ML23097A087

High Burnup Fuel Source Term Accident Sequence Analysis

ML23097A087
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Issue date:	04/14/2023
From:	Albright L, Brooks D, Shawn Campbell, Faucett C, Gilkey L, Humphries L, Keesling D, Luxat D, Phillips J, Katie Wagner NRC/RES/DSA, Sandia
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ML23097A085	List: ML23097A087 ML23102A005
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SAND2023-01313
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SANDIA REPORT SAND2023-01313 Printed April 2023 High Burnup Fuel Source Term Accident Sequence Analysis L.I. Albright, L. Gilkey, D. Keesling, C. Faucett, D.M. Brooks, K.C. Wagner, L.L.

Humphries, J. Phillips, D.L. Luxat Prepared by Sandia National Laboratories Albuquerque, New Mexico 87185 and Livermore, California 94550

Issued by Sandia National Laboratories, operated for the United States Department of Energy by National Technology & Engineering Solutions of Sandia, LLC.

NOTICE: This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government, nor any agency thereof, nor any of their employees, nor any of their contractors, subcontractors, or their employees, make any warranty, express or implied, or assume any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represent that its use would not infringe privately owned rights. Reference herein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government, any agency thereof, or any of their contractors or subcontractors. The views and opinions expressed herein do not necessarily state or reflect those of the United States Government, any agency thereof, or any of their contractors.

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ABSTRACT To extend NUREG-1465 recommendations, representative radiological releases to containment -

patterned after NUREG-1465 - have been evaluated for LWRs utilizing high burnup fuel with enrichments between 5-8% and 5-10% for PWRs and BWRs, respectively. Representative radionuclide releases are generated by applying non-parametric bootstrap methods to MELCOR simulation results spanning multiple accident scenarios that lead to significant core damage and in-containment source terms. In-containment source terms presented in this analysis are generally consistent with NUREG-1465 except where the state-of-knowledge and/or modeling state-of-practice have significantly evolved. Such differences include longer in-vessel phase durations accompanied by larger radionuclide releases to containment. Larger releases are observed, in part, because of increased sampling of low-pressure accident scenarios relative to NUREG-1465; low-pressure accident scenarios are known to cause larger releases to containment. Finally, this analysis demonstrates that in-containment source terms are essentially unchanged by increased burnup or elevated enrichment and that the most significant variation in source term continues to arise from differences between accident scenarios.

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ACKNOWLEDGEMENTS The authors would like to thank Fred Gelbard (SNL) for his review of this work. The authors also express gratitude to Mohsen Khatib-Rahbar (Energy Research, Inc.) for his efforts and supplemental sensitivity studies that supported this work.

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TABLE OF CONTENTS Abstract ............................................................................................................................................3 Acknowledgements ..........................................................................................................................4 Table of Contents .............................................................................................................................5 Table of Figures ...............................................................................................................................7 Table of Tables ..............................................................................................................................11 Executive Summary .......................................................................................................................14 Acronyms and Definitions .............................................................................................................23

1. Introduction ..............................................................................................................................25 1.1. Study Motivation ............................................................................................................25 1.2. Regulatory Use of Source Terms ....................................................................................25 1.2.1. TID-14844 Source Terms ...................................................................................29 1.2.2. NUREG-1465 Source Terms ..............................................................................30

2. Objectives and Scope ...............................................................................................................35 2.1. Alternative Source Term .................................................................................................36 2.2. Selected Reactors for Analysis .......................................................................................37 2.2.1. PWR with Large-Dry Containment - Surry .......................................................37 2.2.2. PWR with Ice Condenser Containment - Sequoyah ..........................................38 2.2.3. BWR Mark I - Peach Bottom .............................................................................38 2.2.4. BWR Mark III - Grand Gulf ..............................................................................39 2.3. Alternative Source Term Reactor Cores Analyzed .........................................................39 2.3.1. BWR Core Specification.....................................................................................39 2.3.2. PWR Core Specification .....................................................................................43 2.3.3. Isotopic Inventories .............................................................................................47

3. Methodology and assumptions ................................................................................................51 3.1. Overall Study Methodology............................................................................................51 3.2. General MELCOR Modeling Approach .........................................................................52 3.2.1. Reactor Coolant System or Nuclear Steam Supply System Failure Modes .......52 3.2.2. Treatment of Fuel Degradation and Relocation ..................................................58 3.2.3. Lower Plenum Debris/Coolant Heat Transfer and RPV Lower Head Failure....58 3.2.4. Core Plate Failure ...............................................................................................59 3.2.5. Ex-Vessel Molten Corium-Concrete Interaction (MCCI) ..................................59 3.2.6. Containment Failure Modeling ...........................................................................59 3.2.7. Fission Product Release Kinetics ........................................................................66 3.2.8. Cesium and Iodine Chemical Forms ...................................................................71 3.3. Accident Tolerant/High Burnup Fuel Severe Accident Phenomena Identification and Ranking Table ...........................................................................................................72 3.4. Non-parametric Approach ..............................................................................................74 3.5. Additional Methodological Assumptions .......................................................................75

4. Accident Selection ...................................................................................................................79 4.1. Selection of Representative Accident Scenarios.............................................................79 4.2. BWR Accident Scenarios ...............................................................................................83 4.2.1. BWR Mark I Containment Accident Scenarios ..................................................83 5

4.2.2. BWR Mark III Containment Accident Scenarios ...............................................85 4.3. PWR Accident Scenarios ................................................................................................87 4.3.1. PWR Large-Dry Containment Accident Scenarios ............................................87 4.3.2. PWR Ice Condenser Containment Accident Scenarios ......................................88 4.4. NUREG-1465 Accident Scenarios .................................................................................90 4.4.1. Boiling Water Reactors .......................................................................................91 4.4.2. Pressurized Water Reactors ................................................................................92

5. Results ......................................................................................................................................93 5.1. High Burnup, High Assay Low-Enriched Uranium Fuel ...............................................96 5.1.1. Gap Release Phase ............................................................................................100 5.1.2. Early In-Vessel Phase .......................................................................................106 5.1.3. Late In-Vessel Phase .........................................................................................109 5.1.4. Ex-Vessel Phase ................................................................................................111 5.2. Additional Sensitivity Analyses....................................................................................112 5.2.1. Fuel Thermal Conductivity ...............................................................................114 5.2.2. Porosity of In-Vessel Particulate Debris ...........................................................117 5.2.3. Diameter of In-Vessel Particulate Debris .........................................................121 5.2.4. Particulate Debris Falling Velocity ...................................................................123 5.2.5. Fuel Rod Lifetime .............................................................................................125 5.2.6. Fuel Relocation Temperature ............................................................................129 5.2.7. Hot Leg Creep Rupture .....................................................................................133 5.3. Release Rates ................................................................................................................135 5.4. Aerosol Retention and Removal Mechanisms ..............................................................139

6. Summary ................................................................................................................................142 References ....................................................................................................................................146 Appendix A. Plant-Independent MELCOR Modeling practices employed in this analysis ..149 Appendix B. HBU Accident Progression and Source Term Reference Analyses .................151 Appendix C. Accident Sequence Event timing tables ............................................................232 Appendix D. Supporting source term tables ...........................................................................260 Distribution ..................................................................................................................................266 6

TABLE OF FIGURES Figure ES-1 Representative early in-vessel radionuclide release fractions to containment.

