
Table 4.7.21 Fuel temperaturecoefficientfor LEU EOL Temperature Fuel Temperature Coefficient Endpoint Midpoint Total 10Cr kff Akeff Ap 1$1 Worth I IF_).l __I_' (I°FI)_1 26.85 (80.33) 1.00146 0 76.85 (170.3) 0.00313 0.00305 -0.41281 126.85 (260.3) 1.01142 -0.41281 226.85 (440.3) 0.01078 0.01065 -1.44153 326.85 (620.3) 1.00064 -1.85434
- --t426.85 (800.3) 0..041160 0.0.1172 -1.58628 526.85 (980.3) 0.98904 -3.44063 926.85 (1700 726.85 (1340) 0.95307 0.03597 0.03816 -5.16436 926.85 (I 700) .0.95307 -8.60498 The following plots and table are for analysis performed for LEU BOL. The maximum fuel temperature is 663.12'C (1225.62°F) and occurs 0.0655 seconds after initiation of a 1.4%Ak/k pulse producing a peak pulse, power of 2.06 GW. The fuel temperature and power as a function of time are shown in Figure 4.7.58 and Figure 4.7.59 respectively. The fuel temperature surface plot at 0.0655 seconds, and axial and radial temperature profiles are shown in Figure 4.7.60 through Figure 4.7.62.
The maximum fuel temperature as a function of pulse size is shown in Figure 4.7.63. The maximum pulse power and the total energy of a pulse after 0.25 seconds are shown in Figure 4.7.64. According to the RELAP5/MOD3.3 model a pulse greater that 2.1%Ak/k pulse would be U\VNR LEU Conversion Analysis 131 August 2008
required to exceed the fuel temperature safety limitof 1150ýC (2100-F). A summary table of the pulse conditions produced as a function of pulse size i's shown in Table 4.7.22. 0 Maximum Temperature of Hot Rod after 1.4% Ak/k pulse (LEU BOL) 1200 2032 1000 1532 U7 800 663.12 CL 600 1032 0. E E C, 400 03
. -Max Hot Rod Thermocouple 532 200
- - Operating Limit= 830 °C
- - Safety Analysis Limit = 1150 °C 0 32 0 0.05 0.1 0.15 0.2 0.25 Time (s]
Fig*ure 4. 7.58 Maximum fuel temperatureof hot rod after 1.4 %Alk/k pulse (LEU BOL) 132 August 2008 LJWNR Conversion Analysis UWNR LEU Conversion Analysis 132 August 2008
Power and Total Energy of 1.4% Ak/k Pulse (LEU BOL) 1.E+10 - 1.E+09 2.06E+09 S.E+09 - 1.E+08 1.E+08 - 1.E+0.7 1.E 7 1.E+o76 1 1.E+06 O 1.E+06 e I.E+05 1.E+05 - I..E+,0 . - Power 1.E04
-- Energy 1.E+03 1.E+03 0 0.05 0.1 0.15 0.2 0*25 Time [s]
Figure4. 7.59 Power and total energy of 1.4 I-lk/l pulse (LEU BOL) 40 I . . . . . . . . ..... .... 600 35 E 25 o 400 20 E 300
'5 O)
- 100 0 1 2 Radial Distance from Fuel Centerline [cm]
Figure 4.7.60 Temperature s'urface /lot of'hot rod at 0. 0655s after I. 4%Ak/k pulse (ILEU BOL) 133 August 2008 UWNR LEU Conversion Analysis Analysis 133 August 2008
Axial Temperature Distribution of hot rod at 0.0655 seconds after 1.4% Ak/k pulse (LEU BOL) Axial Distance from Bottom of Fuel [in] 0 2 4 6 8 10 12 14 700 1292 650 70 1292i._ 1192 600 ..... - T _ _-_--_ _. 1092 E.
$50 992 500 892 'V 450 450 9'V 1
400 400 - 7111117 ~~771ZZIZ 792 692 0. E 350 350 692 a, E 9 300 300 592 592 250 250 492 492 200 ý39,2 0 5 10 15 20 25 30 35 40 Axial'Distance from Bottom of Fuel [cm] Figure 4. 7.61 Axial temperaturedistributionof hot rod after 1.4%oAk/k pulse (LEU BOL) Radial Temperature Distribution of hot rod at 0.0655 seconds after 1.4% Ak/k pulse (LEU BOL) Radial Distance from Fuel Centerline [in] 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 700 -J..... 1232 600 1032 500-832
- 400 ____ ~A * * * ,
632 300 E 200 432 E a, 1oo0 4 232 0 +---r---T-,-------- -r - , 32 0 0.2 0.4 0.6 0.8 1 1.2 1,4 1.6 1.8 Radial Distance from'Bottom of Fuelr[cm] FiiýUre 4.7.62 Radialtemperaturedistribution of hot rod aqfer 1.4%klk/k pulse (LEU BOL) 134 August 2008 UWNR LEU Analysis Conversion Analysis LEI.J Conversion 134 August 2008
. MaximumPulse Temperature vs. Pulse Size (LEU BOL) Pulse Size [$] 1.3 1.S 1.7 1.9 2.1 2.3 2.5 2.7 1200 2172 1100 1972
- o. 900 1 -1772 Z goo 1572
-, 800 700 1372
- 700) 1172 C E 600 .
- Max FuelTemp E
500 972
-~--- Safety Analysis Limit 400 ~ ---- ------------ - ------------- ------------- - ------- - SafetyA. alysis-Li -i 772
- - Operating Limit 300 ......... 572 1 1.2 1.4 1.6 1.8 2 Pulse Size [%Ak/k]
Figure 4. 7.63 Maximum pulse tenmperature as a.finction o/pulse size (l.EU BOL) Maximum Power and Total Energy vs. Pulse Size (LEU BOL) Pulse Size [$] L3 1.5 1.7 1.9 2.1 2.3 2.5 2.7 9 60 51 ... 50 a 7 a U £ 6 U 40
- 5 U 30
- 0 ________
4 0 U 0 20
- 13. 3 U
2 Total Energy
- Max Power 10 1
0 0 1.2 1.4 1.6 1.8 2 Pulse Size [%Ak/k] Figure 4. 7.64 Maximum power and total eneryv as a function of pulse size (LEU BOL) 135 August 2008 UWNR LEU Conversion Analysis LEU Conversion An.alysis 135 August 2008
Table 4.7.22 Sumnmary'of pulse conditions as a function ofpulse size (LEU BOL) Pulse Size Maximum Fuel Maximum Total Energy
= 0.00782 +-0.000134 Temperature Pulse Power Released Ak/k $ °F OC GW MJ 1 1.279 345.19 653.34 0.289 11.703 1.1 1.407 430.04 806.07 0,607 15.694 1.2 1.535 512.05 953.69 1.013 19.459 1.3 1.662 589.30 1092.74 1.500 23.164 1.4 1.790 663.12 1225.62 2.064 26.866 1.5 1.918 734.35 1353.83 2.716 30.593 1.6 2.046 803. i 5 1477.67 3.446 34.285 1.7 2,174- 867.95 1594.31 4.253 37.863 1.8 2.302 929.15 1704.47 5.153 . 41.388 1.9 2.430 987.85 1810. ý1.3 6.143 44.888 2.0 2,558 1044.45 1912.01 7.209 48.377 2.1 2.685 1099.35 2010.83 8.403 51.871 0
The following plots and table are for analysis performed for LEU MOL. The maximum fuel temperature is 726.95TC (1340.5 1F) and occurs 0.065 seconds after initiation of a 1.4%Ak/k pulse producing a peak pulse power of 2.52 GW. The fuel temperature and power plots as a function, of time are shown in Figure 4.7.65 and. Figure 4.7.66 respectively. The fuel temperature surface plot at 0.065 seconds, and axial and radial, temperature profiles are shown in Figure 4.7.67 through Figure 4.7;69. The maximum fuel temperature as a function of pulse size is shown in Figure 4.7.70. The maximum pulse power and the total energy of a pulse after 0.25 seconds are shown in Figure UWNR U"R, LEU Conversion LEU Analysis Conversion Analysis 136 136 August 2008 August 2008 0
4.7.71. According to the RELAP5/MOD3.3 model a pulse `greater that 2:O%Ak/k pulse would be required to exceed the fuel temperature safety limit of I I 500 C (21 00°F). A summary table of the pulse conditions produced as a function of pulse size is shown in Table 4.7.23. Maximum Temperature of Hot Rod after 1.4% Ak/k pulse (LEU MOL) 1200 F- - - 2032 1000
*' 800 1532
- 800 726.95 1
, o1032 ',
M C E406 4- Max Temp 2- Max Hot Rod Thermocouple 532 2- - Operating Limit =830 'C 0 .. . ............---- safety Analysis Ltimt 1150LC 3 0.05 . 0.1 0.15 0.2 0.25 Time [s] Figure4.7.65 Marimum fuel temperatureof hot rod after 1.4 %Adk/k pulse (LEU MOL) Analysis 137 August 2008 UWNRLEU Conversion Analysis [JWNR.LEU Conversion 133 August 2008
Power and Total Energy of 1.4% Ak/k Pulse (LEU MOL) 1.E+10 11.E+09 2.52E+09 1.E+09 1.E+08 I.E+08 1.E+07 I.E*07 o 1,E+06 1E+05 S-wower
-Po er 1.E+04 I.E+04
-Energy 1.+03I,1.E+03 0 0.05 0.1 0.15 0.2 0.25 Time [sI
`i,,ure 4.7. 66 Power and total energv of 1.4 %,Ak'k pulse (LEU MOL) 40 . .......
400 . .. . . . . . . 35 T 4-600 30 LL_
's 500 E 25 C
o*20 400 E
'(j3 (13 77z 2 5
100 0 0 1 2 Radial Distance from Fuel Centerline lcmj F'*gure 4.7.6 7 Teniper'turesu 00ce plotI/'ho/ rod at O. 0650s after Ile4; k/k pulse (LU EM)L) UWNR LEU Conversion Analysis 138, August 2008 0
ý 4D Axial Temperature Distribution of hot rod at 0.065 seconds after 1.4%Ak/k pulse (LEU MOL) Axial Distance from Bottom of Fuel [in] 0 2 4 6 8 10 112 14 800 1392 700
+ ....... 1192 600 a, 19r2 CL 500 a,
792 1=0. E 400 E 300 592 200 392 0 5 10 i5 20 2S 30 35 40 Axial Distance from Bottom of Fuel [cm] F'igure 4.7.68 Axial temperaturedistributionof hot rod after 1.4%kik pulse (LEU MOL) Radial Temperature Distribution of hot rod at 0.065 seconds after 1.4%Ak/k pulse (LEU MOL) Radial Distance from Fuel Centerline [in] 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 800 .....
. 1432 700 .. . ... ........... . ... .... . .. ----- --.. .-------......
1232 600 ____- __,____.:_ _._ O2 6001032 832 " M 400 --------- _ _ C 632 0. E 300. - EU 200 _ 432 100 - - --- ---- 232 0 0.2 0.4 0.6 0.8 1 1.2 1.4 16 1.8 Radial Distance from Bottom of Fuel [cm] I-'igure 4:7.69 Radial temperature distributionof hot rod after .4%Alk/k pulse (LEU MOL) 139 August 2008 Conversion Analysis LEU Conversion UWNR LEU Analysis 139 August 2008
Maximum Pulse Temperature vs. Pulse Size (LEU.MOL) Pulse Size [$1 to 1.3 1.5 1.7 1.9 2.1 2.3 2.5 1200 2172
,£
.1100 1972 1000 1772 '
4. 900 1572 800 1372 ' 4.' 700 E~ 1172 C.
- 0. 600 E
- Max Fuel Temp 972 I 500
*- - Safety Analysis Limit 400 772
-- - Operating Limit 300 572 1 1.2 1.4 1.6 1.8 2 Pulse Size [%Ak/k]
Figure 4. 7. 70 Mafi~mum pulse temperature as a finction Qtpulse size (LEU MOL) Maximum Power and Total Energy vs. Pulse Size (LEU MOL) Pulse Size [$} 1.3 1.5 1.7 1.9 2.1 2.3 2.5 9 60 I
.7 40 =7 2.-s 30 >
4
- 0. 3 0 20 wtD 4 T 2
- __
- Total a.__ Enery .,Max Power- 20 10 1
0 1 1.1 1.2 1.3 1.4 1.5 1.6 1.7 1.8 1.9 2 Pulse Size [%Ak/k] F'igure 4. 7. 71 Maximum power and total energy as a function ofpulse size (LEU MIOL) UWNR LEU Analysis Conversion Analysis LEU Conversion 140 140 August 2008 August 2008 0
Table 4.7.23 Sum mary qf pulse conditions as a function of pulse size (LEU MOL) Pulse Size Maximum Fuel Maximum Total Energy
= 0.00774.+/- 0.000148 Temperature Pulse Power Released Ak/k $ °C OF GW M3 1 1.292 385.44 725.79 0.395 14.20 1.] 1.421 476.03 888.85 0.784 18.78 1.2 1.550 563.85 1046.93 1.267 23.17 1.3 1.680 647.11 1196.80 1.844 27.53 1.4 1.809 726.95 .1340.51 2.520 31.91 1.5 1.938 803.45 1478.21 3.294 36.20 1.6 2.067 875.25 1607.45 4.168 401.41 1.7 2.196 943.55 1730.39 5.143 44.59 1.8 2.326 1009.15 1848.47 6.222 48.77 1.9 2.455 1072.75 1962.95 7.399 52.96 2 2.584 1134.55 2074.19 8.698 57.18 0 .The following plots and. table are for analysis performed for the LEU core at EOL. The maximum fuel temperature is 723.49°C (1334.28'F) and occurs 0.595 seconds after initiation of a 1.4%Ak/k pulse producing a peak pulse power of 3.06 GW. The fuel temperature and power plots as a function of time are shown in Figure 4.7.72 and Figure 4.7.73 respectively. The fuel temperature surface plot at 0.595 seconds, and axial and radial temperature profiles are shown in Figure 4.7.74 through Figure 4.7.76 The maximum fuel temperature as a function of pulse size is shown in Figure 4,7.77. The maximum pulse power and the total energy of a pulse after 0.25 seconds are shown in Figure 4.7.78. According to the RELAP5/MOD3.3 model a pulse greater than 2.0%Ak/k would be UWNR LEU Conversion Analysis 141 August 2008
required to exceed the fuel temperature safety limit of 1150°C (2100'F). A summary table of the pulse conditions produced as a function of pulse sizeis shown in Table 4.7.24. Maximum Temperature of Hot Rod after 1.4% Ak/k pulse (LEU EOL) 1200 .... 2032 1000 800 -- - - -1532...
~.J 800 723.49 - "__ _
h.. . .. I 600 -032 1032 E G. 400 ____ E
- Max Temp
- Max Hot Rod Thermocouple 532 200 - - Operating Limit = 830'C 0 -- Safety Analysis Limit = 1150 'C 32 .
0 0.05 0.1 0.15 0.2 0.25 Time (s] Figure 4.7. 72 Maximum fuel temperatureof hot rod after 1.4 %Ik~k pulse (LEU EOL) UWNR LEU Conversion Analysis J 142 August 2008
Power and Total Energy of 1.4% Ak/k Pulse (LEU EOL) I.E+10 1.E+09 106E+09 1.E+09 1.E+08 1.E+08 1.E+07 O1E+07 1.E+06 1 .E+06 a E+05E+05 Power P.E+04 11+04
- Energy I.E+03 -U 1,E+03 0 0.05 0.1 0.15 0.2 0.25 Time [s]
F'igure 4.7. 73 Power and total energv of .4 %dk/k pulse (LO.U EOL) 40 ~700 35 E .00 30 0500 E 20 20 400 1'5 300 75 1 -* 200 0 0 1 2 Radial Distance from Fuel Centerline (cm] F'igure 4. 7:74 Temperature surfice plot of hot rod at 0.0595s qfier l.4%Atk/k pulse (LEU EOL) UWNR LEU Conversion Analysis *143) August 2008
Axial Temperature Distribution of hot rod at 0.0595 seconds after 1.4% Ak/k pulse.(LEU EOL) Axial Distance from Bottom of Fuel [in] 0 2 4 6 8 10 12 14 800 1392 700 1192 600 I * ' __%_* 9 60 - 992 4) Soo a) 400 792 400....... . ........ .. T.... .. . . . .E E E) 4, I-300 592 300 S9 200 392 0 5 10 15 20 25 30 35 40 Axial Distance from Bottom of Fuel [cm] Figure 4.7.75 Axial temperaturedistributionof hot rod qafer 1.4%lk/k pulse (LEU EOL) Radial Temperature Distribution-of hot rod at 0.0595 seconds after 1.4% Ak/k pulse (LEU EOL) Radial Distance from Fuel Centerline [in] 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 800 1432
.9 700 9 -
9 1232 600 1032 5oo ,.,.......... ~ ~ . .. CL 832 r: 400 632 300
- 4) E 432 4) 200 100 232
.-........... I............. "-. ... .... .. n... .... .............
... ....'.. T ... - _ . . .
0* 32 0 0.2 0 .4 0.6 0.8 1 1.2 1.4 1.6 1.8 Radial Distance from Bottom of Fuel [cm] Figure 4.7.76 Radial fernperaturedistribution of hot rodafter 1.4%A klk pulse (LEU EOL) 144 August 2008 UWNR LEU Conversion Analysis LEIJ Conversion Analysis 144 August 2008
Maximum Pulse Temperature vs.,Pulse Size (LEU EOL) Pulse Size [$] 1.4 1,6 1.8 2.0 2.2 2.4 2.6 1200 -- 1 . 1 2172 1100 1972 1000 1772 " "'900 .................. ........... ............................................
. . . . . . . . ..... ............. .. 1772..........
P S- -1572 o
= 8008.
- 00. . . .. . . . . .