Core types: (1) 60 GWd/MTU LEU, (2) 80 GWd/MTU LEU, (3) 60 GWd/MTU HALEU, (4) 80 GWd/MTU HALEU..................................................................................................... 16 Figure 1-1 Radionuclide transport paths for LWR designs (U.S. Nuclear Regulatory Commission, 2020). ................................................................................................................ 28 Figure 1-2 Role of Accident Progression, Source Term, and Consequence Analysis Computer Codes and Applicable Regulatory Requirements (U.S. Nuclear Regulatory Commission, 2020) ................................................................................................................. 29 Figure 2-1 BWR Fractional Decay Heat Variation with Cooling Time (Cumberland, Sweet, Mertyurek, Hall, & Wieselquist, 2021) .................................................................................. 42 Figure 2-2 BWR Decay Heat Relative to Reference Fuel (Cumberland, Sweet, Mertyurek, Hall, & Wieselquist, 2021) ..................................................................................................... 43 Figure 2-3 PWR Decay Heat Relative to Reference Fuel (Hall, Cumberland, Sweet, &

Wieselquist, 2021) .................................................................................................................. 45 Figure 2-4 PWR Fractional Decay Heat Variation with Cooling Time (Hall, Cumberland, Sweet, & Wieselquist, 2021) .................................................................................................. 46 Figure 3-1 Stages of in-containment source term Analysis ....................................................... 52 Figure 3-2 PWR Reactor Circulation System Natural Circulation Flows ................................. 57 Figure 3-3 BWR Mark I Containment Nodalization (Gauntt, Radel, Salay, & Kalinich, 2008) 60 Figure 3-4 BWR Mark I Reactor Building Nodalization (including containment failure flow pathways) (Leonard, Gauntt, & Powers, 2007) ...................................................................... 61 Figure 3-5 BWR Mark III drywell nodalization (Gauntt, Radel, Salay, & Kalinich, 2008) ..... 62 Figure 3-6 BWR Mark III Containment Nodalization (Gauntt, Radel, Salay, & Kalinich, 2008) 63 Figure 3-7 PWR Large-Dry Containment Nodalization............................................................ 64 Figure 3-8 PWR Large-Dry Containment Nodalization (Plan View) ....................................... 65 Figure 3-9 PWR Ice Condenser Containment Nodalization ..................................................... 66 Figure 3-10 RT-6 Release of Cesium as a function of test sample temperature (Gauntt, Goldman, Kalanich, & Powers, 2016) .................................................................................... 67 Figure 3-11 Booth Model Fits to RT-6 instantaneous diffusion coefficients (Gauntt, Goldman, Kalanich, & Powers, 2016) .................................................................................... 69 Figure 3-12 Comparison of RT-6 Cs Fractional Release Measurements against HBU and LBU Booth Model Predictions (Gauntt, Goldman, Kalanich, & Powers, 2016).................... 70 Figure 3-13 MELCOR Vapor Pressure curves for RN classes CsM (includes Cs2MoO4), CsI, and Mo .................................................................................................................................... 72 Figure 4-1 Peach Bottom Event Progression for Characteristic Accident Scenarios ................ 85 Figure 4-2 Grand Gulf Event Progression for Characteristic Accident Scenarios .................... 87 Figure 4-3 Surry Event Progression for Characteristic Accident Scenarios ............................. 88 Figure 4-4 Sequoyah Event Progression for Characteristic Accident Scenarios....................... 90 Figure 5-1 Non-parametric bootstrap of early in-vessel phase duration ................................... 94 Figure 5-2 MELCOR calculated fractional release to containment during the gap release phase. Releases for noble gases, halogens, and alkali metals are all less than NUREG-7

1465 recommendations. Releases for Te group, Ba/Sr group, and Mo group are greater than NUREG-1465 recommendations. ................................................................................. 102 Figure 5-3 MELCOR calculated accident phase durations. Phase durations for the gap release, early in-vessel, and late in-vessel phases are generally greater than NUREG-1465 recommendations. Ex-vessel phase durations are consistent with or less than NUREG-1465 recommendations. ........................................................................................................ 103 Figure 5-4 MELCOR calculated fractional release to containment during early in-vessel phase. Releases for noble gases are comparable to NUREG-1465 recommendations.

Releases for Ba/Sr group, lanthanides, and Ce group radionuclides are less than NUREG-1465 recommendations. Releases for halogens, alkali metals, Te group, Ru group and Mo group radionuclides are generally greater than NUREG-1465 recommendations. .............. 107 Figure 5-5 MELCOR calculated fractional release to containment during late in-vessel phase. Releases for noble gases, Te group, Ba/Sr Group, Ru group, and Mo group radionuclides are generally greater than NUREG-1465 recommendations. Releases for halogens and alkali metals are generally greater than NUREG-1465 recommendations for BWRs and less than NUREG-1465 recommendations for PWRs. ....................................... 110 Figure 5-6 MELCOR calculated fractional release to containment during ex-vessel phase.

Releases for noble gases are greater than NUREG-1465 recommendations. Releases for halogens, alkali metals, Te group, Ba/Sr Group, Ru group, Mo group, lanthanides, and Ce group radionuclides are generally less than NUREG-1465 recommendations. .................. 112 Figure 5-7 Peak Fuel Temperature Transient in Peach Bottom Short-Term (left) and Surry (right) SBO reference simulation and reduced fuel conductivity simulation. ...................... 115 Figure 5-8 Fission Product Release to Containment for Peach Bottom Short-Term SBO Fuel Thermal Conductivity Sensitivity ......................................................................................... 116 Figure 5-9 Fission Product Release to Containment for Surry SBO Fuel Thermal Conductivity Sensitivity ....................................................................................................... 117 Figure 5-10 Fission Product Release to Containment for Peach Bottom Short-Term SBO In-Vessel Debris Bed Porosity Sensitivity............................................................................ 119 Figure 5-11 Fission Product Release to Containment for Surry SBO In-Vessel Debris Bed Porosity Sensitivity ............................................................................................................... 120 Figure 5-12 Fission Product Release to Containment for Peach Bottom Short-Term SBO In-Vessel Particle Diameter Sensitivity ................................................................................ 122 Figure 5-13 Fission Product Release to Containment for Surry SBO In-Vessel Particle Diameter Sensitivity.............................................................................................................. 123 Figure 5-14 Fission Product Release to Containment for Peach Bottom Short-Term SBO Particle Debris Falling Velocity Sensitivity ......................................................................... 124 Figure 5-15 Fission Product Release to Containment for Surry SBO Particle Debris Falling Velocity Sensitivity .................................................................................................. 125 Figure 5-16 Fission Product Release to Containment for Peach Bottom Short-Term SBO Fuel Rod Lifetime Sensitivity ............................................................................................... 128 Figure 5-17 Fission Product Release to Containment for Surry SBO Fuel Rod Lifetime Sensitivity 129 Figure 5-18 Peak Fuel Temperature Transient in Peach Bottom Short-Term (left) and Surry (right) SBO Fuel Relocation Temperature Sensitivity ................................................ 131 Figure 5-19 Fission Product Release to Containment for Peach Bottom Short-Term SBO Fuel Relocation Temperature Sensitivity.............................................................................. 132 8

Figure 5-20 Fission Product Release to Containment for Surry SBO Fuel Relocation Temperature Sensitivity ........................................................................................................ 133 Figure 5-21 Fission Product Release to Containment for Surry SBO Hot Leg Creep Rupture Sensitivity................................................................................................................ 135 Figure B-1 Surry 1b Pressure Vessel collapsed water level for sequence 1b .................... 153 Figure B-2 Surry 1b Vessel Pressure ................................................................................. 154 Figure B-3 Surry 1b motor-driven auxiliary feedwater flowrate ....................................... 155 Figure B-4 Surry 1b Containment Pressure ....................................................................... 156 Figure B-5 Surry 1b Containment Temperature................................................................. 157 Figure B-6 SU1b Peak Core Temperatures ........................................................................ 158 Figure B-7 Surry 1b Core Damage Percentage .................................................................. 159 Figure B-8 Surry 1b Corium Ejection ................................................................................ 160 Figure B-9 Surry 1b Noble Gas Release Fraction to Containment .................................... 161 Figure B-10 Surry 1b Halogen Release Fraction to Containment........................................ 162 Figure B-11 Surry 1b Alkali Metals Release Fraction to Containment ............................... 163 Figure B-12 Surry 1b TE group release fraction to containment ......................................... 164 Figure B-13 Surry 1b Ba/Sr groups release fraction to containment ................................... 165 Figure B-14 Surry 1b Ru group release fraction to containment ......................................... 166 Figure B-15 Surry 1b Mo group release fraction to containment ........................................ 167 Figure B-16 1b Lanthanides release fraction to containment............................................... 168 Figure B-17 Surry 1b Ce group release fraction to containment ......................................... 169 Figure B-18 Sequoyah 4a mass flow rate through single loop failed RCP seal. .................. 171 Figure B-19 Sequoyah 4a TDAFW flow rate to single and lumped (triple) SGs. ............... 172 Figure B-20 Sequoyah 4a single and lumped (triple) SG levels. ......................................... 173 Figure B-21 Sequoyah 4a primary system pressures. .......................................................... 174 Figure B-22 Sequoyah 4a primary system pressures. .......................................................... 175 Figure B-23 Sequoyah 4a RPV Level .................................................................................. 176 Figure B-24 Sequoyah 4a Containment Pressure ................................................................. 177 Figure B-25 Sequoyah 4a Melt Fraction of Ice in Ice Condensers ...................................... 178 Figure B-26 Sequoyah 4a Peak Fuel Temperature in COR Rings 1 through 5.................... 179 Figure B-27 Sequoyah 4a Core Damage Fraction................................................................ 180 Figure B-28 Sequoyah 4a Noble Gases Release to Containment ........................................ 182 Figure B-29 Sequoyah 4a Halogens Release to Containment .............................................. 183 Figure B-30 Sequoyah 4a Alkali Metals Release to Containment ....................................... 184 Figure B-31 Sequoyah 4a Te Group Release to Containment ............................................. 185 Figure B-32 Sequoyah 4a Ba/Sr Group Release to Containment ........................................ 186 Figure B-33 Sequoyah 4a Ru Group Release to Containment ............................................. 187 Figure B-34 Sequoyah 4a Mo Group Release to Containment ............................................ 188 Figure B-35 Sequoyah 4a Lanthanides Release to Containment ......................................... 189 Figure B-36 Sequoyah 4a Ce Group Release to Containment ............................................. 190 Figure B-37 PB1a Core Water Level ................................................................................... 193 Figure B-38 PB1a Steam Dome Pressure............................................................................. 194 Figure B-39 PB1a Steam Dome Temperature ...................................................................... 195 Figure B-40 PB1a Drywell (Upper) Pressure....................................................................... 196 Figure B-41 PB1a Wetwell (Lower) Pressure ...................................................................... 197 Figure B-42 PB1a Drywell (Upper) Vapor Temperature ..................................................... 198 9

Figure B-43 PB1a Wetwell (Lower) Liquid Temperature ................................................... 199 Figure B-44 Peak Temperatures ........................................................................................... 200 Figure B-45 PB1a Core Damage .......................................................................................... 201 Figure B-46 PB1a Corium Ejection ..................................................................................... 202 Figure B-47 PB1a Noble Gases Release to Containment .................................................... 203 Figure B-48 PB1a Halogens Release to Containment.......................................................... 204 Figure B-49 PB1a Alkali Metals Release to Containment ................................................... 205 Figure B-50 PB1a Te Group Release to Containment ......................................................... 206 Figure B-51 PB1a Ba/Sr Group Release to Containment .................................................... 207 Figure B-52 PB1a Ru Group Release to Containment ......................................................... 208 Figure B-53 PB1a Mo Group Release to Containment ........................................................ 209 Figure B-54 PB1a Lanthanides Release to Containment ..................................................... 210 Figure B-55 PB1a Ce Group Release to Containment ......................................................... 211 Figure B-56 GG5a Core Water Level .................................................................................. 213 Figure B-57 GG5a Steam Dome Pressure ............................................................................ 214 Figure B-58 GG5a Steam Dome Temperature ..................................................................... 215 Figure B-59 GG5a Wetwell Pressure ................................................................................... 216 Figure B-60 GG5a Wetwell Temperature ............................................................................ 217 Figure B-61 GG5a Drywell Pressure ................................................................................... 218 Figure B-62 GG5a Drywell Temperature ............................................................................ 219 Figure B-63 GG5a Peak Core Temperatures ....................................................................... 220 Figure B-64 GG5a Core Damage Fraction........................................................................... 221 Figure B-65 GG5a Ejected Mass.......................................................................................... 222 Figure B-66 GG5a noble gases containment release fraction .............................................. 223 Figure B-67 GG5a Halogens containment release fraction.................................................. 224 Figure B-68 GG5a Alkali metals containment release fraction ........................................... 225 Figure B-69 GG5a Te group containment release fraction .................................................. 226 Figure B-70 GG5a Ba/Sr group containment release fraction ............................................. 227 Figure B-71 GG5a Ru group containment release fraction .................................................. 228 Figure B-72 GG5a Mo group containment release fraction ................................................. 229 Figure B-73 GG5a Lanthanides containment release fraction ............................................. 230 Figure B-74 GG5a Ce group containment release fraction .................................................. 231 10

TABLE OF TABLES Table ES-1 Phase durations (in hours) and release fractions for all core variations (60 GWd/MTU, 80 GWd/MTU, LEU and HALEU).* ................................................................. 18 Table ES-2 Representative phase durations (in hours) and release fractions with those recommended in NUREG-1465. Larger radionuclide releases are highlighted.* .................. 21 Table 1-1 Regulatory Guide 1.3rev2 and 1.4rev2 Compared with TID-14844 Source Term . 30 Table 1-2 NUREG-1465 Revised Radionuclide Groups ......................................................... 31 Table 1-3 NUREG-1465 Release Phase Durations .................................................................. 32 Table 1-4 NUREG-1465 recommended radionuclide release fractions for BWRs ................. 32 Table 1-5 NUREG-1465 recommended radionuclide release fractions for PWRs .................. 33 Table 2-1 Fuel Design Parameters from ORNL BWR Study (Cumberland, Sweet, Mertyurek, Hall, & Wieselquist, 2021) .................................................................................. 41 Table 2-2 Fuel Design Parameters from ORNL PWR Study (Hall, Cumberland, Sweet, &

Wieselquist, 2021) .................................................................................................................. 47 Table 2-3 BWR Isotopic Inventory across MELCOR Radionuclide Classes .......................... 48 Table 2-4 PWR Isotopic Inventory across MELCOR Radionuclide Classes .......................... 49 Table 3-1 HBU and LBU Booth Model Parameters ................................................................ 69 Table 3-2 Phase Timing Criteria .............................................................................................. 76 Table 4-1 Boiling Water Reactor High Burnup Accident Sequence Matrix............................ 81 Table 4-2 Pressurized Water Reactor High Burnup Accident Sequence Matrix ..................... 82 Table 4-3 Risk-Significance of NUREG-1465 BWR Accident Sequences (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) ................................................................................ 91 Table 4-4 Risk-Significance of NUREG-1465 Pressurized-Water Reactor Accident Sequences (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) ..................................................... 92 Table 5-1 Recommended BWR phase durations (in hours) and release fractions with uncertainties for all core variations (60 GWd/MTU, 80 GWd/MTU, LEU and HALEU).

Late in-vessel releases for halogen and Te group radionuclide releases are large relative to other radionuclide groups during that accident phase............................................................. 95 Table 5-2 Recommended PWR phase durations (in hours) and release fractions with uncertainties for all core variations (60 GWd/MTU, 80 GWd/MTU, LEU and HALEU). .... 96 Table 5-3 Comparison of recommended BWR phase durations (in hours) and release fractions for all core types. (1) - 60 GWd/MTU LEU, (2) - 80 GWd/MTU LEU, (3) - 60 GWd/MTU HALEU, and (4) - 80 GWd/MTU HALEU. ....................................................... 98 Table 5-4 Comparison of recommended PWR phase durations (in hours) and release fractions for all core types. (1) - 60 GWd/MTU LEU, (2) - 80 GWd/MTU LEU, (3) - 60 GWd/MTU HALEU, and (4) - 80 GWd/MTU HALEU. ....................................................... 99 Table 5-5 Comparison of Recommended BWR phase durations (in hours) and release fractions with those recommended in SAND2011-0128 and NUREG-1465. Darker shades

- larger than recommendations from both previous studies. Lighter shades - larger than NUREG-1465 or SAND2011-0128 release recommendations. ........................................... 104 Table 5-6 Comparison of recommended PWR phase durations (in hours) and release fractions with those recommended in SAND2011-0128 and NUREG-1465. Darker shades

- larger than recommendations from both previous studies. Lighter shades - larger than NUREG-1465 or SAND2011-0128 release recommendations. ........................................... 105 Table 5-7 In-vessel debris bed porosity sensitivity accident timings..................................... 118 11