*,700 1372 'M - *1172 C E 600
- Max Fuel Temp 972 E
500 400-
- Safety Analysis Limit
- - Operating Limit 77 300 r ,572 1 1.2 1.4 1.6 1.8 2 Pulse Size [%Ak/k]
F'iuure 4.7 77 Maximum pulse temperatureas aihnction of pulse size (LEU EOL) Maximum Power and Total Energy vs. Pulse Size (LEU EOL) Pulse Size [$} 1.4 1.6 1.8 2.0 2.2 2.4 2.6 10 .. . . ..--- - -*. . . -: - = . . . . 70 ___ 60 7 -__- __50 4 U 30 3 o l U 20 2 w Total Energy
- Max Power io 1 1,1 1.2 1.3 1.4 1.5 1.6 1.7 1,8 1.9 2 Pulse Size [%Ak/kl Figure 4. 7. 78 Waximum power and total energy as ajUnction qf pulse size (EI-U 1-70/.)
UWNR LEU Conversion Analysis 145 August 200i8
Table 4.7.24 Summary of pulse conditions as afiinctioh ofpulse~size (LEU EOL) Pulse Size Maximum Fuel Maximum Total Energy 0.007389 +/- 0.00017 Temperature Pulse Power Released Ak/k $ OC OF GW Mi 1.353 401.51 754.72 0.578 16.606 1,1 1.489 485.50 905.90 1.039' 21.318 12 1.624 568.28 1054.90 1.606 25.938 1.3 "1.759 647,38 1197.28 2,276 30.563 1.4 1.895 723.49 1334.28 3.056 35.180 1.5 2.030 796.05 1464.89 3.945 39.764 1.6 2.165 865.65 1590,17 4.940 44.358 1.7 2.301 933.05 1711.49 6.072 48.984 1.8 2,436 998.55 1829,39 7.310 53.653 1,9 2.571 1062.55 1944.59 8.616 58.373 2 2.707 1124.95 2056.91 10.122 63.002 Results presented in Table 4.7.22 through Table 4.7.24 show -that MOL isthe most limiting stage of reactor life by the maximum fuel temperature predicted by a particular pulse size. At no time will a 1.4%Ak/k pulse ever exceed the 830'C (1526°F) fuel temperature operational limit or the 1150'C (2100'F) fuel temperature safety limit. Therefore keeping the current technical specification limit of a maximum pulse of 1.4%Ak/k will provide sufficient margin fromboth the fuel and operational temperature limits.' Conversion from HEU to LEU fuel will not significantly alter the maximum fuel temperature after a pulse. 146 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 146 August 2008
O 5 REACTOR COOLANT SYSTEMS There are no changes due to the LEU conversion. 147 August 2008 UWNR U"R LEU Conversion Analysis LEU Conversion Analysis 147 August 2008
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,nalysis 148 August 2008 0
6 ENGINEERED SAFETY FEATURES There are no changes due to the LEU conversion. 149 August 2008 UWNR LEU Conversion Analysis LEU Conversion Anaiysis 149 August 2008
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- 7. INSTRUMENTATION AND CONTROL SYSTEMS There are no changes due to the LEU conversion.
UWNR LEU Conversion Analysis 151 August 2008
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O 8 ELECTRICAL POWER SYSTEMS There are no changes due to the LEU conversion. UWNR LEU Conversion Analysis 153 August 2008
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9 AUXILIARY SYSTEMS There are no changes due to the LEU conversion. 155 August 2008 UWNR LEU U"R Conversion Analysis LEU Conversion Analysis 155 August 2008
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10 EXPERIMENTAL FACILITIES AND UTILIZATION 10.1 Summary Description The conversion to LEU fuel is not expected to significantly impact any experimental facilities. 10.2 Experimental Facilities The reactivity effects of flooding the beam ports and pneumatic tube, as well as sending a cadmium sample in the pneumatic tube and whale system, have been modeled with MCNP5 for the LEU core. These are compared to previously reported values from the HEU SAR. Table 10.2,1 Experimental FacilityReactivity Effects Condition Previous HEU LEU Calculated % SARW Reported % Ak/k Ak/k Flooding all 4 beam ports + 0.0005 + 0.06 +/- 0.008 Flooding pneumatic tube + 0.0002 + 0.007 +/- 0.005 Cadmium in pneumatic tube - 0.0005 - 0.009 +/- 0.006 Cadmium in most reactive whale tube (C8) N/A - 0.04 +/- 0.006 UWNR LEU Conversion Analysis 157 August 2008
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11 RADIATION PROTECTION PROGRAM AND WASTE MANAGEMENT There are no changes due to the LEU conversion. UWNR LEU Conversion Analysis 159 August 2008
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12 CONDUCT OF OPERATIONS 12.1 Organization and Staff Qualifications The conversion to LEU fuel will not impact the existing UWrNR organization and staff qualifications. 12.2 Procedures The UWNR procedure for disassembly and reassembly of a fuel bundle will be updated to use currently available tools. New procedures will be written for receipt of new LEU fuel and shipment of Standard and FLIP fuel. No other procedure changes are anticipated. All procedure changes and additions will be approved according to existing Technical Specification 6.5, and will not be included with this LEU conversion report. 12.3 Operator Training and Requalification The existing operator training and requalification program at UWNR will not be impacted by the conversion to LEU fuel. The content of the training material will be revised under theexisting program to reflect changes to LEU fuel. Operators will be trained on changes to the SAR and the procedures before the conversion takes place. Operators will be trained on new procedures for fuel receipt and shipment before these actions are carried out. 161 August 2008 UWNR UWNR LEU Conversion Analysis LEU Conversion Analysis 161 August 2008
12.4 Emergency Plan Changes to the UWNR Emergency Plan will be made in accordance with 10 CFR 50.54(q), and will not be included with this LEU conversion report. 12.5 Physical Security Plan No changes to the UWNR security plan are required as part of the conversion to LEU fuel. However, if any changes are required, they will be submitted in accordance with 10 CFR 50.54. 12.6 Quality Assurance Close communication between University of Wisconsin Nuclear Reactor, General Atomics, and the Idaho National Laboratory is maintained through the safety analysis development and reactor component design process. This communication ensures that the appropriate technical and functional requirements from the reactor safety basis are carried forward through the design, and fabrication of the new reactor components. UWNR staff makes certain that the drawings and fabrication specifications indicate the correct dimensions and other design parameters that must be met for each item to be in compliance with the reactor's safety basis. These documents are used to definitively communicate each component's requirements to UWNR personnel, INL procurement personnel, General Atomic personnel, and the QA Engineers from each organization. UWNR staff then relies on the well established QA processes at the INL to ensure that the final product meets the requirements per the drawings and specifications. UWNR LEU Conversion Analysis 162 August 2008
- The INL has a Quality Assurance Program that meets or exceeds the requirements for procuring items and services as established by the DOE. The INL has audited CERCA as required to ensure that their QA program meets or exceeds the requirements of NQA-I. These requirements are contained in 10 CFR 830 Subpart A, Quality Assurance Requirements, DOE Order 414.1C, Quality Assurance, and NQA- 1-2000, Quality Assurance Requirements for Nuclear Facility Applications. These requirements establish the methods that the INL, and their suppliers, must use to procure material and fabricate items. More specifically, the process requirements relevant to the procurement of saflety related items for the UWNR are:
" Items and services shall be procured to meet established requirements and perform as specified. [DOE Order 414.1 C, Atitachment 2, 3.g.(1)] [10 CFR 830.122(g)(1)]
- Prospective suppliers shall be evaluated and selected on the basis of specified criteria..
[DOE Order 414.1 C, Attachment 2, 3.g(2)] [10 CFR 830.122(g)(2)]
" Processes shall be established and implemented to assure that approved suppliers continue to provide acceptable items and services. [DOE Order 414.1C, Attachment 2, 3.g(3)] [10 CFR 830.122(g)(3)]
To complete the process, source inspections of the items are performed by qualified TNL Inspectors before being released to the UWNR, after which, receipt inspections of the items are performed by UWNR personnel. 163 August 2008 UWNR LEU U"R Conversion Analysis LEU Conversion Analysis 163 August 2008
12.7 Reactor Reload and Startup Plan The UWNR was reloaded with TRIGA Standard LEU fuel in 1967, then partially refueled with FLIP fuel in 1974, 1978, then finally with all FLIP fuel in 1979. The Reload and Startup plan for the new LEU core will be modeled after the successful startup plans for these previous cores. The startup report will be submitted in accordance with NUREG-1537 Chapter 12.6 and Technical Specification 6.7.1. The UWNR Reload & Startup plan will incorporate the following steps:
" Remove all HEU fuel bundles from the core grid-box and place in the fuel storage pit before any loading of LEU fuel. Remove the HEU instrumented fuel elements from the core grid-box and store near the storage pit away from the core grid-box.
- Prior to loading any LEU fuel in the reactor core, install a neutron source and 2 fission counters in the core and verify they are operational. Install the fission counters on opposite sides of the core. Remove the rod drives for blades 1-3, leaving the control blades installed in the core. Leave the regulating blade in the fully inserted position.
- All fuel additions shall be made with an operator in the control room monitoring the count rate. This control room operator shall be in constant communication with the fuel handler and supervisor.
" Load the first fuel bundle with transient rod guide tube in the core, then install all 4 scrammable control element drives. Perform scram-time testing and.verify all drives are operational before proceeding with any further fuel additions.
164 August 2008 UWNR LEU U"R Conversion Analysis LEU Conversion Analysis 164 August 2008
- Monitor the criticality of the new core using standard 1/M plots where M is the ratio of the count rate with n fuel bundles in the core divided by the initial count rate. Obtain the count rate from both of the installed fission counters via counter-scalers in the control room. Obtain count rates for the 1/M plot with all 5 control elements fully withdrawn.
- Obtain the initial count rate with the first center fuel bundle and neutron source installed, with all 5 control elements (3 scrammable blades, I scrammable transient rod, 1 non-scrammable blade) installed and fully withdrawn.
- Make all further fuel bundle additions with 2 scrammable control elements half withdrawn and all other control elements fully inserted. In the event of unanticipated criticality, the operator in the control room shall have the ability to scram these 2 control elements.
a Load additional fuel bundles one at a time, updating the I/M plot after each addition (with all control elements fully withdrawn). Add fuel bundles in a symmetric pattern, building the reactor core from the center out. IFE bundles will be loaded and the IFE then installed before updating the 1/M plot for that bundle.
- After criticality has just been reached, calibrate all control elements using the rising period rod bump method. Based on the control element integrated worth curves, calculate the core excess reactivity and shutdown margin. Perform a rough measurement of shutdown margin using the rod drop method.
- Continue adding fuel bundles one at a time until reaching the operational core. Perform a rough measurement of shutdown margin using the rod drop method after each fuel bundle addition after achieving criticality.
UW'NR LEU Conversion Analysis 165 August 2008
- After loading to the operational core, calibrate all control elements using the rising period rod bump method and calculate core excess reactivity and shutdown margin.
- Calibrate the fuel temperature channel.
- Perform the regular pre-startup checklist procedure.
" Perform a power calibration at or near 500 kW. " Perform an initial increase to full power in stepwise increments. Log, all fuel temperatures and control element positions at each power level.
- Repeat the stepwise increase to full power and re-record all temperatures and control element positions. Due to clad deformation during the initial increase to full power and the resultant fuel to clad gap change, the fuel temperatures and reactivity may change after initial bum-in.
" Calculate the power coefficient of reactivity and fuel temperature coefficient of reactivity. " Perform a power calibration at 1000 kW.
- Perform pulse mode operational tests. Initial pulse will be 1.00% Ak/k reactivity insertion, and each successive pulse will be in 0.05% Ak/k increments until the maximum pulse of near, but not to exceed, 1.40% Ak/k reactivity insertion has been reached. Peak power, peak fuel temperature, and integrated pulse powerwill be recorded for each pulse.
Graphs will be prepared for peak fuel temperature vs. reactivity insertion, integrated pulse power vs. reactivity insertion, and peak power vs. the square of reactivity insertion. All 3 graphs will be studied for linearity. UWNR LEU Conversion Analysis ./ 166 Auvust 2008
12.7.1 Acceptance Criteria If, during the reload and startup plan, any of the following acceptance criteria are not met, reactor operation shall be suspended. Reactor operation may be resumed once the Reactor Director and Reactor Supervisor agree on and implement (as appropriate) modifications to the reactor reload and test plan. 12.7.1.1 Initial Criticality MCNP modeling of the LEU core was used to predict critical loading with LEU fuel bundles. The predicted critical loading is 18 bundles if all 14 graphite reflectors are removed, or 16 bundles if all 14 graphite reflectors are installed. The acceptance criteria for initial criticality will be +/- 2 fuel bundles to account for inaccuracies in the MCNP model. 12.7.1.2 Control Element Worth by Rod Drop Method After achieving criticality, control element worth is estimated by the rod drop method after each fuel bundle addition. The calculated shutdown margin shall meet or exceed the Technical Specification limit. 12.7.1.3 Control Element Worth by Rising Period Rod Bump Method Control element worth for the first critical loading and operational core is measured by the rising period rod bump method. The calculated shutdown margin shall meet or exceed the Technical Specification limit. 167 August 2008 UWt'.JR LEU U"R Conversion Analysis LEU Conversion Analysis 167 August 2008
12.7.1.4 Shutdown Margin and Excess Reactivity Excess reactivity is the amount of positive reactivity the core would have if all control elements were withdrawn in the cold clean condition. Shutdown margin is the amount of negative reactivity in the core with the highest worth element j(anticipated to be blade 3) and the regulating blade (non-scrammable) fully withdrawn and all other elements fully inserted. Excess reactivity and shutdown margin shall be calculated for the operational core and must be within Technical Specification limits. Note that under the current HEU SAR, there is no Technical Specification for excess reactivity. 12.7.1.5 Power and Temperature Coefficient The power coefficient is the slope of the power vs. reactivity curve. The temperature coefficient is theslope of the fuel temperature vs. reactivity curve. Power and temperature coefficients will be determined over several different power intervals during the initial increase to full power. The power and fuel temperature coefficients must be negative over all operating ranges. 12.7.1.6 Pulse Mode Operational Testing The graphs of peak fuel temperature vs. reactivity insertion, integrated pulse power vs. reactivity insertion, and peak power vs. the square of reactivity insertion must show a linear dependence. However, some departure from linearity at higher reactivity insertion values is acceptable and may be observed due to longer transient rod ejection times. UJWNR 1 EIT Conversinn Analvsis 168 ivw Aupiust 2008l
O 13 ACCIDENT ANALYSIS While most results of the accident analysis are not affected by the conversion to LEU fuel, the analysis is being updated to reflect currently approved methodologies. 13.1 Maximum Hypothetical Accident (MHA) The maximum hypothetical accident for UWNR is postulated as damage to a fuel element resulting in failure of the fuel cladding. It is postulated that this damage occurs after continuous operation at 1.3 MW and that it occurs in the fuel element with the highest power density. This power is 22.886 kW for the revised HEU analysis and 25.169/25.037/24.551 kW for the LEU BOL/MOL/EOL analysis. The previous HEU SAR assumed a power of 23 kW. Licensed power is 1.0 MW, and the SCRAM setpoint is at 1.25 MW, thus 1.3 MW represents the maximum S power accounting for uncertainty in the power level. The previous HEU SAR assumed infinite 1.25 MW operations. In the event of a cladding failure, the fuel will corrode and leach into the water at a rate of about "100 micrograms of U-ZrH per square centimeter of exposed fuel surface per day for shutdown conditions . Therefore, only the gaseous and highly volatile fission products that have collected in the space between fuel and cladding would be the activity contributing to personnel hazards. This assumption is unchanged due to the conversion to LEU fuel. UWNR LEU Conversion Analysis 169 August 2008
13.1.1 MHA Fission Product Inventory in Fuel Element 0 The quantity of these volatile and gaseous fission products was determined by the use of the ORIGEN2 computer code version 2.1, using the PWRUS cross-section library. A single fuel pin was simulated at the power level of the hottest rod. BOL was modeled at about 50 MWd J continuous operation because most of the fission products analyzed were near saturation by that time. MOL was modeled at 800 MWd and EOL at 1800 MWd continuous operation. Fuel inventories for the previous HEU SAR, revised HEU analysis, and LEU analyses are listed below. The LEU inventories are generally higher than HEU due to the increased power in the hot rod, which is a result of having fewer fuel bundles in the core. 0 UWNR LEU Conversion Analysis 170 August 2008 0
'V Table 13.1.1 MHA Fission ProductInventory Isotope Previous Revised LEU BOL LEU MOL LEU EOL HEU SAR HEU 'Inventory Inventory Inventory Inventory Inventory (Ci) (Ci) (c (Ci) (Ci Br-82 3.OOOE+01 2.276E-01 6.084E-02 4.854E-01 1.164E+00 Br-83 1.050E+02 1.015E-E+02 1.13 1E+02 1.099E+02 1.066E+02 Br-84 1.940E+02 1.908E+02 2.123E+02 2.052E+02 1.977E+02 Br-85 2.530E+02 2.372E+02 2.634E+02 2.545E+02 2.449E+02 Br-87 6.000E+02 4.082E+02 4.535E+02 4.366E+02 4.181E+02 1-130m 2.000E+02 4.308E-01 8.114E-02 9.580E-01 2.462E+00 1-131 5,630E+02 5.471E+02 6.004E+02 6.112E+02 6.177E+02 1-132 8.550E+02 8.158E+02 9.114E+02 9.076E+02 9.132E+02 1-133 1.282E+03 1.276E+03 1.418E+03 1.405E+03 1.399E+03 1-134 1.554E+03 1.441E+03 1.602E+03 1.582E+03 1,569E+03 1-135 1.185E+03 1.188E+03 1.321E-E+03 1.308E+03 1.302E+03 1-136 6.020E+02 5.787E+02 6.467E+02 6.354E+02 6.273E+02 Kr-83m 1.050E+02 1.01 5E+02 1.130E+02 1.099E+02 .1,06,6E%+/-02 Kr-85m 2,530E+02 2.398E+02 2.662E+02 2.574E+02 2.477E+02 Kr-85 5.1OOE+01 3.536E+00 4.976E-01 5.783E+00 1.196E+01 Kr-87 4.860E+02 4,849E+02 5.383E+02 5.185E+02 4.969E+02 Kr-88 6.990E+02 6.848E+02 7.603E+02 7.324E+02 7.016E+02 Kr-89 8.550E+02 8.691E+02 9.648E+02 9.272E+02 8.855E+02 Xe-131rm 5,000E+00 5.41 5E+00 5.809E+00 6.123E+00 6.021 E+00 Xe-133m 3.1OOE+01 3.732E+01 3.744E+01 4.137E+01 4.153E+01 Xe-133 1.282Et03 1.276E+03 1.344E+03 1.405E+03 1.400E+03 Xe-135m 3.500E+02 2.158E+02 2.394E+02 2.411E+02 2.447E+02 Xe-135 1.243E+03 7.897E+02 9.552E+02 9.294E+02 8.926E+02 Xe-137 1.1 85E+03 1.134E+03 1.260E+03 1.245E+03 1.237E+03 Xe-138 8.940E+02 1.180E+03 1.311E+03 1.286E+03 1.264E+03 13.1.2 MHA Fission Product Release Fraction The release of fission products from UZrH fuel elements has been extensively studied by GA and others. The results of this work indicate that the release of fission product gases into the gap between fuel and cladding is given by the following relationship 23:
171 August 2008 UWNR LEU Conversion Analysis U"R Analysis 171 August 2008
RF =1.5x10- 5 + 3.6x 10 3 el3 xOI Equation 13.1.1 where: RF = release fraction T = maximum fuel temperature (°K) At 1.3 MW, the maximum fuel temperature in the hottest pin is calculated for the revised HEU and LEU analyses in section 4.7.4 of this report and summarized in Table 4.7.7, Table 4.7.12, Table 4.7.15, and Table 4.7.17. This temperature is 575.65 'C for the revised HEU analysis, and 594.40/596.27/575.92 'C for the LEU BOL/MOL/EOL analyses. The previous HEU SAR assumed a maximum temperature of 600 'C. At these temperatures; the release fraction is calculated to be 7.936E-4 for the previous HEU SAR, 5.163E-4 for the revised HEU analysis, and 7.201E-4/7.440E-4/5.188E-4 for the LEU BOL/MOL/EOL analyses. Because of the higher temperatures and release fractions, the resulting doses from the MHA are expected to be higher 0 for the LEU case than HEU. Applying these fractions to the total inventory of the fuel element gives the released activity in the table below. IJWNR LEU Conversion Analysis 172
. ,m August 2008
-- wl* .......