Table 5-8 In-vessel particle diameter sensitivity accident timings ........................................ 121 Table 5-9 Reduced fuel rod lifetime function ........................................................................ 126 Table 5-10 Increased fuel rod lifetime function ................................................................. 126 Table 5-11 SOARCA fuel rod lifetime function ................................................................ 127 Table 5-12 Fuel rod lifetime sensitivity accident timings .................................................. 127 Table 5-13 Fuel relocation temperature sensitivity accident timings................................. 130 Table 5-14 Comparison of recommended BWR constant release rates with those of SAND2011-0128 and NUREG-1465. Darker shades - larger than recommendations from both previous studies. Lighter shades - larger than NUREG-1465 or SAND2011-0128 release recommendations. ..................................................................................................... 137 Table 5-15 Comparison of recommended PWR constant release rates with those of SAND2011-0128 and NUREG-1465. Darker shades - larger than recommendations from both previous studies. Lighter shades - larger than NUREG-1465 or SAND2011-0128 release recommendations. ..................................................................................................... 138 Table 5-16 Derived BWR release fractions including and excluding the suppression pool inventory for all core variations (60 GWd/MTU, 80 GWd/MTU, LEU and HALEU). ....... 140 Table C-1 Accident Sequence Event Timings for Surry, Case 1a .............................................. 232 Table C-2 Accident Sequence Event Timings for Surry, Case 1b .............................................. 233 Table C-3 Accident Sequence Event Timings for Surry, Case 1c .............................................. 234 Table C-4 Accident Sequence Event Timings for Surry, Case 1d .............................................. 235 Table C-5 Accident Sequence Event Timings for Surry, Case 1f .............................................. 236 Table C-6 Accident Sequence Event Timings for Sequoyah, Case 4a ....................................... 237 Table C-7 Accident Sequence Event Timings for Sequoyah, Case 4b ....................................... 238 Table C-8 Accident Sequence Event Timings for Sequoyah, Case 4c ....................................... 239 Table C-9 Accident Sequence Event Timings for Sequoyah, Case 4d ....................................... 240 Table C-10 Accident Sequence Event Timings for Sequoyah, Case 4e ..................................... 241 Table C-11 Accident Sequence Event Timings for Sequoyah, Case 4f ..................................... 242 Table C-12 Accident Sequence Event Timings for Sequoyah, Case 4g ..................................... 243 Table C-13 Accident Sequence Event Timings for Peach Bottom, Case 1a .............................. 244 Table C-14 Accident Sequence Event Timings for Peach Bottom, Case 1b .............................. 245 Table C-15 Accident Sequence Event Timings for Peach Bottom, Case 1c .............................. 246 Table C-16 Accident Sequence Event Timings for Peach Bottom, Case 1d .............................. 247 Table C-17 Accident Sequence Event Timings for Peach Bottom, Case 2a .............................. 248 Table C-18 Accident Sequence Event Timings for Peach Bottom, Case 2b .............................. 249 Table C-19 Accident Sequence Event Timings for Peach Bottom, Case 2c .............................. 250 Table C-20 Accident Sequence Event Timings for Peach Bottom, Case 3 ................................ 251 Table C-21 Accident Sequence Event Timings for Peach Bottom, Case 4 ................................ 252 Table C-22 Accident Sequence Event Timings for Grand Gulf, Case 5a .................................. 253 Table C-23 Accident Sequence Event Timings for Grand Gulf, Case 5b .................................. 254 Table C-24 Accident Sequence Event Timings for Grand Gulf, Case 5c .................................. 255 Table C-25 Accident Sequence Event Timings for Grand Gulf, Case 6a .................................. 256 Table C-26 Accident Sequence Event Timings for Grand Gulf, Case 6b .................................. 257 Table C-27 Accident Sequence Event Timings for Grand Gulf, Case 7 .................................... 258 Table C-28 Accident Sequence Event Timings for Grand Gulf, Case 8 .................................... 259 Table D-1 Comparison of BWR LEU phase durations and release fractions with uncertainties for all LEU core variations (60 GWd/MTU LEU, 80 GWd/MTU LEU). ...... 260 12

Table D-2 Comparison of PWR LEU phase durations and release fractions with uncertainties for all LEU core variations (60 GWd/MTU LEU, 80 GWd/MTU LEU). ...... 260 Table D-3 Comparison of BWR 60 GWd/MTU LEU phase durations and release fractions with uncertainties. ................................................................................................................. 261 Table D-4 Comparison of PWR 60 GWd/MTU LEU phase durations and release fractions with uncertainties. ................................................................................................................. 261 Table D-5 Comparison of BWR 80 GWd/MTU LEU phase durations and release fractions with uncertainties. ................................................................................................................. 262 Table D-6 Comparison of PWR 80 GWd/MTU LEU phase durations and release fractions with uncertainties. ................................................................................................................. 262 Table D-7 Comparison of BWR HALEU phase durations and release fractions with uncertainties for all HALEU core variations (60 GWd/MTU HALEU, 80 GWd/MTU HALEU)................................................................................................................................ 263 Table D-8 Comparison of PWR HALEU phase durations and release fractions with uncertainties for all HALEU core variations (60 GWd/MTU HALEU, 80 GWd/MTU HALEU)................................................................................................................................ 263 Table D-9 Comparison of BWR 60 GWd/MTU HALEU phase durations and release fractions with uncertainties. .................................................................................................. 264 Table D-10 Comparison of PWR 60 GWd/MTU HALEU phase durations and release fractions with uncertainties. .................................................................................................. 264 Table D-11 Comparison of BWR 80 GWd/MTU HALEU phase durations and release fractions with uncertainties. .................................................................................................. 265 Table D-12 Comparison of PWR 80 GWd/MTU HALEU phase durations and release fractions with uncertainties. .................................................................................................. 265 EmailInternal........................................................................................................................... 266 EmailExternal (encrypt for OUO) .......................................................................................... 266 13

EXECUTIVE

SUMMARY

To enhance safety as well as operational efficiency and economics, the U.S. nuclear power industry has undertaken an initiative to introduce new fuels into operating reactors. For imminent deployment, the industry has focused on new fuels that can achieve higher burnup. These types of fuels have been termed high burnup (HBU) fuel. To achieve these higher burnups, it is likely that HBU fuel will need to utilize higher enrichments. Maximum enrichments of 10% have been considered to achieve the higher burnups desired. These higher enrichment fuels are high-assay low enriched uranium (HALEU) fuels. Thus, the immediate operational plans of nuclear utilities in the United States include the loading of HALEU/HBU fuel into operating light water reactors (LWRs).

Regulatory source terms are developed based on the radionuclide release determined for postulated, accident scenarios and are deeply embedded in the U.S. Nuclear Regulatory Commissions (NRCs) regulatory policy and practices, reflecting the evolution of the current licensing process over the past 50 years. The licensing process is based on the concept of defense-in-depth, in which power plant design, operation, siting, and emergency planning comprise multiple levels of nuclear safety. This approach encourages nuclear plant designers to incorporate several lines of defense to maintain the effectiveness of physical barriers between radiation hazards and workers, members of the public, and the environment - for both normal operation and accident conditions.

This process of developing source terms was initially prescribed and defined in TID-18444 Calculations of Distance Factors for Power and Test Reactor Sites (U.S. Atomic Energy Commission, 1962). TID 18444 recommendations were based conservatively on assumptions of the maximum credible accident for a reactor postulated at that time. It has evolved to allow a mechanistic process as defined in NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants (U.S. Nuclear Regulatory Commission, 1995). Mechanistic source terms have been developed for a range of typical LWR fuel burnups, which are typically less than 40 GWd/MTU. An earlier, peer reviewed effort, in SAND2011-0128, Accident source terms for light-water nuclear power plants using high-burnup or MOX fuel, investigated the impact of increasing burnup on NUREG-1465 source terms (Powers, Leonard, Gauntt, Lee, & Salay, 2011)

(Energy Research, Inc., 2011). However, source terms existing prior to this report do not reflect the fuels being proposed for deployment in operating reactors in the U.S. as summarized in the NRC memorandum dated May 13, 2020, titled Applicability of Source Term for Accident Tolerant Fuel, High Burn Up and Extended Enrichment (Agencywide Documents Access and Management System (ADAMS) Accession Number ML20126G376) (U.S. Nuclear Regulatory Commission, 2020). This report details the development of such source terms following the methodology outlined in SAND 2011-0128 for operating LWRs loaded with the proposed fuels.

Analysis Objectives Evaluate the impact of increased burnup and enrichment on accident source terms.

The use of postulated source terms is an important feature of the regulatory practices and policies adopted by the NRC. Postulated source terms of this type consider the release of radionuclides to the containment, referred to as in-containment source term. Three previous in-containment 14

source term studies are of particular importance and relevance to this analysis and have been incorporated in various regulatory guides (e.g., Regulatory Guide 1.183):

• Technical Information Document (TID)-14844, Calculation of Distance Factors for Power and Test Reactor Sites, (1962).

• NUREG-1465, Accident Source Terms for Light-Water Nuclear Power Plants (February 1995).

• Peer-reviewed SAND2011-0128, Accident Source Terms for Light- Water Nuclear Power Plants Using High-Burnup or MOX Fuel (Powers, Leonard, Gauntt, Lee, & Salay, 2011).

Representative in-containment source terms developed in this analysis are developed and presented in a manner consistent with the NUREG-1465 in-containment source terms and SAND2011-0128. The results of this analysis serve as an update to the in-containment source terms reported in SAND2011-0128. The process used to develop representative in-containment source terms is as follows:

• Select a set of unmitigated accident scenarios, not reflecting B.5.b and Flex, that lead to significant core damage and in-containment source terms utilizing available BWR and PWR plant PRAs.

• Select a set of reactor plant models that are representative of a broad range of power plants operating across the United States.

• Develop radionuclide inventories and core decay heats for assumed core loading patterns (i.e., spatial fuel burnup and enrichment distributions, later referred to as core type) using the SCALE code package.

• Perform best-estimate MELCOR simulations for each core loading pattern, reactor plant model, and relevant accident scenarios that lead to significant core damage and in-containment source terms.

• Develop a distribution of containment fission product release magnitudes and durations at different phases of an accident using statistical methods and report the representative in-containment source term1.

• For consistency with the definition of in-containment source terms in RG 1.183 and previous in-containment source term analyses (NUREG-1465 and SAND2011-0128), the in-containment source term reported in this analysis include the entire radionuclide inventory within containment.

• In general, the MELCOR plant model, modeling approach, and best practices follow the approach established in SAND2011-0128 with enhancements developed during the SOARCA uncertainty project, as well as post 9/11, and post-Fukushima Daiichi regulatory studies.

• Phenomenological uncertainties identified in NUREG/CR-7283, Peer Review of Accident Source Terms for Light-Water Nuclear Power Plants Using High-Burnup and Mixed Oxide 1

This analysis uses the 50th percentile values, which was also used in SAND2011-0128. NUREG-1465 used the 75th percentile.

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Fuel, are interrogated in this report through a set of sensitivity calculations (Khatib-Rahbar, 2021).

Key Finding 1:

An increase in burnup and enrichment does not strongly impact the in-containment source term.2 Representative in-containment source terms developed in this analysis strongly indicate that accident sequence differences dominate variation in source term to containment. The variation in the in-containment source terms across the accident sequences is larger than the variation in the in-containment source terms across the core types. Figure ES-1 shows the limited variation in early in-vessel radionuclide release fractions of principally released radionuclide species for the different reactor and core types. As shown, an increase in burnup and enrichment does not strongly impact the in-containment source term in terms of radionuclide release fractions.

Changes to radionuclide inventories as a result of increased burnup and enrichment, however, may impact subsequent dose calculations and feasibility of HBU/HALEU implementation.

Figure ES-1 Representative early in-vessel radionuclide release fractions to containment. Core types: (1) 60 GWd/MTU LEU, (2) 80 GWd/MTU LEU, (3) 60 GWd/MTU HALEU, (4) 80 GWd/MTU HALEU.3 2

Improved modeling fidelity does change the fission product release fractions and timings relative to previous source term studies including NUREG-1465 and SAND2011-0128.

3 Each bar in the chart represents the representative in-containment source term for a given combination of reactor and core types considered in the analysis. BWR results consider the population of Peach Bottom and Grand Gulf datasets (that meet the associated core type specification), and PWR results consider the population of Surry and Sequoyah datasets (that meet the associated core type specification). LEU core types assume 5 wt% enrichment, HALEU core types assume 8 wt% enrichment for PWRs and 10 wt% enrichment for BWRs.

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Representative in-containment source terms for HALEU/HBU fuels with corresponding standard deviation expressed as percent uncertainty are presented in Table ES-1. Findings include:

• Semi-volatile4 and non-volatile radionuclide releases during the gap release phase.

Heterogeneous modeling of the reactor core in MELCOR allows overlap between gap release and early in-vessel phase phenomena during each phase. That overlap permits quantifiable semi-volatile and non-volatile radionuclide releases to occur during the gap release phase.

• Largest fractional releases (volatiles and semi-volatiles) during the early in-vessel phase. Most radionuclide releases to containment occur during the early in-vessel phase for both reactor types. This follows from the fact that the most significant period of core degradation and liberation of the fission products from both the fuel matrix and cladding fission product barriers occurs during the early in-vessel phase. Volatile and semi-volatile fission products are the primary contributors to containment releases during the early in-vessel phase.

• Restricted radionuclide release during the late in-vessel and ex-vessel phases.

Radionuclide release during the late in-vessel and ex-vessel phases for the noble gases are limited by remaining radionuclide inventories after the larger releases that occurred during the early in-vessel phase.

4 Semi-volatile radionuclides are defined here as radionuclides that become volatile under certain conditions (e.g.,

oxidizing conditions). Example semi-volatile radionuclide include Mo, Ba, and others.

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Table ES-1 Phase durations (in hours) and release fractions for all core variations (60 GWd/MTU, 80 GWd/MTU, LEU and HALEU).*

Gap Release Early In-vessel Late In-vessel Ex-vessel Phase Duration 0.70 6.7 44.6 3.1 Noble Gases 0.016 0.95 0.005 0.011 Halogens 0.005 0.71 0.16 0.017 Alkali Metals 0.005 0.32 0.021 0.009 BWR Te Group 0.003 0.56 0.19 0.003 Ba/Sr Group 0.0006 0.005 0.002 0.038 Ru Group <1.0E-6 0.006 7.9E-05 <1.0E-6 Mo Group 1.9E-05 0.12 0.002 2.3E-05 Lanthanides <1.0E-6 <1.0E-6 <1.0E-6 3.6E-05 Ce Group <1.0E-6 <1.0E-6 0.0 0.003 Gap Release Early In-vessel Late In-vessel Ex-vessel Phase Duration 1.3 4.0 24.0 1.9 Noble Gases 0.026 0.93 0.010 0.018 Halogens 0.007 0.58 0.031 0.020 Alkali Metals 0.003 0.50 0.013 0.015 PWR Te Group 0.006 0.55 0.019 0.005 Ba/Sr Group 0.001 0.002 0.0001 0.011 Ru Group <1.0E-6 0.008 5.4E-05 <1.0E-6 Mo Group 2.0E-05 0.15 0.002 0.002 Lanthanides <1.0E-6 <1.0E-6 <1.0E-6 1.4E-05 Ce Group <1.0E-6 <1.0E-6 <1.0E-6 0.0006

• This analysis uses the 50th percentile.

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A comparison of in-containment source terms developed in this analysis and recommended in NUREG-1465, are presented in Table ES-2. The following findings reflect the evolution in the severe accident knowledge base and modeling practices since NUREG-1465:

• Evolution of severe accident state-of-practice. Modern MELCOR models the distribution of heat in the core and heat removal from the core more efficiently than the STCP. MELCOR also models heterogeneous core degradation using a 2D representation of the reactor core and improved fuel failure models.

• Longer release phase durations. Longer phase durations are responsible, in part, for the larger releases observed during the early in-vessel phase.5

• Larger early releases for halogens, alkali metals, Te, and Mo group radionuclides.

Increased early releases are primarily a result of increased sampling of low-pressure accident scenarios; releases are known to be more significant for low-pressure accident scenarios.

Additional effects can be attributed to changes in the state-of-knowledge since the NUREG-1465 analysis. The prevailing expert opinion at the time predicted a greater degree of early in-vessel retention of Te due to a possible reaction with unoxidized Zircaloy cladding, and chemisorption onto steel structures. Similarly, Mo was assumed to have very low volatility as a non-reactive metal. More recent experimental findings including results from the Phébus-FP program (Clement & Haste, 2004) suggest more volatile chemical forms of both radionuclide groups.6 Key Finding 2:

The larger early releases to containment found in this study are a result of early primary pressure boundary failure.

o Occurrence of SRV seizure for BWRs during the early in-vessel phase. The SBO sequences considered in this present study for BWRs were predominately ones with a thermally induced seizure of the SRV prior to lower head breach. This resulted in the vessel achieving a low-pressure condition in the majority of the SBO scenarios considered. This has a notable effect on the ability of fission products generated during the early in-vessel phase to be released from the vessel into containment. Across the scenarios for which SRV seizure resulted in the vessel depressurizing prior to lower head breach, the early in-vessel CsI releases exhibited relatively higher releases to containment from the time the gap release phase ended to the time of lower head failure. By contrast, the much smaller set of SBO scenarios in which the vessel remained at high-pressure prior to lower head breach exhibited lower CsI releases over the time from cessation of the gap phase to lower head breachthe magnitude of CsI release for a condition with the vessel remaining at high-pressure is consistent with previous high-pressure accident sequence findings.

o HLCR occurrence for PWRs during the early in-vessel phase. Releases of the halogen, alkali metals, Te, and Mo group radionuclide classes are notably larger for PWRs because of a high incidence of hot leg creep rupture events during the early in-5 SAND2011-0128 generally predicted longer phase durations than NUREG-1465 as well.