0
Table 13.1.2 MHA Released Fission Product Inventories Isotope Previous Revised LEU BOL LEU MOL LEU EOL HEU SAR HEU Released Released Released Released Released (Ci) (Ci) (Ci) (Ci) (Ci) Br-82 2.381E-02 1.175E-04 4.381E-05 3-611E-04 6.039E-04 Br-83 8.333E-02 5.240E-02 8.145E-02 8.176E-02 5.53 1E-02 Br-84 1.540E-01 9.851E-02 1.529E-01 1.527E-01 1.026E-01 Br-85 2.008E-01 1.225E-01 1.897E-01 1.893E-01 1.271E-01 Br-87 4.761 E-01 2.108E-01 3.266E-01 3.248E-01 2.169E-01 I-130m 1.587E-01 2.224E-04 5.843E-05 7.127E-04 1.277E-03 1-131 4.468E-01 2.825E-01 4.324E-01 4.547E-01 3.205E-01 1-132 6.785E-01 4.212E-01 6.563E-01 6.752E-01 4.738E-01 1-133 1.017E+00 6.588E-01 1.021E+00 1.045E+00 7.258E-01 1-134 1.233E+00 7.440E-01 1.154E+00 1.177E+00 8.140E-01 1-135 9.404E-01 6.134E-01 9.513E-01 9.731E-01 6.755E-01 1-136 4.777E-01 2.988E-01 4.657E-01 4.727E-01 3.255E-01 Kr-83rn 8.333E-02 5.240E-02 8.138E-02 8.176E-02 5.531E-02 Kr-85m 2.008E-01 1.238E-01 1.917E-01 1.915E-01 1.285E-01 Kr-85 4.047E-02 1.826E-03 3.583E-04 4.302E-03 6.205E-03 Kr-87 3.857E-01 2.504E-01 3.877E-01 3.857E-01 2.578E-01 Kr-88 5.547E-0 I 3.536E-01 5.475E:O 1 5.449E-01 3.640E-01 Kr-89 6.785E-01 4.487E-01 6.948E-01 6.898E-01 4.594E-01 Xe-131m 3.968E-03 2.796E-03 4.183E-03 4.555E-03 3.124E-03 Xe-133m 2.460E-02 1.927E-02 2.696E-02 3.078E-02 2.155E-02 Xe-133 1.017E+00 6,588E-01 9.679E-01 1.045E+00 7.264E-01 Xe-135m 2.778E-01 1.1 14E-01 1.724E-01 1.794E-01 1.270E-01 Xe- 135 9.864E-01 4.077E-01 6.879E-01 6.914E-01 4.631 E-0 1 Xe- 137 9.404E-01 5.855E-01 9.074E-0 1 9.262E-01 6.418E-0 1 Xe-138 7.095E-01 6.092E-01 9.441E-01 9.567E-01 6.558E-01 For the purpose of further calculations, it is assumed that:
" All gaseous fission products (Kr, Xe) are released to the room air whether the pool is filled with water or not. " For soluble volatiles (Br, I), calculations assume all activity is absorbed in pool water for calculations of pool water activity.
UWNR LEU Conversion Analysis 173 August 2008
For soluble volatiles (Br, I), calculations assume ali"activity is released to the room air for calculations of air activity, regardless of whether the pool is drained or not. 13.1.3 MHA Activity in Pool Water If 100% of the soluble fission products (Br, I) are absorbed in the pool water, the resulting activity would be filtered out and concentrated in the demineralizer. The demineralizer pumps pool water through at a rate of only 18 gpm, so it would take about a day for most of the activity to be deposited into the resins. The soluble fission product release from Table 13.1.1 (Br and I isotopes) was decayed for 24 hours and input into a Dantsys 2-D computer code to model the dose rate outside the demineralizer. The previous HEU SAR calculated a dose rate of 88 mR/hr at 1 meter from the demineralizer. For the revised HEU analysis the calculated dose rate is 84.2 mR/hr. For the LEU analysis the calculated dose rate is 90.8 mR/hr at BOL, 93.5 mR/hr at MOL, and 65.2 mR/hr at EOL. If the release fraction is accounted for, all dose estimates would be less than I mR/hr. These dose rates would not interfere with any necessary emergency repairs in the demineralizer area. 13.1.4 MHA Fission Product Release to Air within the Reactor Laboratory For calculating doses resulting from a release to the room air, it is assumed that the ventilation system is inoperable and the pool is drained. The ventilation system includes a backup exhaust fan, and both fans are on a building emergency diesel power circuit, so a loss of ventilation is considered a remote possibility. Complete loss of pool water is also a very remote possibility, but both assumptions are made for the MHA. UWNR LEU Conversion Analysis 174 August v 2008
The effective whole-body dose is calculated using the foll6wing equation24 Hf = h,.yf- Equation 13.1.2 where: HEff = effective whole-body dose from ith nuclide (rem). hEff = effective dose coefficient for the ith nuclide (rem-m 3 /Ci-s) X.= concentration of the ith nuclide in the cloud (Ci/m 3) t = time of cloud exposure (s) Equation 13.1.2 is used to calculate effective whole-body external dose by assuming submersion in an infinite cloud. Doses are calculated for each isotope individually and then summed. Dose to the thyroid is calculated using the following equation25: H,h = zVthh Equation 13.1.3 where: Hth = dose to thyroid from ith nuclide (rem) X = concentration of the ith nuclide in the cloud (Ci/m 3) V = breathing rate = 1.2 m 3/hr = 3.33E-4 m 3/s t = exposure time (s) hth = thyroid dose coefficient for the ith nuclide (rem/Ci) The released fission product inventory from Table 13.1.2 is assumed to spread uniformly in the reactor confinement volume. This volume is assumed to be 2000 m 3 which is a bounding assumption; a more realistic volume is at least 2500 m3 . Confinement concentrations are listed in Table 13.1.3. A total evacuation time of 10 minutes is assumed with a 5 minute evacuation time from confinement and an additional 5 minute evacuation time from the building. The doses in Table 13.1.4 through Table 13.1.6 are *for the 5 minute evacuation from confinement. As will be shown in section 13.1.5.1, the occupational worker will be subject to an additional dose equal to approximately 3% of the dose received in confinement. 175 August 2008 UANR LEU UWNR Conversion Analysis LEU Conversion Analysis 175 August 2008
As an example, the effective whole-body external dose and thyroid dose are calculated for 1-131 in the case of the revised HEU analysis. The effective whole-body external dose is:
= hE,1 3 lX, 3 HEJ = 6.73E -2 rem -mj3(1.41E - 4 )(300s)= 2.85mrem The thyroid dose is:
H,h I,131 = Z-A3Vthh = ( I - 4 CiY 3.33E M (300s 1.41E 1.08E6-e
= 15.3rem Other isotopes are calculated in the same fashion. The following isotopes are neglected due to their short half-lives (less than 10 minutes) and lack of dose coefficients: Br-87, 1-130m, 1-136, Kr-89, Xe-137. External doses are shown in Table 13.1.4 below, and thyroid doses are shown in table Table 13.1.5.
"JWNR .008 I.IJ Conversion Analvsis
- 176
. ,v Auvust
Table 13.1.3 MHA Occupational LablConcentration Isotope Previous Revised LEU BOL LEU MOL LEU EOL HEU SAR HEU Lab Conc. Lab Conc. Lab Conc. Lab Conc. Lab Conc. (Ci/mY) (Ci/m 3) (Ci/m3) (CU/M3) (Ci/m3) Br-82 1.190E-06 5.876E-08 2.191E-08 1.806E-07 3.020E-07 Br-83 4.166E-06 2.620E-05 4.07213-05 4.088E-05 2.765E-05 Br-84 7.698E-06 4.926E-05 7.644E-05 7.633E-05 5.129E-05 Br-85 1.004E-05 6.123E-05 9.484E-05 9.467E-05 6.353E-05 Br-87 2.381E-05 1.054E-04 1.633E-04 1.624E-04 1.085E-04 1-130m 7.936E-06 1.1 12E-07 2.922E-08 3.564E-07 6.387E-07 1-131 2.234E-05 1.412E-04 2.162E-04 2.274E-04 1.602E-04 1-132 3.393E-05 2.106E-04 3.282E-04 3.376E-04 2.369E-04 1-133 5.087E-05 3.294E-04 5.106E-04 5.226E-04 3.629E-04 1-134 6.166E-05 3.720E-04 5.768E-04 5.885E-04 4.070E-04 1-135 4.702E-05 3.067E-04 4.757E-04 4.866E-04 3.378E-04 1-136 2.389E-05 1.494E-04 2.329E-04 2.364E-04 1.627E-04 Kr-83m 4.166E-05 2.620E-05 4.069E-05 4.088E-05 2.765E-05 Kr-85m 1.004E-04 6.190E-05 9.585E-05 9.575E-05 6.426E-05 Kr-85 2.024E-05 9.128E-07 1.792E-07 2.151E-06 3.103E-06 Kr-87 1.928E-04 1.252E-04 1.938E-04 1.929E-04 1.289E-04 Kr-88 2.774E-04 1.768E-04 2.738E-04 2.724E-04 1.820E-04 Kr-89 3.393E-04 2.244E-04 3.474E-04 3.449E-04 2.297E-04 Xe-131m 1.984E-06 1.398E-06 2.092E-06 2.278E-06 1.562E-06 Xe-133m 1.230E-05 9.634E-06 1.348E-05 1.539E-05 1.077E-05 Xe- 133 5.087E-04 3.294E-04 4.839E-04 5.226E-04 3.632E-04 Xe-135m 1.389E-04 5.571E-05 8.620E-05 8.968E-05 6.348E-05 Xe-135 4.932E-04 2.039E-04 3.439E-04 3.457E-04 2.316E-04 Xe-137 4.702E-04 2.927E-04 4.537E-04 4.631 E-04 3.209E-04 Xe-138 3.547E-04 3.046E-04 4.721E-04 4.784E&04 3.279E-04 tPrevious HEU SAR assumed pool scrubbing 90% of Br and I isotopes when reporting lab concentrations. Analysis Conversion Analysis 177 August 2008 UWNR UWNR LEU Conversion 177 August 2008
Table 13.1.4 MHA OccupationalExternal Dose by Isotope Isotope Effective Revised HEU LEU BOL LEU MOL LEU EOL Dose Coef. 2 4 External External External External Worker Dose Worker Dose Worker Dose Worker Dose (rem-m3 /Ci-s) (rem) (rem) (rem) (rem) Br-82 4.810E-01 8.478E-06 3.1611E-06 2.605E-05 4.357E-05 Br-83 1.413E-03 1.111 E-05 1.727E-05 1.733E-05 1.173E-05 Br-84 3.482E-01 5.145E-03 7.985E-03 7.973E-03 5;357E-03 Br-85* Br-87* 1-130m* 1-131 6.734E-02 2.853E-03 4.367E-03 4.593E-03 3.237E-03 1-132 4.144E-01 2.618E-02 4.080E-02 4.197E-02 2.945E-02 1-133 1.088E-01 1.075E-02 1.666E-02 1.706E-02 1.184E-02 1-134 4.810E-01 5.368E-02 8.324E-02 8.492E-02 5.873E-02 1-135 2.953E-01 2.717E-02 4,213E-02 4.310E-02 2.992E-02 1-136* Kr-83m 5.550E-06 4.363E-08 6.775E-08 6.807E-08 4.604E-08 Kr-85m 2.768E-02 5.14013-04 7.958E-04 7;950E-04 5.335E-04 Kr-85 4.403E-04 1.206E-07 2.367E-08 2.841E-07 4.098E-07 Kr-87 1.52413-01 5.72513-03 8.864E-03 8.820E-03 5.895E-03 Kr-88 Kr-89* Xe-131m 3.774E-01 1.439E-03 2.002E-02 6.036E-07 3.100E-02 9.032E-07 3.085E-02 9.835E-07 2.0611E-02 6.744E-07 0 Xe-133m 5.069E-03 1.465E-05 2.050E-05 2.340E-05 1.638E-05 Xe-133 5.772E-03 5.704E-04 8;380E-04 9.050E-04 6.289E-04 Xe-135m 7.548E-02 1.261E-03 1.952E-03 2.031E-03 1.437E-03 Xe-135 4.403E-02 2.693E-03 4.543E-03 4.567E-03 3.059E-03 Xe-137* I Xe-138 2.135E-01 1.951E-02 3.023E-02 3.064E-02 2.100E-02
*Short-lived isotopes neglected for dose calculations.
UWNR LEU Conversion Analysis 178 August 2008 0
Table 13.1.5 MHA OccupationalThyroOl Dose by Isotope Isotope Thyroid Revised HEU LEU BOL LEU MOL LEU EOL Dose Coef.2 5 Thyroid Thyroid Thyroid Thyroid Worker Dose Worker Dose Worker Dose Worker Dose (rem/Ci) (rem) (rem) (rem) (rem) 1-130m* 1-131 1.080E+06 1.526E+01 2.336E+01 2.456E+01 1.731E+01 1-132 6.438E+03 1.356E-01 2.113E-01 2.174E-01 1.525E-01 1-133 1.798E+05 5.923E+00 9.181E+00 9.398E+00 6.526E+00 1-134 1.066E+03 3.964E-02 6.147E-02 6.271E-02 4.337E-02 1-135 3.130E+04 9.600E-01 1.489E+00 1.523E+00 1.057E+00 1-136* 1
*Short-lived isotopes neglected for dose calculations.
The total external and thyroid dose for a worker in confinement for 5 minutes is shown below, along with the Total Effective Dose Equivalent (TEDE). Table 13.1.6 MHA Total OccupationalDose during 5 minute evacuation External Dose Thyroid Dose TEDE* (rem) (rem) (rem) Previous HEU SAR 0.01 N/A N/A Revised HEU Analysis 0.176 22.3 0.846 LEU BOL Analysis 0.273. 34.3 1.30 LEU MOL Analysis 0.278 35.8 1.35 LEU EOL Analysis 0.192 25.1 0.945
- TEDE was calculated by using a thyroid weighting factor of 0.03 (10 CFR 20.1003).
All doses are below the occupational limit of 5 rem whole-body and 50 rem thyroid dose (10 CFR 20.1201). No comparison can be made to the previous HEU SAR thyroid dose because the previous calculation took credit for pool scrubbing of soluble isotopes and therefore is not representative of the MHA. An appropriate comparison of thyroid dose is made with the previous HEU SAR in section 13.1.6. 0 UWNR LEU Conversion Analysis LEU Conversion Analysis 179 179 August 2008 August 2008
13.1.5 MHA Release of Fission Products to Unrestricted Areas The analysis for unrestricted areas considers two exposure categories: an occupant of the Mechanical Engineering Building who is exposed during the evacuation, and a member of the public standing just outside the building. The nearest residence is also included in the category of someone standing just outside the building due to the close proximity of the nearest residence from the Mechanical Engineering Building (approximately 80 in). Doses are calculated for the MHA in this section, but in the following sections doses are also calculated using more realistic assumptions including intact pool and operable ventilation for comparison. 13.1.5.1 MHA Dose to Building Occupant In the event of a serious release, the nearby occupants of the Mechanical Engineering Building would be evacuated. However, doses are calculated for the occupants as they are evacuating. The assumptions are:
" Evacuation time of 10 minutes.