6 SAND2011-0128 also predicted larger releases than NUREG-1465 during the early and late in-vessel release phases for halogens, Te group, and Mo group radionuclides.

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vessel phase. HLCR causes rapid depressurization of the primary system and low-pressure conditions prior to vessel breach. Both vaporized and unvaporized fission products travelling through the primary circuit are released to containment through HLCR sites increasing in-containment source terms during the early in-vessel phase. In the Surry SOARCA UA, HLCR was also observed in the majority of cases. The subset of simulations that do not exhibit HLCR and remain at high-pressure until vessel failure exhibit lower in-containment source terms for halogen, alkali metal, Te, and Mo group radionuclides as well as the other, non-volatile radionuclide classes - the magnitude of release among these classes reflect advancements in the state-of-knowledge (i.e., Te and Mo group releases are in-line with the current state-of-knowledge).

o Early primary pressure boundary failure sensitivity. As stated above, the timing of primary pressure boundary failure is known to impact containment releases. To investigate in-containment source term sensitivity to primary pressure boundary failure timing, a sensitivity calculation is performed with hot leg creep rupture suppressed -

SOARCA UA results indicate that hot leg creep rupture is the more likely event. The sensitivity calculation with suppressed hot leg creep rupture clearly shows significantly reduced in-containment source terms.

Key Finding 3:

Releases to containment are significantly reduced if the primary pressure boundary remains intact as in the previous studies.

• Enhancement in radiochemistry state-of-knowledge impacting early in-vessel phase releases. The state-of-knowledge of radiochemistry impacting radionuclide releases to containment has also advanced. For example, 100% of the I inventory is assumed to combine with Cs to form CsI; any remaining Cs mass is assumed to form Cs2MoO4. This is consistent with the findings of the Phébus-FP program (Clement & Haste, 2004), as adopted during SOARCA. Thus, the dominant forms of I and Mo releases are associated with chemical compounds that include Cs. Larger Te releases are also consistent with the findings of the Phébus-FP program (Clement & Haste, 2004), which observed more efficient transport of Te group materials to containment. Lanthanide and Ce group radionuclide classes remain below the threshold release fraction (1.0x10-6)7 for all simulations performed in this analysis.8

• Reduced ex-vessel releases. The contribution of the ex-vessel release phase to the overall in-containment source term is reduced. Large releases during the earlier, prolonged accident phases deplete most of the releasable volatile and semi-volatile fission product inventories from core debris.9 7

The threshold release fraction is a metric defined by the authors of this study to reflect release fractions too small to contribute substantively to in-containment source terms.

8 SAND2011-0128 incorporated the same state-of-knowledge enhancements and exhibited larger Te and Mo releases than NUREG-1465.

9 Ex-vessel releases were reduced in the SAND2011-0128 report for the same reasons.

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Table ES-2 Representative phase durations (in hours) and release fractions with those recommended in NUREG-1465. Larger radionuclide releases are highlighted.*

Gap Release Early In-vessel Late In-vessel Ex-vessel NUREG- NUREG- NUREG- NUREG-2021 2021 2021 2021 Report 1465 1465 1465 1465 Phase Duration 0.70 0.50 6.7 1.5 44.6 10.0 3.1 3.0 Noble Gases 0.016 0.050 0.95 0.95 0.005 0.0 0.011 0.0 Halogens 0.005 0.050 0.71 0.25 0.16 0.010 0.017 0.30 BWR Alkali Metals 0.005 0.050 0.32 0.20 0.021 0.010 0.009 0.35 Te Group 0.003 0.0 0.56 0.050 0.19 0.005 0.003 0.25 Ba/Sr Group 0.0006 0.0 0.005 0.020 0.002 0.0 0.038 0.10 Ru Group <1.0e-6 0.0 0.006 0.003 7.9E-05 0.0 <1.0e-6 0.003 Mo Group 1.9E-05 0.0 0.12 0.003 0.002 0.0 2.3E-05 0.003 Lanthanides <1.0e-6 0.0 <1.0e-6 0.0002 <1.0e-6 0.0 3.6E-05 0.005 Ce Group <1.0e-6 0.0 <1.0e-6 0.0005 0.0 0.0 0.003 0.005 Gap Release Early In-vessel Late In-vessel Ex-vessel Phase Duration 1.3 0.50 4.0 1.3 24.0 10.0 1.9 2.0 Noble Gases 0.026 0.050 0.93 0.95 0.010 0.0 0.018 0.0 Halogens 0.007 0.050 0.58 0.35 0.031 0.10 0.020 0.25 Alkali Metals 0.003 0.050 0.50 0.25 0.013 0.10 0.015 0.35 PWR Te Group 0.006 0.0 0.55 0.050 0.019 0.005 0.005 0.25 Ba/Sr Group 0.001 0.0 0.002 0.020 0.0001 0.0 0.011 0.10 Ru Group <1.0e-6 0.0 0.008 0.003 5.4E-05 0.0 <1.0e-6 0.003 Mo Group 2.0E-05 0.0 0.15 0.003 0.002 0.0 0.002 0.003 Lanthanides <1.0e-6 0.0 <1.0e-6 0.0002 <1.0e-6 0.0 1.4E-05 0.005 Ce Group <1.0e-6 0.0 <1.0e-6 0.0005 <1.0e-6 0.0 0.0006 0.005 th

• This analysis uses the 50 percentile values, which was also used in SAND2011-0128. NUREG-1465 used the mean release for volatiles and 75th percentile for non-volatiles.

Phenomenological uncertainty postulated to contribute to in-containment source term uncertainty is investigated for two broad categories of key phenomena identified in the HBU/HALEU Phenomena Identification and Ranking Table (PIRT), NUREG/CR-7283:

1. Alteration of thermophysical properties of the fuel and cladding at HBU - at increased burnup, the fuel thermal conductivity will decrease.

o Fuel thermal conductivity - A sensitivity calculation is performed at lower fuel thermal conductivity.

2. Fuel fragmentation and sintering at HBU - at increased burnup, fuel fragmentation and sintering is postulated. Increased fragmentation and sintering of fuel rods could result in earlier fuel degradation and failure.

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o Porosity of in-vessel particulate debris - Sensitivity calculations are performed with varied in-vessel particulate debris porosities.

o Diameter of in-vessel particulate debris - Sensitivity calculations are performed with varied in-vessel particulate debris diameters.

o Particulate debris falling velocity - A sensitivity calculation is performed at lower particulate debris falling velocity.

o Fuel relocation temperature - Sensitivity calculations are performed with varied fuel relocation temperatures. A sensitivity study was also performed with the eutectics model activated.

o Fuel rod lifetime - Sensitivity calculations are performed with varied fuel rod lifetimes.

In general, these sensitivity calculations show minimal impact on in-containment source terms.

Knowledge gaps identified in NUREG/CR-7283 are thus not found to have a significant impact on the overall results of this study. The variability in source term across accident scenarios is much larger than the variability in source term, for a given accident scenario, due to phenomenological uncertainty.

In summary, representative BWR and PWR source terms to containment for HALEU/HBU fuels have been developed using MELCOR consistent with the NUREG-1465 methodology.

Representative in-containment source terms do not strongly depend on increased burnup and enrichment. The occurrence of early primary pressure boundary failure in a broad range of scenarios led to larger early releases to containment, and smaller late releases to containment relative to previous analyses. Containment releases are significantly reduced if the integrity of the primary pressure boundary is maintained.

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ACRONYMS AND DEFINITIONS Abbreviation Definition AC Alternating current ADS Automatic depressurization system AFW Auxiliary Feedwater AST Alternative source term ATF Accident tolerant fuel ATWS Anticipated transient without scram BWR Boiling water reactor CCFL Counter current flow CDF Core damage frequency DC Direct current ECCS Emergency core cooling system ECDF Empirical cumulative distribution function GG Grand Gulf HALEU High-assay low-enriched uranium HBU High burnup HLCR Hot leg creep rupture HPCI High pressure coolant injection system HPSI High-pressure safety injection LBLOCA Large-break loss of coolant accident LEU Low-enriched uranium LOCA Loss of coolant accident LPCI Low-pressure coolant injection LPSI Low-pressure safety injection LTSBO Long-term station blackout LWR Light water reactor NRC Nuclear Regulatory Commission ORNL Oak Ridge National Laboratories PB Peach Bottom PIRT Phenomena Identification and Ranking Table PORV Pilot-operated relief valve PRA Probabilistic risk assessment PRT Pressurizer relief tank PWR Pressurized water reactor QoI Quantity of interest RCIC Reactor core isolation cooling system RCP reactor coolant pump RCS Reactor coolant system 23

Abbreviation Definition RHR Residual heat removal SBLOCA Small-break loss of coolant accident SBO Station blackout SOARCA State-of-the-Art Reactor Consequence Analyses SQN Sequoyah SRV Safety relief valve STCP Source Term Code Package STSBO Short-term station blackout SU Surry TDAFW Turbine-driven auxiliary feedwater TID Technical information document TMI-2 Three Mile Island Unit-2 24

1. INTRODUCTION 1.1. Study Motivation Recently, industry has initiated a process to introduce new fuels into operating reactors to enhance operational efficiency and economics. In the imminent deployment timeframe, industry is focused on employing new fuels that can achieve higher burnup, which are commonly referred to as high burnup (HBU) fuel. To achieve these higher burnups, it is likely that HBU fuel will also incorporate higher enrichmentspossibly as high as 8% (peak at 10 wt% for BWRs). These higher enrichment fuels are commonly known as high-assay low enriched uranium (HALEU) fuels. Thus, the immediate operational plans of nuclear utilities in the United States include the loading of HALEU/HBU fuel into operating LWRs.

As discussed below, fission product release magnitudes and timings have been established for use in regulatory applications when typical Light Water Reactor (LWR) plant operations achieved lower degrees of burnup. HALEU/HBU fuels could result in modifications to decay heat levels and inventories of radionuclides in the fuel, which can impact the magnitude and timing of fission product release from fuel and ultimately into containment under accident conditions.

1.2. Regulatory Use of Source Terms The following excerpt has been adapted from the NRC Non-Light Water Reactor (Non-LWR)

Vision and Strategy, Volume 3 - Computer Code Development Plans for Severe Accident Progression, Source Term, and Consequence Analysis (U.S. Nuclear Regulatory Commission, 2020).

Regulatory source terms are deeply embedded in the NRCs regulatory policy and practices, as the current licensing process has evolved over the past 50 years. This approach encourages nuclear plant designers to incorporate several lines of defense in order to maintain the effectiveness of physical barriers between radiation hazards and workers, members of the public, and the environment - for both normal operation and accident conditions. The approach centers on the concept of design basis accidents (DBAs), which aim to determine the effectiveness of each line of defense. The DBAs establish and confirm the design basis of the nuclear facility, including its safety-related structures, systems, and components and items important to safety. This ensures that the plant design meets the safety and numerical radiological criteria set forth in regulations and subsequent guidance. From this foundation, specific safety requirements have evolved through a number of criteria, procedures, and evaluations as reflected in the regulations, guides, standard review plans, technical specifications, and license conditions, as well as TID, WASH, and NUREG documents.

The various regulatory source terms for LWRs, used in conjunction with the DBAs, establish and confirm the design basis of the nuclear facility, including items important to safety, ensuring that the plant design meets the safety criteria set forth in the U.S. Code of Federal Regulations (CFR) (e.g., 10 CFR 100.11, Determination of Exclusion Area, Low Population Zone, and Population Center Distance; 10 CFR 50.67, Accident Source Term; 10 CFR 50.34(a)(1)(iv); General Design Criterion 19, Control Room, of Appendix A, General Design Criteria for Nuclear Power Plants, to 10 CFR Part 50, 25

Domestic Licensing of Production and Utilization Facilities) and in subsequent staff guidance.

The NUREG-0800 Standard Review Plan (SRP) for the review of safety analysis reports for LWR nuclear power plants contains specific examples of the various regulatory radiological source terms and provides information on the staffs regulatory guides which were developed for LWRs. The various regulatory source terms discussed in the SRP include the following:

• Accident source term is based on DBAs to establish and confirm the design basis of the nuclear facility and items important to safety while ensuring that the plant design meets the safety and numerical radiological criteria set forth in the CFR (e.g., 10 CFR 50.34(a)(1)(iv), GDC 19, and subsequent staff guidance). SRP Chapter 15 addresses this topic.

• Equipment qualification source term is used to assess dose and dose rates to equipment. SRP Section 3.11, Environmental Qualification of Mechanical and Electrical Equipment; SRP Section 12.2, Radiation Sources; Regulatory Guide 1.89, Environmental Qualification of Certain Electric Equipment Important to Safety for Nuclear Power Plants; and Regulatory Guide 1.183, Alternative Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors, Appendix I, address this topic.

• Post-accident shielding source term is used to assess vital area access, including work in the area. SRP Section 12.2; Item II.B.2 of NUREG-0737, Clarification of TMI Action Plan Requirements, issued November 1980; RG 1.89; and RG 1.183 address this area.

• Design-basis source term is based on 0.25-1 percent fuel defects to determine the adequacy of shielding and ventilation design features. SRP Section 12.2 provides further guidance.

• Anticipated operational occurrences source term is based on the technical specifications or the design-basis source term, whichever is more limiting, to determine the effects of events like primary-to-secondary leakage and reactor steam source term. SRP Section 11.1, Coolant Source Terms, gives reactor coolant (primary and secondary) and reactor steam design details.

• Normal operational source term is based on operational reactor experience, as described in American National Standards Institute/American Nuclear Society N18.1, Selection and Training of Nuclear Power Plant Personnel. SRP Section 11.1 and Section 11.2, Liquid Waste Management System, give further guidance for reactor coolant (primary and secondary) and reactor steam design details, and SRP Section 11.3, Gaseous Waste Management System, gives system design features used to process and treat liquid and gaseous effluents before being released or recycled.

This process of developing accident radiological source terms was initially very prescriptive and defined in TID-18444 Calculations of Distance Factors for Power and Test Reactor Sites. It was replaced by a mechanistic process as defined in NUREG-1465, Accident Source Terms for Light- Water Nuclear Power Plants. Both accident 26

source term characterizations are focused on LWRs. The mechanistic source term described in NUREG -1465 provides the framework for developing accident radiological source terms using methods and codes such as MELCOR for severe accident analysis.

The NRC staff has concluded that an ongoing code development process is appropriate for incorporating new information on accident source terms especially as priorities regarding the different technologies emerge. An applicant may propose changes in source term parameters (timing, release magnitude, and chemical form) from those contained in the applicable guidance, based on and justified by design-specific features. Regulatory Position 2 of Regulatory Guide 1.183 provides attributes of an acceptable alternative accident source term.

To generate an acceptable source term, certain modeling capabilities must be either adapted from current light water capabilities, added for new phenomena specific to new technologies, or ignored for those physics models specific to LWR application. Figure 1-1 below depicts the radionuclide (RN) transport path from release from the fuel to release to the environment for an LWR. Deposition and resuspension of aerosols on surfaces, evaporation and condensation on aerosols and structures, agglomeration of aerosols, chemisorption on surfaces, and bubble transport through coolant are examples of existing phenomena developed for LWR analysis.

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Figure 1-1 Radionuclide transport paths for LWR designs (U.S. Nuclear Regulatory Commission, 2020).

The role of the computer codes used to generate accident source term and consequences is depicted in Figure 1-2. NRCs Office of Nuclear Regulatory Research (RES) is responsible for the development of the computer codes and follows the information flow shown in Figure 1-2. The figure also shows an overview of regulatory uses of the codes by the Office of Nuclear Reactor Regulation (NRR) who is responsible for siting and licensing of new reactor designs. Future uses of the information by the Office of Nuclear Material Safety and Safeguards (NMSS) are also shown (U.S. Nuclear Regulatory Commission, 2020).

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Figure 1-2 Role of Accident Progression, Source Term, and Consequence Analysis Computer Codes and Applicable Regulatory Requirements (U.S. Nuclear Regulatory Commission, 2020) 1.2.1. TID-14844 Source Terms Historically, Regulatory Guide 1.3, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Boiling Water Reactors, Revision 2 (June 1974) and Regulatory Guide 1.4, Assumptions Used for Evaluating the Potential Radiological Consequences of a Loss of Coolant Accident for Pressurized Water Reactors, Revision 2 (June 1974) provided guidance on the safety assessment requirements of 10 CFR 50.34 and by extension the siting requirements of 10 CFR 100.11. The in-containment source term provided within the regulatory guides was derived from TID-14844 (DiNunno, Baker, Anderson, &

Waterfield, 1962). Application of the TID-14844 source term is still permissible today as 10 CRF 100 notes further guidance is provided in TID-14844 and may be used as a point of departure for consideration of particular site requirements which may result from evaluation of the characteristics of a particular reactor, its purpose and method of operation. Presently, most domestic operating reactors have used the TID-14844 in-containment source term for regulatory applications.