- Released fission products will first disperse uniformly throughout the reactor confinement plus auxiliary support space.
- Released fission products will then disperse uniformly throughout the Mechanical Engineering, building, excluding the fifth floor.
The assumption that the release must first disperse through the auxiliary support space before reaching unrestricted areas is reasonable because all potential leak paths into unrestricted areas must first go through the auxiliary support space. Once the release escapes into unrestricted UWNR LEU Conversion Analysis 180 August 2008 0
areas, it would be quickly dispersed throughout the resf of the building due to the building ventilation system and the large open atrium area connecting floors 1-4. Released activities are taken from Table 13.1.2. Effective whole-body external dose is calculated using Equation 13.1.2, and thyroid dose is calculated using Equation 13.1.3. Room volumes were calculated by estimating the floor area from building drawings, and assuming an 8 ft tall ceiling for a bounding analysis. Volumes are summarized in the table below. Table 13.1. 7 Building Volumes usedfor Ground Release Area Description Floor Area Area Volume Total (ft2) (M3) Volume (mi3) Control Room and Auxiliary Support 4,500 1020 3,020 Space Non-Restricted ME Basement 42,000 9,515 12,535 Non-Restricted ME 1st through 4th 57,500 per floor 13,025 per floor 64,635 Floors 230,000 total 52,100 total The external and thyroid doses are summed for each isotope as done in section 13.1.4. Results are summarized in the table below. Table 13.1. 8 MHA Building Occupant Dosesfor Ground Release External Dose Thyroid Dose TEDE* (mrem) (mrem) (mrem) Revised HEU Analysis 10.9 1,380 52.3 LEU BOL Analysis 16.9 2,120 80.6 LEU MOL Analysis 17.2 2,210 83.6 LEU EOL Analysis 11.9 1,550 58.5
- TEDE was calculated by using a thyroid weighting factor of 0.03 (10 CFR 20.1003).
181 August 2008 UWNR UWNR LEU Conversion Analysis LEU Conversion Analysis 181 Aug ust 2008
Doses are below the limit to the public of 100 mrem (10 CFR 20.1301). The previous HEU SAR did not consider the dose to building occupants outside reactor confinement, so no comparison can be made. 13.1.5.2 MHA Dose to Outside Building Atmospheric dispersion from a ground release is difficult to model close to the source. The Gaussian Plume Model is not valid for distances less than 100 m. Therefore, the ground release calculations in this section follow the method presented in NUREG/CR-238726, which makes the bounding assumption that the concentration divided by the release rate, or x/Q, is equal to 0.01 s/m3, where the release time is 1 hour (3600s). A 1 hour release time must be used because the Gaussian Plume Model is generally not valid for release times shorter than 1 hour, therefore assuming a release time of 1 hour effectively calculates a 1 hour averaged dose. Even if the release was shorter than 1 hour, it would simply result in higher concentrations for a shorter time but the averaged dose would not change. Using the variable x/Q is often confusing and makes comparisons with the previous HEU SAR difficult; therefore this variable was converted to simple concentration for our analysis. By substituting A/3600s for Q, where A is the activity, then the assumption that V/Q is equal to 0.0.1 s/mi3 is equivalent to assuming that the concentration X is equal to A times 2.78E-6 M3. This is equivalent to uniformly diluting the activity throughout a volume of 360,000 in 3 . Using the same dose equations as before (Equation 13.1.2 and Equation 13.1.3) and the same methodology as in section 13.1.4, using the released inventory from Table 13.1.2, the dose to a UWNR LEU Conversion Analysis 182 August 2008 S
member of the public standing outside the evacuated Mechanical Engineering Building is given in the table below. Table 13.1.9 MHA Dose to Outside Building External Dose Thyroid Dose TEDE* (mrem) (mrem) (mrem) Previous HEU SAR 0.153 1,019 30.7 Revised HEU Analysis 11.7 1,490 56.4 LEU BOL Analysis 18.2 2,290 86.8 LEU MOL Analysis 18.6 2,380 90.1 LEU EOL Analysis 12.8 1,670 63.0
- TEDE was calculated by using a thyroid weighting factor of 0.03 (10 CFR 20.1003).
This is also the nearest resident dose, because the nearest residence is within 100 m of the Mechanical Engineering Building (approximately 80 m). Doses are below the limit to the public of 100 mrem (10 CFR 20.1301). 13.1.6 Near MHA With Pool Intact and Ventilation System Inoperable The following analysis for the Near MHA with pool intact is identical to the MHA assumptions except that the pool is assumed to be intact and therefore credit is taken for the water scrubbing soluble fission products out. Even in this analysis, the bounding assumption is made that 10% of the soluble fission products still escape into the confinement atmosphere. The new activity of released fission products is shown in the table below. Conversion Analysis 183 August 2008 UWNR LEU UWNR LEU Conversion Analysis 183 August 2008
Table 13.1.10 Near MHA with Pool Intact Released FissionProduct Inventories Isotope Previous Revised LEU BOL LEU MOL LEU EOL 0 HEU SAR HEU Released Released Released Released Released (Ci) (ca) (Ci) (ca) (ci) Br-82 2.381E-03 1.175E-05 4.381E-06 3.6.11E-05 6.039E-05 Br-83 8.333E-03 5.240E-03 8.145E-03 8.176E-03 5.531E-03 Br-84 1.540E-02 9.851E-03 1.529E-02 1.527E-02 1.026E-02 Br-85 2.008E-02 1.225E-02 1.897E-02 1.893E-02 1.271E-02 Br-87 4.761E-02 2.108E-02 3.266E-02 3.248E-02 2.169E-02 1-130m 1.587E-02 2.224E-05 5.843E-06 7.127E-05 1.277E-04 1-131 4.468E-02 2.825E-02 4.324E-02 4.547E-02 3.205E-02 1-132 6.785E-02 4.212E-02 6.563E-02 6.752E-02 4.738E-02 1-133 1.017E-01 6.588E-02 I.021E-01 1.045E-01 7.258E-02 1-134 1.233E-01 7.440E-02 1.154E-01 1.177E-01 8.140E-02 1-135 9.404E-02 6.134E-02 9.513E-02 9.731E-02 6.755E-02 1-136 4.777E-02 2.988E-02 4.657E-02 4.727E-02 3.255E-02 Kr-83m 8.333E-02 5.240E-02 8.138E-02 8.176E-02 5.531E-02 Kr-85m 2.008E-01 1.238E-01 1.917E-01 1.915E-01 1.285E-01 Kr-85 4.047E-02 1.826E-03 .3.583E-04 4.302E-03 6.205E-03 Kr-87 3.857E-01 2.504E-01 3.877E-01 3.857E-01 2.578E-01 Kr-88 Kr-89 5.547E-01 6.785E-01 3.536E-01 4.487E-01 5.475E-01 6.948E-01 5.449E-01 6.898E-01 3.640E-01 4.594E-01 0 Xe-131m 3.968E-03 2.796E-03 4.183E-03 4.555E-03 3.124E-03 Xe-133m 2.460E-02 1.927E-02 2.696E-02 3.078E-02 2.155E-02 Xe-133 1.017E+00 6.588E-01 9.679E-01 1.045E+00 7.264E-01 Xe-135m 2.778E-01 1.114E-01 1.724E-01 1.794E-01 1.270E-01 Xe-135 9.864E-01 4.077E-01 6.879E-01 6.914E-01 4.631E-01 Xe-137 9.404E-O1 5.855E-01 9.074E-O1 9.262E-01 6.418E-01 Xe-138 7.095E-01 6.092E-01 9.441E-01 9.567E-01 6.558E-01 184 August 2008 UWNR LEU ConVersion Analysis LEU Con'Version Analysis 184 August 2008
0 13.1.6.1 Near MHA With Pool Intact Fission Product Release to Air Within the Reactor Laboratory The analysis from section 13.1.4 was repeated for the Near MHA with pool intact. The total external and thyroid dose for a worker in confinement for 5 minutes is shown below, along with the Total Effective Dose Equivalent (TEDE). Table 13,1.11 Near MHA with Pool Intact OccupationalDose during 5 minute evacuation External Dose Thyroid Dose TEDE* (rem) (rem) (rem) Previous HEU SAR N/A 18.9 N/A Revised HEU Analysis 0.0629 2.23 0.130 LEU BOL Analysis 0.0978 3.43 0.201 LEU MOL Analysis 0.0986 3 58 0.206 LEU EOL Analysis 0.0670 2.51 0.142
- TEDE was calculated by using a thyroid weighting factor of 0.03 (10 CFR 20.1003).
0 All doses are below the occupational limit of 5 rem whole-body and 50 rem thyroid dose (10 CFR 20.1201). 13.1.6.2 Near MHA With Pool Intact Dose to Building Occupant The analysis from section 13.1.5.1 was repeated for the Near MHA with pool intact. The results are shown below. Table 13.1.12 Near MHA with PoolIntact Building OccupantDoses for Ground Release External Dose Thyroid Dose TEDE* (mrem) (mrem) (mrem) Revised HEU Analysis 3.89 138 8.04 LEU BOL Analysis 6.05 212 12.4 LEU MOL Analysis 6.10 ,221 12.7 LEU EOL Analysis 4.15 155 8.81
- TEDE was calculated by using a thyroid weighting factor of 0.03 (10 CFR 20.1003).
UWNR LEU Conversion Analysis 185 August 2008
All doses are within the limit to the public of 100 mrem (10 CFR 20.1301). 13.1.6.3 Near MHA With Pool Intact Dose to Outside Building The analysis from section 13.1.5.2 was repeated for the Near MHA with pool intact. The dose to a member of the public standing outside the evacuated Mechanical Engineering Building is given in the table below. Table 13.1.13 Near MHA With Pool Intact Dose to Outside Building External Dose Thyroid Dose TEDE* (mrem) (mrem) (mrem) Previous HEU SAR 0.084 102 3.14 Revised HEU Analysis 4.19 149 8.66 LEU BOL Analysis 6.52 229 13.4 LEU MOL Analysis 6.57 238 13.7 LEU EOL Analysis 4.47 167 9.49
- TEDE was calculated by using a thyroid weighting factor of 0.03 (10 CFR 20.1003).
These doses are below the 100 mrem limit for public exposure (10 CFR 20.1301). This is also the nearest resident dose, because the nearest residence is within 100 m of the Mechanical Engineering Building (approximately 80 m). 13.1.7 Near MHA With Pool Drained and Ventilation System Operable The following analysis for the Near MHA with ventilation intact is identical to the MHA assumptions except that the ventilation system is assumed to be operable and therefore credit is taken for the dilution and dispersion that would result from an elevated release. The released I 13 M1TUMKTI? - V-"' AI lf-J 1 AZ1 "Qt 9flRA M 0
inventories are identical to those of the MHA shown in Table 13.1.2. Occupational doses within confinement are assumed to be the same as those calculated for the MHA in section 13.1.4. Doses to building occupants are eliminated, because with the ventilation system operable the confinement is keptat a negative pressure and so no release would leak out into the building. Only the dose outside the building is recalculated for the Near MHA with ventilation intact. Note that the previous HEU SAR did not analyze this near MHA condition. 13.1.7.1 Ventilation System Effective Stack Height With the ventilation system running, the release is discharged through a Strobic Air Tri-Stack exhaust fan. The effective stack height is determined by the physical height of the stack, the buoyancy of a hot effluent, and the vertical momentum from the exhaust fan. The roof height of the Mechanical Engineering Building is 26.5 m. The height of the exhaust fan is 4 m, so the true physical height of the stack is 30.5 rn above ground. Typically the buoyancy rise and momentum rise are each calculated, and whichever is the larger effect is used in calculations (the other is neglected). In our case the exhaust temperature is always at room temperature (approximately 720 F). The energy from the exhaust fan will heatrthe air slightly, but in the absence of exhaust temperature measurements this heating effect is neglected. Therefore only in the winter months would the buoyancy rise be significant. Because there would be no buoyancy rise during the summer months, the buoyancy rise is neglected for conservatism. The momentum rise is calculated using the following equation27 3dv AH- Equation 13.1.4 Uh UWNR LEU Conversion Analysis 187 August 2008
where: AH = increase in stack height (m) d = top inside stack diameter (m) [30.5in or 0.7747m] 2 1 v = stack gas exit velocity (m/s) [3400fpm or 17.272m/s at 9600cfm] 29 Uh = wind speed at physical stack height (m/s) Using 3.54m/s for the wind speed (lowest monthly average on record) and the supplied stack 'diameter and exit velocity (which is only valid for a fan flow of 9600cfm), the increase in stack height is calculated using Equation (4) to be 1 1.3m. This value is rounded down to 11 m for further calculations. When added to the 30.5 m physical stack height, this results in an effective stack emission height of 41.5 m. 13.1.7.2 Maximum Ground Level Concentration Ground level concentrations are calculated using the Gaussian Plume Model. The full equation for down-wind concentration is27 AX_ Y QH) expi- 2y [ U expor xp~(H.-z)2 exp 2a'
+exp e
F(H+z) 2(T' 2 Equation 13.1.5 where: X= concentration (Ci/m 3) x = receptor down-wind distance (m) y = receptor cross-wind distance (m) z = receptor height above ground (m) H = effective stack emission height (m) Q = release rate (Ci/s) u = mean wind speed (m/s) (y = cross-wind dispersion coefficient (m) o, = vertical dispersion coefficient (m) UWNR LEU Conversion Analysis 188 August 2008
For calculating concentrations at ground-level (7=O) and under the center of plume travel (y=O), the equation reduces to: X(x,O,0Q; H) exp H Equation 13.1.6 uo-Uo7 CXP 2cTj For urban environments, the same reference gives the following equations for the dispersion coefficients based on the Pasquill Stability Class A-F: A-B: ary =0.32x(l + O.0004x)-° 5 o-, 0.24x(l + 0.00 Ix)" 5 Equation 13.1.7 C: 0rY = 0.22x(4 + 0.0004x)- 5 o- = 0.20x D: a, =0.16x(l+0.0004x) 5 a,--.= 4x(l +o0.0003x)- 5 5 E-F: ory=O. llx(1+ 0.0004x)- o, = 0.08x(l + 0.001 5x)-° Where: x = down-wind distance (m) It is important to note that the equations and coefficients given for the Gaussian Plume Model are not valid for receptors very close to the source (less than 100 m) or for release times less than I hour. For this reason, a 1 hour release is assumed even though a true release could potentially occur in a shorter time period (as short as 26 minutes). A shorter release would result in higher concentrations but shorter exposure times, so the I hour averaged exposure would not change and therefore the assumption of release time does not impact dose estimates. The following figure illustrates the ground level concentration (per Ci/s release rate of the isotope in question) as a function of both distance from the source and Pasquill Stability Class. Class A represents very unstable conditions, and class F represents extremely calm conditions. The wind speed was assumed to be 3.54 m/s (lowest monthly average on record). UWNR LEU Conversion Analysis 189 August 2008
,0 4.05-05 -_______
3,5E-05 3.0E-05 2,5E-05 H.-c
- a. a. 2,OE-05 CUS-c12 1.5E-05 0
1,OE-05 5.0E-06 O.OE+00 0 50 100 150 200 250 300 Down-wind Distance (m) 350 400 450 500 0 Figure 13. 1. 1 Ground Level Concentration vs. !)isiance (Q= I Ci/s) As shown. in Figure. 13.1.1, the maximum relative concentration is 3.5935E-5 CJ/m 3 per Ci/s released, and this maximum occurs at a down-wind distance of 148 m under class C stability conditions. 13.1.7.3 Near MHA With Ventilation Intact Dose to Outside Building After accounting for the operation of the ventilation system, the resulting public concentrations for each isotope and total doses are shown in the tables below for the case of the near MHA with ventilation intact. The calculation of doses follows the methods presented in section 13.1.5.2. UWNR UWNR LEU Conversion Analysis LEU Conversion Analysis 190 190 August 2008 August 2008 0
Table 13.1.14 Near MHA With VentilationIntact: Public Concentrationby Isotope Near MHA with Ventilation Intact Public Concentration Isotope (Ci/m_) Revised HEU LEU BOL LEU MOL LEU EOL Br-82 1.173E-12 4.373E-13 3.605E-12 6.028E-12 Br-83 5.231E-10 8.130E-10 8.161E-10 5.521E-10 Br-84 9.833E-10 1.526E-09 1.524E-09 1.024E-09 Br-85* 1.222E-09 1.893E-09 1.890E-09 1.268E-09 Br-87* 2.104E-09 3.260E-09 3.242E-09 2.165E-09 I-130m* 2.220E-12 5.833E-13 7.114E-12 1.275E-11 1-131 2.820E-09 4.316E-09 4.539E-09 3.199E-09 1-132 4.204E-09 6.552E-09 6.740E-09 4.729E-09 1-133 6.576E-09 1.019E-08 1.043E-08 7.245E-09 1-134 7.426E-09 1.152E-08 1.175E-08 8.126E-09 1-135 6.123E-09 9.496E-09 9.713E-09 6.743E-09 1-136* 2.982E-09 4.649E-09 4.719E-09 3.249E-09 Kr-83m 5.231E-10 8.123E-10 8.161E-10 5.521E-10 Kr-85m 1.236E-09 1.914E-09 1.912E-09 1.283E-09 Kr-85 1.822E- 11 3.577E-12 4.295E-1 1 6.194E- II Kr-87 2.499E-09 3.870E-09 3.850E-09 2.573E-09 Kr-88 3.529E-09 5.465E-09 5.439E-09 3.633E-09 Kr-89* 4.479E-09 6.935E-09 6.886E-09 4.586E-09 Xe-131m 2.791E-11 4.176E-11 4.547E- 1I 3.118E-11 Xe-133m 1.923E-10 2.691E-10 3.072E-10 2.151E-10 Xe-133 6.576E-09 9.661E-09 1.043E-08 7.250E-09 Xe-135m 1.1 12E-09 1.721 E-09 1.790E-09 1.267E-09 Xe-135 4.070E-09 61866E-09 6.902E-09 4.623E-09 Xe-137* 5.844E-09 9.057E-09 9.246E-09 6.406E-09 Xe-138 6.081E-09 9.424E-09 9.550E-09 6.546E-09 191 August 2008 Conversion Analysis LEU Conversion UWNR LEU Analysis 191 August 2008
Table 13.1.15 Near MHA With Ventilation Intact: Dose to Outside Building External Dose Thyroid Dose TEDE* (mrem) (mrem) (mrem) 0 Previous HEU SAR N/A N/A N/A Revised HEU Analysis 0.0422 5.35 0.203 LEU BOL Analysis 0.0655 8.22 0.312 LEU MOL Analysis 0.0667 8.57 0.324 LEU EOL Analysis 0.0459 6.01 0.226
- TEDE was calculated by using a thyroid weighting factor of 0.03 (10 CFR 20.1003).
13.1.8 Near MHA With Pool Intact and Ventilation System Operable The following analysis for the Near MHA with pool and ventilation intact is identical to the MHA assumptions except that the ventilation system is assumed to be operable and the pool is assumed to be intact. The released inventories are identical to those of the Near MHA with pool intact shown in Table 13.1.10 from section 13.1.6. Occupational doses within confinement are assumed to be the same as those calculated for the Near MHA with pool intact in section 13.1.6. 0 Doses to building occupants are eliminated, because with the ventilation system operable the confinement is kept at a negative pressure and so no release would leak out into the building. Only the dose outside the building is recalculated. 13.1.8.1 Near MHA With Pool and Ventilation Intact: Dose to Outside Building Details of modeling operation of the ventilation system are covered in sections 13.1.7.1 and 13.1.7.2. The resulting doses outside the building are shown in the table below for the Near MHA with pool and ventilation intact. Conversion Analysis UWNR LEU Conversion Analysis 192 192 August 2008 August 2008 0
Table 13.1.16 Near MHA With Pool and Ventilation Intact: Dose to Outside Building External Dose Thyroid Dose TEDE* (mrem) (mrem) (mrem) Previous HEU SAR 0.006 10.0 0.306 Revised HEU Analysis 0.0151 0.535 0.0311 LEU BOL Analysis 0.0234 0.822 0.0481 LEU MOL Analysis 0.0236 0.857 0.0493 LEU EOL Analysis 0.0161 0.601 0.0341
- TEDE was calculated by using a thyroid weighting factor of 0.03 (10 CFR 20.1003).
13.1.9 Expected Dose to Public The calculation of public dose outside the building is now repeated using more realistic assumptions in order to compare with the results of the MHA. The MHA assumed the pool was drained and the ventilation system was inoperable, with previous continuous operation at 1.3 MW. The following analysis assumes the pool is intact and the ventilation system is operable,
- with previous continuous operation at 1.02 MW.
Using the methods from section 13.1.1, the fission product inventories were recalculated assuming the maximum power density at 1.02 MW. The release fraction is now calculated assuming the maximum fuel temperatures at 1.02 MW. These temperatures are 475.38 'C for the revised HEU analysis and 490.15/491.51/475.43 0 C for the LEU BOL/MOL/EOL analyses, taken from Table 4.7.7, Table 4.7.12, Table 4.7.15, and Table 4.7.17. Using these temperatures, the release fraction is 7.549E-5 for the revised HEU analysis and 1.005E-4/1.032E-4/7.556E-5 for the LEU BOL/MOL/EOL analyses. In addition to the smaller release fractions, credit is now taken for pool water scrubbing the soluble fission products (Br and I). The bounding assumption is made that 10% of these soluble fission products will still escape into the confinement O UWNR LEU Conversion Analysis 193 August 2008
atmosphere. These new assumptions result in the fission. product inventories and released activities shown in the tables below. 0 Table 13.1.17 Expected Fission Product Inventories Isotope Revised HEU LEU BOL LEU MOL LEU EOL Expected Expected Expected Expected Inventory Inventory Inventory Inventory (Ci (Ci) (Ci) (Ci) Br-82 1.751E-01 4.203E-02 3.824E-01 9.028E-01 Br-83 7.808E+01 8.877E+01 8.656E+01 8.271E+O1 Br-84 1.468E+02 1.667E+02 1.616E+02 1.534E+02 Br-85 1.825E+02 2.068E+02 2.005E+02 1.901E+02 Br-87 3.140E+02 3.561E+02 3.439E+02 3.245E+02 1-130m 3.314E-01 5.266E-02 7.565E-01 1.912E+00 1-131 4.208E+02 4.710E+02 4.817E+02 4.790E+02 1-132 6.275E+02 7.150E+02 7.148E+02 7.085E+02 1-133 9.815E+02 1.113E+03 1.107E+03 1.085E+03 1-134 1.108E+03 1.257E+03 1.246E+03 1.2188E+03 1-135 9.138E+02 1.036E+03 1.030E+03 1.010E+03 1-136 4.452E+02 5.076E+02 5.004E+02 4.868E+02 Kr-83m Kr-85m Kr-85 7.808E+01 1.845E+02 2.720E+00 8.874E+01 2.090E+02 3.905E-01 8.654E+01 2.027E+02 5.700E+00 8.270E+01 1.923E+02 1.158E+01 0 Kr-87 3.730E+02 4.226E+02 4.084E+02 3.856E+02 Kr-88 5.268E+02 5.970E+02 5.768E+02 5.445E+02 Kr-89 6.685E+02 7.576E+02 7.302E+02 6.871E+02 Xe-131 m 4.165E+00 4.557E+00 4.680E+00 5.319E+00 Xe-133m 2.871E+01 2.938E+01 3.259E+01 3.223E+01 Xe-133 9.815E+02 1.054E+03 1.107E+03 1.086E+03 Xe-135m 1.660E+02 1.878E+02 1.899E+02 1.899E+02 Xe-135 6.075E+02 8.028E+02 7.858E+02 7.516E+02 Xe-137 8.723E+02 9.884E+02 9.807E+02 9.598E+02 Xe-138 9.077E+02 I1.029E+03 1.013E+03 9.808E+02 UWNR LEU UWNR Conversion Analysis LEU Conversion Analysis 194 194 August 2008 August 2008 0
Table 13. 1.18 Expected Fission PriductRelease Isotope Revised HEU LEU BOL LEU MOL LEU EOL Expected Expected Expected Expected Released Released Released Released (Ci) (Ci) (Ci) (Ci) Br-82 1.322E-06 4.225E&07 3.948E-06 6.822E-06 Br-83 5.894E-04 8.924E-04 8.937E-04 6.250E-04 Br-84 1.1 08E-03 1.676E-03 1.668E-03 1.1 59E-03 Br-85 1.377E-03 2.079E-03 2.070E-03 1.436E-03 Br-87 2.370E-03 3.580E-03 3.550E-03 2.452E-03 1-130m 2.502E-06 5.294E-07 7.81 OE-06 1.445E-05 1-131 3.177E-03 4.735E-03 4.973E-03 3.619E-03 1-132 4.737E-03 7.188E-03 7.380E-03 5.353E-03 1-133 7.409E-03 1.119E-02 1.143E-02 8.198E-03 1-134 8.368E-03 1.264E-02 1.286E-02 9,203E-03 1-135 6.898E-03 1.041E-02 1.063E-02 7.632E-03 1-136 3.360E-03 5.103E-03 5.166E-03 3.678E-03 Kr-83m 5.894E-03 8.921 E-03 8.934E-03 6.249E-03 Kr-85m 1.392E-02 2.101E-02 2.093E-02 1.453E-02 Kr-85 2.053E-04 3.926E-05 5.885E-04 8.750E-04 Kr-87 2.816E-02 4.248E-02 4.216E-02 2.914E-02 Kr-88 3.976E-02 6.002E-02 5.955E-02 4.114E-02 Kr-89 5.047E-02 7.616E-02 7.539E-02 5.192E-02 Xe-131m 3.144E-04 4.581 E-04 4.832E-04 4.019E-04 Xe-133m 2.167E-03 2.954E-03 3.365E-03 2.435E-03 Xe-133 7.409E-02 1.060E-01 1.143E-01 8.206E-02 Xe-135m 1.253E-02 1.888E-02 1.961E-02 1.435E-02 Xe-135 4.586E-02 8.070E-02 8.113E-02 5.679E-02 Xe-137 6.585E-02 9.936E-02 1.012E-01 7.252E-02 Xe-138 6.852E-02 1.034E-01 1.046E-01 7.411 E-02 Details of modeling operation of the ventilation system are covered in sections 13.1.7.1 and 13.1.7.2. The resulting public doses are shown in the table below for the case of the expected accident release. 0 UWNR LEU Conversion Analysis LEU Conversion Analysis 195 195 August 2008 August 2008
Table 13.1.19 Expected Dose to Outsi~le Building External Dose Thyroid Dose TEDE* (mrem) (mrem) (mrem) Revised HEU Analysis 1.69E-3 6.OIE-2 3.50E-3 LEU BOL Analysis 2.58E-3 9.OOE-2 5.28E-3 LEU MOL Analysis 2.59E-3 9.37E-2 5.40E-3 LEU EOL Analysis I .82E-3 6.79E-2 3.86E-3
- TEDE was calculated by using a thyroid weighting factor of 0.03 (10 CFR 20.1003).
All of these doses are orders of magnitude below the 100 mrem limit. 13.2 Rapid Addition of Reactivity Accident This accident postulates the worst case result of insertion of excess reactivity of the maximum allowed experiment reactivity worth or ejection of the transient rod (1.4%Ak/K) while the reactor is operating at 1.3 MW. The hot pin located at D5 SW has the highest pin power of 22.886 kW for HEU BOL and would experience the highest fuel temperature after the pulse. In addition, it is assumed that the inlet water temperature is 54.44°C (130'F) and that the water level is 5.7912m (19 feet) above the core. This provides the most limiting input conditions for this accident. The limitation of experiment reactivity to 1.4%Ak/K insures that reactivity insertions from experiment removal or failure will result in consequences no worse than those considered here. The existing interlocks and administrative limits on firing the transient rod will not change as a result of the LEU conversion. 13.2.1 Temperature after Pulse for HEU BOL Firing a pulse while at 1.3 MW would cause the reactor to scram from power level and fuel temperature scrams. According to technical specifications, the instant a simulated signal reaches UWNR LEU Conversion Analysis 196 August 2008 0
. the value of the LSSS to the instant the slowest scrammdble control element reaches its fully inserted position shall not exceed 2 seconds. In this analysis, it is assumed that a 1.4%Ak/k ($1.859) pulse is fired while at 1.3 MW and that no blades fall in until the 2 second mark where all blades then fall into the core instantaneously, inserting -8.86%Ak/k (-$11.766) of reactivity.
-8.86%Ak/k was the calculated MCNP5 total scrammable blade worth at HEU BOL. Therefore, the entire pulse power and energy release as seen in Figure 1.3.2.1 are used to determine the temperature profile after'a pulse at 1.3 MW as seen in Figure 13.2.2. The delayed neutron fraction for HEU BOL is calculated to be 0.00753.
Core Power and Energy vs. Time after 1.4% Ak/K Pulse at 1.3 MW 1.E.10 L..........1.E+10 1.94 GW - Power
-Energy 11+09 --- - 1.E+09 O-"
1I1.E+08 1E+0 7 L. . . . ...-------.--............................. .. .... ... .... ..... .. ... ... . ....... .... .. . .. . . .. .... .. .. .. .... . ... .. . ... . ... 1.. 1.. 0 7E 1.E+06 o C 0 0,2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2 Time Is) Figure 13.2. 1 Core Power and Energy vs. Time afier /. 4%4 k/k Pulse al 1.3 MJWplotted br 2s (HIEU BOL) 197 August 2008 UWNR LEU UWNR Conversion Analysis LEU Conversion Analysis 197 August 2008
Max Hot Rod Temperature after 1.4% k/k Pulse at 1.3 MW (HEU BOL) 1200
- 2132 1100 -
1932 1000
! - 1732
.- 900 L - 1532 S800 E
70 -1332 700 600 .. . .. .. . . . . .. .. . . .- 1132
-- -- Safety Analysis Limit I Soo 932 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2 Time [s]
Figure 13.2.2 Temperature Profile vs, Time after 1.4%Jk/k pulse at 1.3 MW plotted for 2s (HEU BOL) After the blades fall in, the power and temperature are plotted out to 100 seconds. Immediately after the blades fall in at 2 seconds from the initiation of the 1.4%Ak/k pulse, the peak temperature is reached at 864.05'C (1587.297F). The peak fuel temperature is 285.95°C (512.710F) below the fuel temperature safety limit of 1150°C (2100'F). At no time during the accident will the fuel temperature be exceeded within the core for this accident scenario. The core power and temperature profile plots after the blades have fallen into, the core are shown in Figure 13.2.3 and Figure 13.2.4. Analysis 198 August 2008 UWNR LEU UWNR LEU Conversion Conversion Analysis 198 August 2008
Core Power after Blades SCRAM at 2s following 1.4% Ak/k pulse at 1.3 MW (HEU BOL) 1.E+07 11E+05 a.0 1.E+04 0 10 20 30 40 50 60 70 80 90 100 Time (s] Figure 13, 2.3 Core Power vs. Timne after 1.4%Akik pulse(at 1,3 MW following SCRAM (IHEU BOL) Maximum Hot Rod Temperature after Blades SCRAM at 2s following 1.4% Ak/k Pulse at 1.3 MW (HEU BOL) 1200 _ S2032 864.05 S800 ..... . 600 --- - 1032 2400
-- Safety Analysis Limit= 1150'C 0 L 7.--- ...-.. - -- ----.... .... . --,.... ......... ......... ... .. 4 32 "0 10 20 30 40 50 60 70 80 90 100 Time [s]
t-'igure 13.2,4 Temperature Profile vs. Time af 1.4% k/k pulse at 1.3 MWfollowing SCRAM (lIEU BOL) 199 August 2008 UWNR LJWNR LEU Conversion Analysis LEU Conversion Analysis 199 August 2008
It is important to note that RELAP5 / MOD3.3 is predicting that the fuel temperature would continue to rise, even after the pulse had terminated within 0.05 seconds, The code is predicting the heat transfer coefficient will decrease, because the power released during the pulse occurs when the cladding temperature is hot. In addition, RELAP5/MOD 3.3 does not model axial heat conduction within the rod, and thus is overestimating the temperature in the rod itself; since the only heat removal is to the water surrounding the rod. After power falls more rapidly due' to the insertion of the Control rod's, the heat transfer rate is able to recover and keep the maximum fuel temperature under I 150TC (2100°F). Even with these limitations, a 1.4%Ak/k pulse at 1.3 MW would not cause fuel damage or release fission products from the reactor. For comparison, the maximum fuel temperature after the pulse at 0.0410 seconds is 784.75°C (1444.55°F) and the maximum fuel temperature after 2,00 seconds is 864;05'C (1587.29°F). These temperature surface plots are shown adjacent to each other as seen in Figure 13.2.5. 200 August 2008 Conversion Analysis UWNR LEU Conversion Analysis 200 August 2008
40[...................... 40000[.......*. 35 700 35 E o700 30 - 30 6 -600 , E 25 £25 0 50 0 o2 20 500 20 15 40000 r--, 10 5 5 0 1 2 0 1 2 Radial Distance from Fuel Centerline [cm] Radial Distance from Fuel Centerline [cm] pulse at 1,3 MW at 0.0410s (left)& 2.0OOs (right) (HEUOBOL) Figure 13.2,5 Fuel temperatureplot after 1.4 %Ykik 13.2.2 Temperature after Pulse for LEU BOL Using the same methodology as stated in 13.2.1, the temperature after a 1.4%Ak/k($1.790) pulse at 1.3 MW for the LEU BOL core can be tound. The power and energy plot for the first 2 seconds is shown in Figure 13.2.6 and the maximum fuel temperature for the first 2 seconds is shown in Figure 13.2.7. After the scrammable blades with a total worth calculated by MCNP5 of -8.188%Ak/k (-$10.470) fall into the core instantaneously at 2 seconds, the core power is plotted in Figure 13..2.8 and the maximum fuel temperature is shown in Figure 13.2.9. The delayed neutron fraction is calculated for LEU BOL to be 0.00782. UWNR LEU Conversion Analysis 201 August 2008
Core Power and Energy after 1.4% ak/k pulse at 1.3 MW (LEU BOL) 1.E+ 10 --- - -____ ---- 1.E+09 1.69E+09 Power
- Energy 1.E+0 9 - . .. .. . .......
------............. . . . . .. . . .. . . . . . . 11E+ 08 0
1.E+06 11E+075* 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2 rime [s] Fque13. 2.6 Power aind totcal energi, after 1.4%k plsat13fW(.&BL Maximum Hot Rod Temperature after 1.4 A%k/k Pulse at 1.3 MW (LIEU BOL) 1200 1932 1000 .
.1153 a.800...............-..-..- 0.
E E I- 332 700 ________ 600 . . . ........ .......... .. 3132
-- Safety Analysis Limit= 1150*C 500 -- - - - - - - - - - ~ - . . . .~932 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 18 2 Time [s]
Figtire 13.2. 7 Mlaxitum lhot rod temnperatulre afier 1.4Y,)Aklk pulse at 1.3 MWf (LI'U BOI) Analysis 202 August 2008 UWNR LEULEU Conversion Conversion Analysis 202 August 2008
Core Power after Blades SCRAM at 2s following 1.4% Ak/k pulse at 1.3 MW (LEU BOL) I.E+06 1,E+05 0o. 1,.E+04 1.E+03 0 10 20 30 40 5o 60 70 80 90 100 Time [s] Figure 13.2,8 Core power after blades SCRAM in at 2 secondvs./ilowing l.4%Ak/k pulse at 1.3 MW (LEU OL) Maximum Hot Rod Temperature after Blades SCRAM at 2s following 1.4% Ak/k Pulse at 1.3 MW (LEU BOL) 1200 2032 1000 905.65 1532 W" 800 M. E, 600
.1032 a,2 E 400
*Max Temp 532 200 f
ýLy na A
'. i~;yIT Yý irIrC ;+= ~
iln 0 T y * ,r 32 10 20 30 40 50 60 70 80 90 100 Time [s] Figure 13.2.9 Max fiiel temperature afier blades SCRA A4 at 2s fJlloig a. 1.4% ,Abkk pulse at 1.3 MW (LEU BOL) UWNR LEU Conversion Analysis 203 August 2008
The maximum fuel temperature during a 1.4%Ak/k pulse'. at 1.3 MW for the LEU BOL case is 905.65°C (1662.1 7 0F) 2.252 seconds. after initiation of the pulse. The peak fuel temperature is 244.35°C (437.83'F) below the fuel temperature safety limit of 11'50°C (2100'F). The maximum. pulse power during this transient was determined to be .1.69 GW at 0.0250 seconds from initiation of the pulse. While pulsing at 1.3 MW is clearly an accident, no fuel failure is predicted during the transient. For comparison, the maximum fuel temperature after the pulse at 0.0460 seconds is 805.55°C (1481.99°F) and the maximum fuel temperature after 2.252 seconds is 905.65 0 C (1622.17'F). These temperature surface plots are shown adjacent to each other as seen in Figure 13.2.10. 40 ........ .............. 800 40 ... ..... ......... 900 35 ~~700 35
, 800 F U 30 0_6000 - 30 o 0 o M
.E,20 E 20 500 o o
*400 15 15 C* 400 10 .* 300. 10
.-*300 200 200 01.0 i 01 2 0 12 Radial Distance from Fuel Centerline [cm] Radial Distance from Fuel Centerline [cm]
Figure 13.2.10- Fuel temperatureplotafier1.4 %iAklkpulse at 1.3 MWat 0.0460s (left) & 2.252s (right) (LEUBOL) UWNR LEU Conversion Analysis 204 August 2008 0
13.2.3 Temperature after Pulse for LEU MOL Using the same methodology as stated in 13.2. ], the temperature after a 1.4%Ak/k ($1,809) pulse at 1.3 MW for the LEU MOL core can be found. The power and energy plot for the first 2 seconds is shown in Figure 13.2.11 and the maximum fuel temperature for the first 2 seconds is shown in Figure 13.2.12. After the scrammable blades fall into the core at 2 seconds with a calculated MCNP5 worth of -8.498%Ak/k (-$10.979), the core power is plotted in Figure 13.2.13 and the maximum fuel temperature is shown in Figure 13.2.14. The delayed neutron fraction is calculated for LEU MOL to be 0.00774. Core Power and Energy after 1.4% Ak/k pulse at 1.3 MW (LEU MOL) 1,E+10 I _ _Power 1.Ef09 2.04E+09 Energy 1.E+09__.. . .. . .. .. . . . .. . . ... ... 1.E+08 1.E+07 1.E+06 1E+06 -r-- -- -. r 11,E,--
.E÷05 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2 Time [s]
Figure 13.2.11 Power and total energy a/ier 1.4%,4k/k pulse at 1.3 MW (LEU MOL) 205 August 2008 Conversion Analysis UWNR LEU Conversion Analysis 205 August 2008
Maximum Hot Rod Temperature after 1.4 A%k/k'Pulse at 1.3 MW (LEU MOL) 1200 iF 2132 1100 1932 1000 1732
ý900 ...
C, 1532 0j 800 E E Ci 1332 700 600 I, - - Max Temp Safety Analysis Limit = 1150°C 1132 500 932 0 0.2 0.4 0.6 0.8 1 1.2 1.4 1.6 1.8 2 Time [si
...... [s ]
Figure 13.2.12 Maximum hot rod temperature after 1.4%oAk/k pulse at 1.3 MW (LEUMOL) Core Power after Blades SCRAM at 2s following 1.4% Ak/kpulse at 1.3MW(LEUMOL) 1.E+07 1.E+06 1.E+05 0 CL 1.E+04 1.E+03 0 10 20 30 40 50 60 70 80 90 100 Time [s] Figure 13.2.13 Core power after blades SCRAM in at 2 seconds following 1.4Ik/k pulse at 1.3 MW (LEU MOL) 206 August 2008 UWNR LEU Conversion Analysis Analysis 206 August 2008
Maximum Hot Rod Temperature after Blades SCRAM at 2s following 1.4%Ak/k Pulse at 1.3 MW (LEU MOL) 1200 976.05 800 S0 0 1032 E________________E E 400 - 0 MaxTemp 532
-Safety Analysis Limit = 1150*C 0 ,32 0 10 20 30 40 50 60 70 80 90 100 Time [s]
Figure 13.2.14 Max fuel temperatureafter blades SCRAM at 2s fillowing a. 1.4%zlk/k pulse at 1.3 MW (LEU MOL) The maximum fuel temperature during a 1.4%Ak/k pulse at 1.3 MW for the LEU MOL case is 976.05-C (1788.89-F) 2.252 seconds after initiation of the pulse. The peak fuel temperature is 173.95-C (311.1 I'F) below the fuel temperature safety limit of 1I50'C (2100'F). The maximum pulse power during this transient was determined to be 2.04 GW at 0.0250 seconds from initiation of the pulse. While pulsing at 1.3 MW is clearly an accident, no fuel failure is predicted during the transient. For comparison, the maximum fuel temperature after the pulse at 0.0485 seconds is 854.15'C (1569.47°F) and the maximum fuel temperature after 2.252 seconds is 976.05'C (1788.89'F). These temperature surface plots are shown adjacent to each other as seen in Figure 13,2.15. UWNR LEU Conversion Analysis 207 August 2008
40 [ . . . . . .. . . ........... 40 ... .......... ................ 40 00 800 900 35 35
-700, 30 0700 E 25 E 25 0 600 0320 50 o20 E E o 00 is 400 15 Cu (1@ 0 10 15210 7-300
*
- 300.
5 NO 200 200
- 0. 0 '
0 1 2 0 1 2 Radial Distance from Fuel Centerline [cm] Radial Distance from Fuel Centerline [cm] Figure 13.2 15 Fuel temperatureplot qfter 1.4 9o0/Ž,kk pulse at 1.3 MW at 0.0485s (left) & 2.252s (right) (LEU MOL) 13.2.4 Temperature after Pulse forLEU EOL Using the same methodology as stated in 13.2. 1, the temperatureafter a 1.4%Aklk ($1.895) pulse at 1.3 MW for the LEU EOL core can be found. The power and energy plot for the first 2 seconds is shown in Figure 13.2.16 and the maximum fuel temperature for the first 2 seconds is shown in Figure 13.2.17. After the scrammable blades fall into the core at 2 seconds with a calculated MCNP5 worth oi'-8.756%Ak/k (-$11.850), the core power is plotted in Figure 13.2.18 and the maximum fuel temperature is shown in Figure 13.2.19. The delayed neutron fraction is calculated for LEU EOL to be 0.007389. 208 August 2008 UWNR LEU U*NR Conversion Analysis LEU Conversion Analysis 208 August 2008
0 1.E+10 Core Power and Energy after 1.4% Ak/k pulse at 1.3 MW (LEU EOL) I.E+09 I .E+/-09 1.E+08 S11E+08 11E+07 0 I.E0 11E+06 1.E+06 I.E+05 0 .0.2 0.4 0.6 08 1 1.2 1.4 1.6 1.8 2
'Time Is)
Fiiure 13.2.16 Power and total energv after 1.4%ik/k pulse at 1.3 MW (LEU EOL) 0 Maximum Hot Rod Temperature after 1.4 A%k/k Pulse at 1.3 MW (LEU EOL) 1200 2132 1100 1932 1.732 1000 900 --. --- ---- . . -....... 800.- 1532 E E 60; 1332 700 Max Temrp 1 1132 600 [s]. - - Safety Analysis Limit = 1150'C 500 .. . ..... . 4 .. ....
....... ... ............ . *.................... ..... 932 0 0.2 0.4 0.6 0.8 1.2 1.4 .1 6 1,8 2 Time Time (s].
-"igure 3.2.17
/ Mat(ximum hot rod temperature a/ter I. 4%*IAk/k pulse at 1.3 MW (LEU EOL)
U\VNR LEU Conversion Analysis 209 August 2008
Core Power after Blades SCRAM at 2s following 1.4% Ak/k pulse at 1.3 MW (LEU EOL) I 1.E+06 I.E+05 0 I.E+04 -f-- 1.E+03 0 10 20 30 40 50 60 70 80 90 100 Time [s] Figure 13.2.18 Core power after blades SCRAM in at 2 seconds following 1.4°/lJ k/k pulse at 1.3 MW (LEU EOL) Maximum Hot Rod Temperature after Blades SCRAM at 2s following 1.4% Ak/k Pulse at 1.3 MW (LEU EOL) W 1200 2032
ý997.05 1000 1532 800 600 1----- M 1032 IV E 400 E
a, a,
-Max Temp 532 200 I ..-. -
- - Safety Analysis Limit= 1150°C 0 10 20 30 40 32 0 10 20 30 40 so 60 70 80 90 100
.Time [s)
Figure 13.2.19 Max fuel temperatureaoier blades SCRAM at 2s.following a./, 4%Ak/k pulse a0i. 3 MW (I.EU EOL) UWNR LEU Conversion Analysis 210 August 2008
The maximum fuel temperature during a 1.4%Ak/k pulse at 1.3 MW for the LEU EOL case is 997.05-C (1826.69'F) 2.127 seconds after initiation of the pulse. The peak fuel temperature is 152.95°C (273.317F) below the fuel temperature safety limit of 1150°C (2100'F). The maximum pulse power during this transient was determined to be 2.57 GW at 0.024 seconds from initiation of the pulse. While pulsing at 1.3 MW is clearly an accident, no fuel failure is predicted during the transient. For comparison, the maximum fuel temperature after the pulse at 0.0.46 seconds is 875.85°C (1608.53°F),and the maximum fuel temperature after 2.127 seconds is 997.05°C (1 826.69'F). These temperature surface plots are shown adjacent to each other as seen in Figure 13.2.20. 40 .. .... ..... . . .. . . . . 40 . ......... 358035 900 35 35 E 2 U- 800 3o *3 0o S25 E 25 0700 A20 0c 20 EE 0500 u 15 400 15:
.400
" 10 10
"'* "*200 200 0 1 2 0 1 2 Radial Distance from Fuel Centerline [cm] Radial Distance from Fuel Centerline [cm]
Figure 13.2.20 Fuel temperature plot after 1.4 %Akk pulse at 1.3 MW at 0.0460s (left) & 2. 127s (right) (LEU EOL) UWNR LEU Conversion Analysis 211 August 2008
It is apparent from the analysis performed for this accident scenario that LEU EOL is the most limiting stage of core lifetime, because it has the lowest negative fuel temperature coefficient, leading to a larger pulse and thus a higher fuel temperature. For all stages in core life, the pulse at, full power will not exceed the fuel temperature safety limit. While pulsing at 1.3 MW is clearly an accident, no fuel failure is expected. 13.3 Reduction-in-Cooling Accidents Although there is little likelihood of complete loss of water from the reactor pool, an analysis is made to demonstrate that such loss will not damage reactor fuel. 13.3.1 Possible Means of Water Loss The LEU conversion will not change the assumed means of pool water loss, which would be a sheared and open beam port. However, :it is desired to update this, calculation to validate other analyses later in this conversion report. The pool water level is routinely maintained at least 20.875 ft above the top of the core (this is the low pool level alarm point) which corresponds to 21.5 ft (6.553 m) above the fuel center. However, tbr conservatism it is assumed that the water level is only 19 ft above the top of the core, or 19.625 ft above the fuel center. The pool has a surface area of 89.13 ft2 (8.281 m3/4). The inner beam port diameter is 0.5 ft (0.1524 m), and the beam ports are at core center. Using these values, the following equation 30 can be used to estirnate the time required to drain the pool. t 2 A=, [I1- Equation 13.3.1 nAi0 UWNR LEU Conversion Analysis 212 August 2008
where: td = time to drain pool to height h (s) g = acceleration due to gravity (32.174 ft/s 2) Ap = cross-sectional area of pool surface (ft2) 2 Ao = cross-sectional area of drain opening (ft ) Cd - discharge coefficient (0.6) h, = initial height of water above drain opening (ft) h = final height of water above drain opening (ft) The calculated drain time is 836s. The previous HEU SAR assumed a time of 400s, but the previous calculation neglected the discharge coefficient. A sheared and open beam port could drain the water level to mid-core height, but water would still be in contact with the fuel and would prevent excessive temperatures. Nevertheless, the following analysis assumes no water is in contact with the fuel. 13.3.2 Radiation Levels Due to Unshielded Core . Calculations of radiation levels at various points in the Reactor Laboratory were made assuming continuous operations at 1.02 MW. The core exposure was 50 MWd for BOL, 800 MWd for MOL, and 1800 MWd for EOL. Fission product inventories for the hottest pin were multiplied by the number of fuel elements (91 for the revised HEU analysis, 83 for the LEU analysis) then divided by the power peaking factor (1.602 for the revised HEU analysis, 1.607/1.598/1.567 for the LEU BOL/MOL/EOL analyses) to calculate the core-wide inventory. The decay time of 836s represents the time required to drain the pool (see section 13.3.1). Doses from direct and scattered radiation were calculated using the MCNP5 code. Results of the current calculations for the revised HEU and the LEU analyses are given in Table 13.3.1 along with numbers from the previous HEU SAR for comparison. 213 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 213 August 2008
Time Table 13.3.1 Calculatedradiationdose rates afterpool water is lost. Core Analysis Beam 3rd Floor Console Pool Curb Pool Top 0 After Port Classroom (R/hr) Over Core Behind Shutdown Floor (R/hr) (R/hr) Curb (R/hr) (R/hr) Previous HEU SAR 1.6 N/A 2.0 10,000 2.6 Revised HEU 7.28 6.88 10.0 11,200 41.8 10 So LEU BOL 5.53 6.07 7.61 9,100 Seconds 31.7 LEU MOL 5.69 6.11 7.83 9,290 32.6 LEU EOL 5.60 5.96 7.70 9,110 32.0 Previous HEU SAR N/A N/A N/A N/A N/A Revised HEU 4.89 4.14 6.72 7,430 28.1 836 Scn LEU BOL 3.64 3.57 5.03 5,890 Seconds 20.9 LEU MOL 3.83 3.65 5.28 6,140 22.0 LEU EOL 3.78 3.57 5.21 6,040 21.7 Previous HEU SAR 0.17 N/A 0.24 1,200 0.30 I Day Revised HEU LEU BOL 1.23 0.751 0.764 0.556 1.69 1.04 1,720 1,110 7.07 4.31 0 LEU MOL 0.963 0.669 1.33 1,400 5.53 LEU EOL 0.976 0.669 1.34 1,420 5.60 Previous HEU SAR 0.08 N/A 0.12 540 0.14 Revised HEU 0.682 0.482 0.938 992 3.92 I Week LEU BOL 0.349 0.324 0.481 561 2.00 LEU MOL 0.543 0.426 0.747 827 3.11 LEU EOL 0.556 0.427 0.766 841 3.19 Previous HEU SAR 0.03 N/A 0.04 140 0.04 Revised HEU 0.353 0.210 0.485 492 2.02 1 Month LEU BOL 0.133 0.110 0.183 206 0.764 LEU MOL 0.283 0.184 0.390 410 1.62 LEU EOL 0.296 0.187 0.407 424 1.70 UWNR LEU Analysis Conversion Analysis LEU Conversion 214 214 August 2008 August 2008 0
These levels are not too high to allow emergency repairs to be made. Facility emergency procedures cover the situation of pool water loss. LEU calculated doses are lower than HEU primarily because of the. higher physical density of the fuel which results in more radiation self-shielding. 13.3.3 Fuel Temperature after Loss of Pool Water In order to determine the fuel temperature during a LOCA; it is necessary to make a few appropriate assumptions. It is assumed the reactor is operating at 1.02 MW (I MW nominal power + 2% uncertainty) for 50 days of continuous operation. While previous accident analyses, such as the pulse at full power were performed at 1.3 MW, it is unreasonable to believe the reactor operators would operate the reactor beyond the scram set point for 50 days straight. The UWNR has never operated continuously, and thus continuous operation at 1.02 MW is still a very limiting assumption. In addition, the hot rod channel thermal hydraulic parameters were changed from previous analysis to be more physically representative of the actual hot rod channel. Previously, the hot rod channel assumed the channel had a limiting flow area due to the presence of the transient rod and still assumed the transient rod was producing power. In order to make the assumption more accurate, the flow area was changed from 4.7429 cm' (0.735 16 in2) to 5.0144 cm2 (0.77723 in 2) and the hydraulic diameter was changed from 1.66318 cm (0.65479 in) to 1.78143 cm (0.70135 in). This analysis still assumed the four quarter rod segments were powered by a rod with the same pin power peaking factor of D5 SW. The assumptions made for all previous analysis are 215 August 2008 UWNR LEU Conversion Analysis UWNR Analysis 215 August 2008
still valid so that the predicted maximum fuel temperatures and CHF values are bounding. For the LOCA, the change in assumption was made so that the actual conditions in the core would be modeled and then the accident scenario would occur. In the event of massive water loss, the reactor would be shut down after receiving the pool high/low alarm. Section 13.3.1 calculated it would take at least 836 seconds to uncover the fuel due to the LOCA. During the first 836 seconds the fuel would still be water cooled and the power of the core would be decreasing from the power of delayed neutron fission and decay heat. At 836 seconds it is assumed that all water cooling is lost and the fuel only has air cooling after this time. During the air cooled portion of the LOCA transient, it is necessary to understand the difference between fuel temperature safety limits in the presence of water vs. water. The fuel temperature safety limit in the presence of water is stated to be I 150'C (2100°F), because the temperature of the clad is at a lower temperature due to contact with the water. However, with air cooling transients, the fuel temperature safety limit is 950'C (1740'F), because the temperatureof the clad is assumed to be the same temperature as the fuel. The methodology for this difference is kindly provided by TRIGA International as shown below. The strength of the fuel element clad is a function of its temperature. The stress imposed on the clad is a function of the fuel temperature as well as the hydrogen-to-zirconium ratio, the fuel UWNR LEU Conversion Analysis
..... j 216 August 2008
burnup, and the free gas volume within the element. The analysis of the stress imposed on the clad and strength of the clad uses the following assumptions:
- 1. The fuel and clad are the same temperature.
- 2. The hydrogen-to-zirconium ratio is 1.6 for LEU 30/20 fuel.
- 3. A space one-eighth inch high within the clad represents the free volume within the element.
- 4. The reactor contains fuel that has undergone burnup equivalent to 54 MW-days for LEU 30/20 fuel.
- 5. Maximum operating temperature of the fuel is 600 0 C.
The fuel elementinternal pressure P is given by: P=Ph+Pfp+Pair where: Ph is the hydrogen pressure; Pfp is the pressure exerted by volatile fission products; and P,,i is the pressure exerted by trapped air. For the hydrogen-to-zirconium ratios greater than about 1.58, the equilibrium hydrogen pressure can be approximated by: K 19740.37 3 1.76 + 10 .3014 x 17 . Ph =exp Tk where: x is the ratio of hydrogen atoms to zirconium atoms, and Tk is the fuel temperature (K). The pressure exerted by the fission product gasses is given by: n RTk E E V 217 August 2008 UWNR UW'NR LEU Conversion Analysis LEU Conversion Analysis 217 August 2008
where: f is the fission product release fractionr; n
- is the number of moles of gas evolved per unit of energy produced E
(mol/MW-day); R is the gas constant (8.206x10-2 L-atmlmol-K); V is the free volume occupied by the gasses (L); and E is the total energy produced in the element (MW-days) The fission product release fraction is given by23 exp -1.34xI04 J f = 1.5x1O -5 +3.6x103 where: T0 is the maximum fuel temperature in the element during normal operation (K). The fission product gas production rate n, E varies slightly with power density. The value 1.19x10 3 mol/MW-day is accurate to within a few percent over the range from a few kilowatts per element to well over 40kW per element. The free volume occupied by thegases is assumed 0 to be a space of one-eighth inch (0.3 175cm) high atthe top of the fuel so that V = 0.3175 7r -ri2 where: ri is the inside radius of the clad (1.745cm), The LEU 30/20 fuel has been tested to 50% bumup so its capability is slightly less than FLIP fuel at 54 MW-days per element. As the fission product gas pressure is proportional to the energy released, assume that the fuel in the reactor has undergone maximum bum-up. UWNR LEU Conversion Analysis 218 August 2008 0
Finally, the air trapped within the fuel element clad will exert pressure 0 Pair = RTk24 where it is assumed that the initial specific volume of the air is 22.4 L/mol. Actually, the air forms oxides and nitrides with the zirconium, so that after relatively short operation the air is no longer present in the free volume inside the fuel element clad. The results confirm the conclusion of NUREG-1282 that the LEU 30/20 fuel has a safety limit of 950'C when.the clad temperature equals the fuel temperature. AA
+
-4 ' t rl
--7
-s t.J P. PG. ,L IL.LV.!
I - - r _ 1u Fl 7 7 ---- T I-0% .0.
"U DO w;') IOU 00 C MVE 14.11OR C, r.
Figure 13.3.1 Strength and applied stress as a function of temperaturefor 1.7 and 1.6 H-Zr TRIGA fuel U"R LEU Conversion Analysis 219 August 2008
In addition to the analysis presented by TRIGA International, the methodology for calculating the cladding strength as a function of temperature was performed in GA-9064, 3 1 and the safety fuel temperature limit in air of 950'C (1740'F) was determined in the TAMU 1979 SAR 32 . To determine the maximum fuel temperature during the LOCA transient, it is assumed that blades 1, 2, 3, and the transient rod SCRAM into the core 2 seconds after the pool level alarm is activated. The regulating blade is not inserted during a SCRAM and is assumed to not be manually inserted into the core. This causes a prompt drop in power, and the dominant source of power for the transient is delayed neutron fission power and decay heat. Using an 80 second delayed neutron period, the power from fission after the blades have dropped in can be calculated. At 836 seconds, the power in the core is determined predominately from the decay heat. Since the decay heat is determined from the previous steady state operation, the axial and radial power distributions are' identical to those used in steady state analysis. In addition, the core power peaking factors are also identical to those used in the steady state analysis. To accurately analyze the LOCA, the problem was split into three parts. These parts were:
- 1. 2-channel steady state model at 1.02 MW, 5.7912m (19 feet) of water and inlet temperature of 54.44°C (1307F)
- 2. 2-channel transient model where the loss of water is modeled by losing water pressure until 101.3 kPa achieved at 836s to simulate losing 5.7912m (19 feet of water).
- 3. 2-channel transient model where the water coolant has been replaced with air. It is assumed the inlet air temperature is 25°C (77°F).
U"R LEU Conversion Analysis 220 August 2008
To determine the overall power transient during the LOCA, the ORIGEN2 data used previously to determine the radiation levels in an unshielded core was also used to determine decay heat. Since the ORIGEN2 data was only constructed for the hot pin, the decay heat for the hot pin was multiplied by the number of rods in the core and divided by the hot rod power peaking factor at the respective time of core life to determine the total core power during the transient. In addition to the decay heat, there is additional fission power that needs to be added for the first part of the transientdue to the influence of delayed neutrons. The prompt negative jump due to the effect of control blades falling in is: Pffier bludes drop - P00) ( For HEU BOL these parameters are: Po = 1.02 MW
=0.753%
Pshutdown = -3.69 1%A k/K Pafiter blades drop= 172830.7831 W Pshutdown is computed by adding the shutdown margin without blade 3 and the regulating blade inserted and the worth of blade 3 as calculated by MCNP5. Following the prompt jump, the fission power decreases as a function of time as follows: Pfsi,,ir~(t) = Pafle blode,drop ex p At shutdown 221 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 221 August 2008
For HEU BOL the parameters are: Pafer blades drop = 172830.783 1 W Atshutdown is the time since SCRAM in seconds
= 80s The summation of the fission power and the decay heat gives the final total core power curve entered into the 2-channel RELAP5/MOD3.3 model, The LOCA power curves did not incorporate the positive reactivity effects of the rods cooling down. While this would increase the fission power, it is not anticipated to effect the total power curves computed for the air cooled transient. The total power curve for HEU BOL is shown in Figure 13.3.2. Further analysis will follow for LEU BOL, MOL, and EOL cases. A zoomed in graph of the total core power during the air cooled portion of the HEU BOL LOCA transient is shown in Figure 13.3.3 Total Core Power During LOCA Transient (HEU BOL) 1,OOE+O7 ----------------- .- -----
Steady State
-Water Cooled Portion of Transient
-Air Cooled Portion of Transient 1.00E+05 1OOE+04 I.OOEi03 -- --. - - - - ~ - -
-5000 5000 15000 25000 35000 45000 55000 65000 75000 85000 Time [s]
Fi*gre 13.3.2 Total core power used in LOCA starting at 1.02 MW (HEU BOL) UWNR LEU Conversion Analvsis j-.-- 222 August 2008 0
Total Core Power during Air cooled portion of LOCA Transient (HEU BOL) 20000 18 0 0 0 . . . . . .. . . . . . . . . . .................... ................................ ............... 16000 - 14000 - 12000 W 0000 3: 0 8000 2000 0 0 10000 20000 30000 40000 50000 60000 70000 80000 90000 Time [s] Figure 13.3.3 Total core power used/br air cooled portion of LOCA transient (HEU BOL)
- After inputting this power profile into each component of the transient, the maximum temperature during the transient can be calculated. During the water cooled portion of the transient, the temperature falls from the steady state temperature of 483.00°C (901.40'F) to 73.46°C (164.23°F) after 836s. At this point in time the water level would reach the top of the fuel, and the remaining water is assumed to vacate the core.
223 August 2008 UWNR LEU Conversion Analysis Analysis 223 August 2008
Maximum Hot Rod Temperature during LOCA Transient (HEU BOL) 1000 1832 900 - Steady State 1632 800142 700 Water Draining out of Pool Transient 1432 u700. 596.89 - CompleteAir Cooled Transient 1232 ?6 S600 S- - Air Cooled Safety Analysis Limit 1032 i 500 T-400 - _ __ ___ ___ __ 832 a) 300 --- ______.632 200............... - 432 100 232 0 , -r,-, 32
-5000 5000 15000 25000 35000 45000 55000 65000 75000 85000 Time [s]
Figure 13.3.4 Temperatureprofile during LOCA transient (HEU BOL) The maximum temperature for HEU BOL during the air cooled portion of the transient was calculated to be 596.89°C (1 106.407F) as seen in Figure 13.3.4. This temperature is below the 950'C (17407F) fuel temperature safety limit with air cooling. This temperature occurs 7,750 seconds (2.15 hours) from the initiation of the accident or 6,914 seconds (1.92 hours) since being in contact with only air. While clearly a LOCA is a significant accident, nodamage to the fuel is predicted. In addition to analysis peiformed for HEU BOL, it is necessary to show that the fuel temperature will not exceed 950'C (1 740'F) for the LEU core at BOL, MOL, and EOL. Since the LEU core has 8 less fuel rods, the expected fuel temperature in the hot rod will be higher than the HEU Analysis 224 August 2008 UWNR LEU Conversion Conversion Analysis 224 August 2008 0s
. core. Table 13.3.2 shows the input parameters used to analyze the LEU core at BOL, MOL, and EOL for the LOCA analysis. Continuous operation for 50 days was assumed for LEU BOL, the identical assumption used for the HEU BOL case. For LEU MOL and EOL, the continuous operation time was calculated to be the burnup at. that stage divided, by the steady state power. The burnup for.LEU MOL and EOL is 800 MWD and 1800 MWD respectively. Table 13.3.2 Input conditions to determine the power profile br the LOCA transient Core Configuration HEU BOL LEU BOL LEU MOL LEU EOL Steady State Power 1.02 MW 1.02 MW 1.02 MW 1.02 MW Infinite Operation Time 50 days 50 days 784.3 days 1764:7 days 13 0.753 % 0.782% 0.774% 0.7389% Pshutdown -3.691%Ak/k -3.593%Ak/k -3.902%Ak/k -5.114%Ak/k Patler bladcs drop 172,830.78 W 182,317.71 W 168,853.09 W 1'28,769.58'W T 80s 80s 80s 80s
- Power at the start of the air oled transient, t=836s 18,369.86 W 18,883.59 W :19,833.55 W 19,719.38 W Using the results shown in Table 13.3.2, the power profile for the LEU core at BOL, MOL, and EOL is shown in Figure 13.3.5 through Figure 13.3.7. These power profiles are then put into the RELAP input decks with the same methodology employed for the HEU BOL case. The maximum fuel temperature is shown in Figure 13.3.8 through Figure 13.3.10.
0 UWNR LEU Conversion Analysis 225 August 2008
Decay Heat for LOCA transient calculated by ORIGEN (LEU BOL) 1.OOE+07
-Steady State
-Water Draining out of Pool Transient 1.OOE+/-06 I-i-
-Air Cooled Portion of Transient 1.00E+05 3
0 0. 1.00E+04 - 1.00E+03 -
-5000 5000 15000 25000 35000 45000 ssooo 65000 75000 85000 Time [s]
Figure 13.3.5 Power Profile used in LOC'A starting at 1.02 MW (LEU BOL) Decay Heat for LOCA transient calculated by ORIGEN (LEUIMOL) 1.OOE+07
- Steady State
- Water Draining out of Pool Transient 1.006406.
- Air Cooled Portion of Transient I .006+05 - - -- - - - - ----- ---
0 1.006+04 - - . . -.-- 1.OOE+03
-5000 5000 15000 25000 35000 45000 55000 65000 75000 85000 Time [s]
Figure 13.3.6 Power Pro/ile used in LOCA starting at 1.02 MW (LEU MOL) 226 August 2008 Conversion Analysis UWNR LEU Conversion UWNR Analysis 226 August 2008
Decay Heat for LOCA transient calculated by ORIGEN (LEU EOL) 1.OOE+07
-Steady Sta te
-Water Draining out of Pool Transient 1.00 E06
-Air Cooled Portion of Transient 0 1.OOE+05
, 1- a CL 1.OOE+04 ...... ----
I.O OE+0 3 .. . . .- ... . . --- .....- . ..--- ... . .. - - - - -r .. ..- -- - -
-s000 5000 15060 25000 3s000 45000 55000 65000 75000 85000 Time [s]
Figure 13.3.7 Power Profile used in LOCA starting ot 1.02 MW (LEU EOL) Maximum Hot Rod Temperature during LOCA Transient (LEU BOL) 9000 1832 1632
-Ste ady State 648.37 - Water Draining out of Pool Transient 1432 700-AT *-Complete Air Cooled Transient 1232 E 0.
E U, 500 400 300
+ -
Air Cooled Safety Analysis Limit 1032 832 632 E 200 432 100 I.- 232 0 32
-5000 5000 15000 25000 35000 45000 55000 65000 75000 85000 Time [s]
Figure 13.3.8 Temperature profile during LOC)A transient (LEU BOL) UWNR LEU Conversion Analysis 227 August 2008
Maximum Hot Rod Temperature during LOCA Transient (LEU MOL) 1000 - 1832 900 - Steady State 1632 800695.1 Water Draining out of Pool Transient 1432 Z7700 *
. - Complete Air Cooled Transient 1232 W 600
- Air Cooled Safety Analysis Limit 1032 832 0 E
0L 400 - E 632 300 200 4 32 0 -32
-5000 5000 15000 25000 35000 45000 55000 65000 75000 85000 Time [s]
F'igure 13 3.9 Temperatureprofile during LOCA transient (LEU MOL) Maximum Hot Rod Temperature during LOCA Transient (LEU EOL) 1000 --- 1832 900 1632
-Steady State 800 1432 700:' " 679.15-
.. . . Water . . .C ...out of Pool
. . Draining Transienrt 3
- Complete Air Cooled Transient 1232 600 CL 500 . . .Air Cooled Safety Analysis Limit 1032 0) 832 400 - .... E.
E 632 U, 0J 300 -...... 200 432 100 -232 232
-5000 5000 15000 25000 35000 45000 55000 65000 75000 85000 Time [s]
Figure 13.3, I10Temperature profile during L OCA transient (LEU EOL) 228 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 228 August 2008
Table 13.3.3 Summary ofLOCA temperature results Core Configuration HEU BOL LEU BOL" LEU MOL LEU EOL Starting temp during steady 483.00°C 498.75°C 498.75"C 482.7 1°C state operation (901.40°F) (929.75-F) (929.75 0 F) (900.88-F) Temperature at end of water 73.46 0 C 74.87 0 C 75.73"C 74.81 C cooled transient (164.23°F) (166.77-F) (168.3 1°F) (166.66-F) Change in temperature from 409.54 0 C 423.88°C 423.02 0 C 407.900 C steady state to 836s (737.17°F) (762.88°F) (761.44-F) (734.22 0 F) Core Power at 836s 18,369.86 W 18,883.59 W 19,833.55 W 19,719.38 W Maximum temperature in 596.89 0 C 648.37 0 C 695.10 0 C 679.15 0 C hotrod during LOCA (I 106.40°F) (1199.07 °F) (1283.18 0 F) (1254.57 0 F) Time of maximum 7,750 sec 7,775 sec 8,450 sec 9,300 sec temperature in hot rod (2.15 hr) (2.16 hr) (2.35 hr) (2.58 hr) In summary, all of the LOCA temperature results are shown in Table 13.3.3. As can be seen, at no time is the predicted temperature greater than the fuel temperature safety limit in air. Even O with many limiting assumptions, such as continuous operation of the reactor, this analysis has confirmed that the reactor will not exceed 950'C (1740'F) during a LOCA. Furthermore, all of the preceding LOCA analyses assumed that the water would be completely drained from the pool resulting in a complete LOCA. However, another accident scenario is possible where the water level drops to the bottom of the beam port exposing 26.67 cm (10.5 in) of active fuel and leaving 11.43 cm (4.5 in) of active fuel submerged in water. This accident, known as a partial LOCA, is much more difficult to analyze, but analysis performed has demonstrated that the complete LOCA is more limiting than the partial LOCA3 3'34 . 229 August 2008 UWNR LEU U\VNR Conversion Analysis LEU Conversion Analysis 229 August 2008
13.3.4 Loss of Coolant Flow Not applicable; natural convection cooling. 13.4 Other Accidents The previous HEU SAR estimated the potential reactivity effect of several accident scenarios involving core. loading. For comparison, these reactivity values have been calculated for the LEU core using MCNP5, summarized in the table below. Table 13.4. / Fuel and reflector handling accident reactivity effects Accident Condition Previous HEU SAR LEU Calculated Reported % Ak/k % Ak/k Add I graphite reflector in maximum worth + 0.230 + 0.22 +/- 0.01 non-occupied space (E8) Add I fuel bundle in maximum worth non- + 0.77 + 0.67 +/- 0.01 occupied space (E8) Add I fuel bundle in place of reflector in N/A + 1.17 +-0.01 maximum worth space (F5) to make 22 bundle core 0 Add I fuel bundle laying horizontally on top of + 0.5 + 0.009 +/- 0.009 core Adding additional fuel bundles or reflectors violates facility procedures for fuel handling and is not credible. Nevertheless, all potential reactivity effects are less than 1.4 %Ak/k and are therefore bounded by the pulse accident analysis covered in section 13.2. Conversion to LEU fuel will not introduce other accidents not previously considered. Analysis UWNR LEU Conversion Analysis UAINR 230 230 August 2008 August 2008 0
14 TECHNICAL SPECIFICATIONS The following sections in the UWNR Technical Specifications will be revised for the LEU conversion, Unless otherwise noted, each section replaces in its entirety the section from the previous HEU SAR. Most changes are simply updating references to LEU 30/20 fuel and removing all references to mixed cores. These represent changes to the approved UWNR Technical Specifications as submitted with the 1973 SAR and revised through amendment number 16 dated August 30, 2006. The UWNR license renewal SAR submitted in 2000 has not yet been approved, and so is not referenced in this LEU conversion report. Changes are back-highlighted. 14.1.14 Fuel Element (Previous HEU SAR Technical Specification 1.14) A fuel element is a single TRIGA fuel rod of LEU*Q32typ. 14.1.18 Standard Core (Previous HEU SAR Technical Specification 1.18) This section will be deleted. 14.1.19 Mixed Core (Previous HEU SAR Technical Specification 1.19) This section will be deleted. 231 August 2008 UWNR LEU Conversion Analysis LW Conversion Analysis 231 August 2008
14.1.20 FLIP Core (Previous HEU SAR Technical Specification 1.20) This section will be deleted. 14.1.21 Operational Core (Previous HEU SAR Technical Specification 1.21) An operational core is a LE.U,7O/20 core for which the core parameters of shutdown margin, fuel temperature, power calibration, and maximum allowable reactivity insertion have been determined to satisfy the requirements of the Technical Specifications. 14.2.1 Safety Limits (Previous HEU SAR Technical Specification 2.1) Revise Specifications and Bases to read the following: Specifications
- a. The temperature in a TRIGA LFIJ 3/20 fuel element shall not exceed II 50'C under any conditions of operation.
The reactor power: level shall not exceed 1500 kW steady state under any conditions of operation. Bases A loss of integrity of the fuel element cladding could arise from a buildup of excessive pressure between the fuel moderator and the cladding if the fuel U"R LEU Conversion Analysis 232 August 2008
temperature exceeds the safety limit. The pressure is caused by air, fission product gases, and hydrogen from dissociation of the fuel moderator. The magnitude of this pressure is determined by the fuel moderator temperature and the ratio of hydrogen to zirconium in the alloy. The safety limit for TRIGA fuel is based on data which indicate that the stress in: the cladding due to hydrogen pressure from the dissociation of zirconium hydride will remain below the ultimate stress provided the temperature does not exceed I 1.50'C and the fuel cladding is water cooled (pages 3-1 to 3-23 of GA-9064). It has been shown by experience that operation of TRIGA reactors at a power level of 1500 kW will not result in damage to the fuel. Several reactors of this type have operated successfully for several years at power levels up to 1500 kW.
' c oO oil* Sa,S* ov S r by4ql
- L**
- rlv TI500,kW corren,*ds to a p~eaki* e temptratr* rae of Thus a Safety Limit on power level of 1500 kW provides an ample margin of safety for operation.
14.2.2 Limiting Safety System Setting (Previous HEU SAR Technical.Specification 2.2) Revise specification and basis to read the following: UWNR LEU Conversion Analysis 233 August 2008
Specification weasure4'ina ntimetdfelcen with ,a vi oe eaigf between 0_87 aind 1-.16, or SOOT as rreasured; iIenni.ýýliitedl fuiel clrnn ývtiapppvrpaig4tr-fýtas 1.116.
- b. The limiting safety system setting for reactor power level shall be 1.25 MW.
Basis The limiting safety system setting is a temperature which, if exceeded, shall cause a reactor scram to be initiated preventing the safety limit from being exceeded. a pint power peaking fac-or of t lst 0.8 see Technical Specification 143'4Ž~), IFiit& IFF thrcouple ~re ies 400QC ten th ndaiiiurn ri-el. _UW,F7 therniconpl reaches 500TC then the TM.1UMthwO~tO11prtLr COul b~e no ~gyrc itr than I 121 .ý? In &itr Cs, tli, %ý1d leav a mrngin to thle :fuel te~pr~uesfey ii of .25TC., w i's This marin u~ed to 5a~low altriatc L!,rmocoupkes to be useifi trhe IF: i ~case' he nixnunradin&ticjmoop4 Analysis 234 August 2008 UWNR Conversion Analysis LEU Conversion UWNR LEU 234 August 2008
A\nay~j i'n section4.7 of- tii; -'nAilss A stat at13'W the peak. fuel temperature in the core will be approximately *6C so that the limiting power level setting provides an ample safety, margin to accommodate errors in power level measurement and anticipated operational transients. In the pulse mode: of operation, the same limiting safety system. setting will apply. However, the'temperature channel will, have no effect on limiting the peak powers generated because of its relatively long time constant (seconds) as compared with the width of the pulse (milliseconds). In this mode, however, the temperature trip will act to reduce the amount of energy generated in the entire pulse transient by cutting off the "tail" of the energy transient in the event the pulse rod remains stuck in the fully withdrawn position. 14.3.2 Pulse Mode Operation (Previous HEU SAR Technical Specification 3.2) Revise the bases to read the following: Bases 103O'o'to the 830'C operationa lim~it recomm1ended byGencra1 Atomlics "Puling Tqjjpevrqture Limit for ThIGA LEQ'Fuel7,, GA-G260117 ,~ffbr 200d7). UWNR LEU Conversion Analysis 235 August 2008
convýerwin.
'_bi Se tola th 1(.,) tt~noi~gh nsru'teid Fel tlilenýei i Obi ~ective f &hth*S$n IJtj Sl)eificatiord The, flstrurnientd Fiuel Eleinient'wicl-ihprovides ii, sigiL1 totefe ~m pwr pak~sing sectior , tlat08 %ith ~ ol40),O iuh aSS tlat11 Analsi,, ir scir 4.7.6 that V'ti'tie Fhr ~ corc -ouajon vvitia pill anlyisal' iso sfiv~s ttt wýithlihIF .i.i a cor loaio Iwit a pllptnowc peaking UWNR LEU Conversion Analysis 236 August v
2008
. 14.5.1 Reactor Fuel (Previous HEU SAR Technical Specification 5.1) Revise the Specifications and Bases to read the following: Specifications. The individual unirradiated TR IYU3. fuel elements shall have the following characteristics: (I) Uranium content: maximum of *Wt-/ nihc,.toiam of1995 v~th~ f19.7 ra nu~iij23 (2) Hydrogen-to-zirconium atom ratio (in the ZrH,): nominal 1.6 H atoms to 1.0 Zr atoms with a maximum H to Zr ratio of 1.65. (3) Natural erbium content (homogeneously distributed): nominal .,,'Wt,1-(4) Cladding: 304 stainless steel, nominal 0.020 inch thick. Bases The fuel UWh cUiiatConn 237nluin o Auu950/qo, This increas in loading wo re,,l suld minaIncrease In powc dcjiityo'es da!l MO.
&-~n inr 1býIilcý] p ty~ f ., sgýs teduty 9iagin~by (ssan content-{of' Libou 5.6%'ý 1pss tharl ihle sg 1 yeSI, Vajeo,
.0W-%4:,(o.'VC the 0UWNR LEU Conversion Analysis 237 August 2008
tha -. ob Thai for'a snle i,wt q , rit'n fl, 114M iloe dernsity I~ byip~h~sf~ Themaximum hydrogen-to-zirconium ratio of 1.65 could result in a maximum stress under accident conditions in the fuel element clad about a factor of two greater than for a hydrogen-to-zirconium ratio of 1.60. This increase in the clad stress during an accident would not exceed the rupture strength of theclad. 14.5.2 Reactor Core (Previous HEU SAR Technical Specification 5.2) Specification (b) and the corresponding basis (b) will be deleted. Specification (a) and basis (a) will be revised to read the following: 0 Specifications
- a. The core shall be an arrangement of TRIGA LE (2 uranium-zirconium hydride fuel-moderator bundles positioned in, the reactor grid plate.
Bases
- a. COýV have been f r yeail a dOCUMCI 'edIT, je nhid n-g 02 el~liave als been optirat&ed a OQricaF ~AtOMICS an(d Tcxas A&Mt an their 'ýsiccessfiul opuraioa UWREUCovrsinAaceltc, m vlbli 238'or aiion-tealysis Auus 200 UWNR LEU Conversion Analysis 238 August 2008 0
ndiate
ýVlconln tht te LU 3/20 core wl aey.aif l 14.5.7 Reactor Pool Water Systems (Previous HEU SAR Technical Specification 5.7)
The basis for specification (a) will be revised as follows. The specification will not change, but has been reproduced below for clarity. Specifications
- a. The reactor core shall be cooled. by natural convective water flow.
Bases
- a. UWNR LEU Ahalsi 239 4Cnri8 alysist n0t8 covctv coln ofthe reactor co6re Is ufficient to .mahintm the ful I saeoni~tion p atoles a power.jevel-it 150 k51W' (thbe powr.aft 0UWNR LEU Conversion Analysis 239 August w
2008
This page is intentionally left blank. nayi 4 uut20 UWRLUCneso U"R LEU Conversion Analysis 240 August 2008
15 OTHER LICENSE CONSIDERATIONS 15.1 Prior Utilization of Reactor Components The conversion to LEU fuel at UWNR will use new, unirradiated, recently fabricated fuel that has been approved by the NRC 2. Existing reflector assemblies and control elements, as well as hydraulic irradiation experimental facilities, will be re-used in the LEU core. 15.2 License Conditions The conversion of UWNR to LEU fuel will require a license amendment to allow possession of the new LEU fuel. Before the conversion order is received, all TRIGA Standard fuel currently in storage will be removed from the facility. Therefore, the license possession limits will be
- amended to remove the TRIGA Standard fuel and to possess and use the new LEU fuel.
Specifically, paragraph 2.B. will be revised to read: Pursuant to the Act and Title 10 CFR, Chapter 1, Part 70, "Special Nuclear Materials," to receive, possess, and use, up to a maximum of 15.0 kilograms of contained uranium 235 at less than 20 percent enrichment in the form of non-power reactor fuel; and to possess, but not use, up to a maximum of 18.0 kilograms of contained uranium 235 at 70 percent enrichment in the form of non-power reactor fuel until it can be removed from the facility; and 150 grams of contained uranium 235 at enrichment above 90 percent in neutron detectors; and 16 grams of contained plutonium in a plutonium-beryllium neutron source in connection with operation of the reactor. Without exceeding the 241 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 241 August 2008
foregoing maximum possession limits, the specific categories of maximum limits are as follows: Maximum U-235 Maximum Pu % Enrichment Exempt Status* (1) 15.0 kg 19.75 Exempt 10 CFR 73.6(a) (2) 0.14 kg 93.22 Not Exempt (3) 0.01 kg 93 Not Exempt (4) 16 grams Exempt 10 CFR 73.6(c) (5) 13.5 kg 70 Exempt 10 CFR 73.6(b) (6) 4.5 kg 70 Not Exempt
*Material is exempt provided that it meets the requirements for exemption pursuant to the cited provisions of 10 CFR 73.
15.3 Decommissioning The conversion to LEU fuel will not impact the UWNR decommissioning plan. UWNR LEU Conversion Analysis 242 August 2008 0
O REFERENCES Cashwell, R. J. "Safety Analysis Report for the University of Wisconsin Reactor." April, 1973. 2 NUREG-1282, "Safety Evaluation Report on High-Uranium Content Low-Enriched Uranium-Zirconium Hydride Fuels for TRIGA Reactors," USNRC, August 1987. 3 General Atomics. "Safety and Accident Analyses and Report Texas A&M University Conversion from HEU to LEU Fuel." December, 2005. 4 General Atomics. "Safety Analysis for the HEU to LEU Conversion of the Washington State University Reactor." June, 2007. GA Document Number: TRD 070.07007 RGE 001. NRS Accession Number: ML080170058. 5 "MCNP-A General Monte Carlo N-Particle Transport Code, Version 5," LA-CP-03-0245, F. B. Brown, Ed., Los Alamos National Laboratory (2003). . 6 Paul W. Humrickhouse and Paul P. H. Wilson, 'Validating a Monte Carlo Model of the University of Wisconsin Nuclear Reactor with Operational Data," Nuclear Technology, Volume 155, Number 2, August 2006. 7 "The REBUS-MCNP Linkage," J. G. Stevens, Argonne National Laboratory (Draft). J.J. Duderstadt and L.J. Hamilton "Nuclear Reactor Analysis", John Wiley & Sons, 1976. 9 Bretscher, M., "P3fnllp for Water-Reflected Fresh Critical Cores of the Oak Ridge Research Reactor Reflected by Water," RERTR Program, Argonne National Laboratory, July 31, 1986.
'0 RELAP5/MOD3.3 Code Development Team, "RELAP5/MOD3.3 Code Manual Volume 1:
Code Structure, System Models, and Solution Methods." 2003, Idaho National Laboratory, Idaho Falls, Idaho.
'1Earl E. Feldman, "Fundamental Approach to TRIGA Steady-State Thermal-Hydraulic CHF Analysis," ANL/RERTR/TM-07-01, December 2007.
12 General Atomics, "Request for additional information: Texas A&M University Nuclear Science Center Reactor, Docket No. 50-128. December 2005. 243 August 2008 UWNR LEU Conversion Analysis LEU Conversion Analysis 243 August 2008
3 Groeneveld, D. C., et. al., "The CHF Lookup Table," Nuclear Engineering and Design, 2007, p. 1-24. 14 Bermath, L., "A Theory of Local Boiling Burnout and its Application to Existing Data," Chemical Engineering Progress Symposium. Series No. 30, Volume 56, pp. 95-116 (1960). 15 Oregon State University Radiation Center, "Safety Analysis for the Conversion of the Oregon State TRIGA Reactor from HEU to LEU Fuel," Docket No. 50-243, Corvallis, Oregon, November 2007. 16Bousbia-Salah, A., et al., "Assessment of RELAP5 Model, for the University of Massachusetts Lowell Research Reactor". Nuclear Technology & Radiation Protection, 2006. 21: p. 3-12. 17 Dunn, F.E., et al. "MNSR Transient Analyses and Thermal Hydraulic Safety Margins for HEU and LEU Cores Using RELAP5-3D Code." In 2007 International Meeting on Reduced Enrichment for Research and Test Reactors." 2007. Prague, Czech Republic. 18 Oregon State University Radiation Center, "Safety Analysis for the Conversion of the Oregon State TRIGA Reactor from HEU to LEU Fuel," Docket No. 50-243, Corvallis, Oregon, November 2007. 19 The RELAP5-3Do Code Development Team, RELAP5-3D© Code Manual, Version 2.4, Volume I, Section 4.5, INEEL-EXT-98-00834, Idaho National Laboratory, June 2005. 20 GA-C26017, "Pulsing Temperature Limit for TRIGA LEU fuel." General Atomics Inc., December 2007. 21 Vitiello, Brian J., "Replacing LEU Hot Rod With Fresh Fuel at MOL and EOL." UWNR Technical Report 2008-02 (August 2008). 22 "Uranium-Zirconium Hydride Fuels for TRIGA Reactors." General Atomics Report UZR-28 (1997). 23 Foushee and Peters, "Summary of TRIGA Fuel Fission Product Release Experiments." General Atomics Report GULF-EES-A 10801 (Sept. 197 1). 24 Eckerman & Ryman, "Federal Guidance Report No. 12: External Exposure to Radionuclides in Air, Water, and Soil," EPA-402-R-93-081 (1993). 25 Hawley, Et Al, "Analysis of Credible Accidents for Argonaut Reactors," NUREG/CR-2079 (1981). UWNR LEU Conversion Analysis 244 August 2008 0
26 Hawley, Et Al, "Credible Accident Analyses for TRIGA and TRIGA-Fueled Reactors," NUREG/CR-2387 (1982). 27 Turner, "Workbook of Atmospheric Dispersion Estimates,?' Second Edition 1994, CRC Press. 28 Email correspondence with Brian Merritt from Strobic Air Corporation, dated 11/1/2007. 29 Email correspondence with Brian Merritt from Strobic Air Corporation, dated 11/2/2007. Attached fan flow diagram generated for model TS2L150C12, speed 1170 rpm, density 0.075 lbs./cu. ft. and values calculated for fan flow of 9600 cfm. 30 Hunsaker and Rightmire, "Engineering Applications of Fluid Mechanics," McGraw-Hill Book Company, 1947 (pp. 69-70). 3' GA-9064, "Safety Analysis Report for the Torrey Pines TRIGA Mark 11 Rector." General Atomics, Inc., January, 5, 1970. 32 "Safety Analysis Report for the Nuclear Science Center Reactor Texas A&M University." June 1979.
- Vitiello, Brian J., "Thermal Hydraulic and Safety Analysis of the University of Wisconsin Nuclear Reactor." University of Wisconsin-Madison, August 2008. (Master's degree thesis) 34 Dunn, F.E. et al, "Analyses of Loss-of-Coolant Accidents for TRIGA-Fueled Research Reactors at Washington. State University, the University of Wisconsin, and Oregon State University." Argonne National Laboratory DRAFT report ANL/RERTR/TM-08-03 (August, 2008).
Analysis 245 August 2008 UWNR LEU Conversion Analysis LEU Conversion 245 August 2008}}