TID-14844 is a prescriptive in-containment source term predating quantifiable, safety significant factors of accident initiating events or the pertinent, severe accident progression phenomena.

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Therefore, this analysis developed source terms generally considered to be conservative and non-mechanistic. The study combined experimental observations of radionuclide release from heated and irradiated UO2 fuel pellets. The combined releases are conservatively taken as instantaneous and coincident with the assumed accident initiating event. The adaptation of the TID-14844 study into Regulatory Guides 1.3 and 1.4 is provided below.

Table 1-1 Regulatory Guide 1.3rev2 and 1.4rev2 Compared with TID-14844 Source Term Released Mass Fraction Regulatory Guide 1.3 Chemical Forms in Regulatory Fission Products TID-14844 and 1.4 Guide 1.3 and 1.4 Noble Gases 1.0 1.0 -

0.25 assumed immediately 0.91 - I2-(gaseous) available for leakage from 0.05 - CsI (solid) primary containment. 0.04 - CH3I (organic)

Remaining 0.25 assumed to plate Iodine 0.50 out -

Solid Particulates 0.01 - -

1.2.2. NUREG-1465 Source Terms Following the accident at Three Mile Island Unit 2 (TMI-2), extensive research efforts into severe accident progression, pertinent phenomena, and off-site consequences were performed.

These efforts culminate in the NUREG-1150, Volume 1, Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants, Final Summary Report, (U.S. Nuclear Regulatory Commission, 1990), where severe accident risks were assessed for five nuclear power plants. This seminal work resulted in the estimates of core damage frequencies (CDFs) and accident consequences. Demonstrating quantification of the safety significant factors not available at the time of the TID-14844 (DiNunno, Baker, Anderson, & Waterfield, 1962).

NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) combined the methodologies employed in NUREG-1150 to develop a mechanistic source term that characterized the magnitude and timing of radionuclide release to containment. The study further refined the speciation of iodine into different chemical forms based on state-of-the-art understanding at the time of the study. The revised radionuclide groups considered by NUREG-1465 include 8 groups and constituent elements, as shown in Table 1-2.

30

Table 1-2 NUREG-1465 Revised Radionuclide Groups Group Group Name Elements 1 Noble Gases Xe, Kr 2 Halogens I, Br 3 Alkali Metals Cs, Rb 4 Tellurium Group Te, Sb, Se 5 Barium/Strontium Ba, Sr 6 Noble Metals Ru, Rb, Pd, Mo, Tc, Co 7 Lanthanides La, Zr, Nd, Eu, Nb, Pm, Pr, Sm, Y, Cm, Am 8 Cerium Group Ce, Pu, Np The in-containment source term release was separated into release phases well characterized by the individual analytical codes comprising the accident sequence simulation. These analytical codes were coupled together to simulate accident progression and source term evolution in the Source Term Code Package (STCP).

Replacing the instantaneous release found in the TID-14844 source term, the release of fission products is divided into the following release phases in NUREG-1465:

• Coolant activity release is notionally small, comparatively to the other release phases.

Therefore, NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) reports no appreciable timing or magnitude for this release phase. Coolant activity releases are principally the result of fuel pin failure during normal operations. The fuel pin failures are due to stochastic failures or fuel pin failures due to rapid coolant inventory loss during the accident initiating events.

• Gap activity release occurs as coolant inventory is depleted during progression of the postulated accident. Fuel pins become uncovered, and the cladding fails by high temperature and stress induced rupture. The fission product inventory, released from the fuel matrix but contained within the gap between the fuel and cladding, is released to the RCS and eventually the containment.

• Early in-vessel release is characterized by further elevated temperatures leading to substantial damage of and loss of fuel geometry. Excessive core temperatures due to loss of coolant and rapid oxidation of the cladding enhances diffusion of fission products within fuel grains to grain boundaries. Vaporized species of fission products are released from the fuel.

Dependent on individual vapor pressures, many coalesce into aerosol particulate while being transported through the RCS. Melt formation and debris relocation downward through the core occurs as core material transverses to the reactor vessel lower head.

• Late in-vessel release occurs once the lower head has failed. Corium is released from the vessel and is transported ex-vessel. Core debris retained within the vessel as well as the continued evolution of fission products retained with the vessel may continue to be released from the RCS to the containment.

31

• Ex-vessel release results from the interaction of the core materials and the containment concrete. Chemical reactions between the concrete material and ejected corium are promoted by the excessive debris temperature and decay heat. Concrete ablation by the hot corium, releases gases and forms of aerosols creating direct containment sourcing of fission products.

A peer review of SAND2011-0128 recommended a reorganization of the NUERG-1465 release phases to combine the gap and early in-vessel release phases into a single phase (Energy Research, Inc., 2011). This recommendation was provided to the NRC for their consideration.

NUREG-1465 recommendations for phase durations are provided in Table 1-3 for each release phase. The NUREG-1465 recommendations for the magnitude of release for the selected species groups are shown in Table 1-4 and Table 1-5 for BWRs and PWRs, respectively. It is worth noting that the NUREG-1465 recommendations for BWRs include larger releases during the ex-vessel phase (i.e., after lower head failure) for each radionuclide group except noble gases, which are primarily released during the early in-vessel phase. Recommendations for PWRs are similar except for halogen releases, which are also larger during the early in-vessel phase.

Table 1-3 NUREG-1465 Release Phase Durations Release Phase PWR Durations BWR Durations Coolant Activity 10 s to 30 s 30 s Gap Activity 0.5 h 0.5 h Early In-Vessel 1.3 h 1.5 h Ex-Vessel 2h 3h Late In-Vessel 10 h 10 h Table 1-4 NUREG-1465 recommended radionuclide release fractions for BWRs Fission Product Release Group Gap Release Early In-Vessel Late In-Vessel Ex-Vessel Noble Gases 0.05 0.95 0 0 Halogens 0.05 0.25 0.01 0.35 Alkali Metals 0.05 0.20 0.01 0.35 Te Group 0 0.05 0.005 0.25 Ba/Sr Group 0 0.02 0 0.1 Noble Metals (Ru and Mo) 0 0.0025 0 0.0025 Lanthanides 0 0.0002 0 0.005 Ce Group 0 0.0005 0 0.005 32

Table 1-5 NUREG-1465 recommended radionuclide release fractions for PWRs Fission Product Release Group Gap Release Early In-Vessel Late In-Vessel Ex-Vessel Noble Gases 0.05 0.95 0 0 Halogens 0.05 0.35 0.01 0.25 Alkali Metals 0.05 0.25 0.01 0.35 Te Group 0 0.05 0.005 0.25 Ba/Sr Group 0 0.02 0 0.1 Noble Metals (Ru and Mo) 0 0.0025 0 0.0025 Lanthanides 0 0.0002 0 0.005 Ce Group 0 0.0005 0 0.005 33

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2. OBJECTIVES AND SCOPE This report provides representative source terms developed utilizing decay heats and radionuclide inventories that represent core average loadings with fuel at elevated burnups. The reactor cores specifically considered in this analysis are

• Core-average burnup of 60 GWd/MTU at 5 wt% enrichment

• Core-average burnup of 80 GWd/MTU at 5 wt% enrichment

• Core-average burnup of 60 GWd/MTU at 8 wt% enrichment (peak at 10 wt% for BWRs)

• Core-average burnup of 80 GWd/MTU at 8 wt% enrichment (peak at 10 wt% for BWRs)

The severe accident analyses are conducted using the newest version of the MELCOR integral accident analysis code (Version 2.2). The plant models utilized in this analysis reflect the best practices developed through the State-of-the-Art Reactor Consequence Analyses (SOARCA) study (Bixler, Gauntt, Jones, & Leonard, 2013) (Ross, Bixler, Weber, Sallaberry, & Jones)

(Sandia National Laboratories, 2013) (Sandia National Laboratories) (Sandia National Laboratories). The regulatory source terms determined from these analyses are reported in a manner consistent with NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) release phases and acknowledged risk significant radionuclide speciation groupings.

The primary objectives of this analysis are to:

• Extend the NUREG-1465 alternative source term (AST) to address LWRs with cores designed to utilize HBU fuel with varied fuel enrichments

• Compare extended enrichment and HBU fuel source terms to historical source term estimates

• Perform sensitivity analyses to address issues specific to HBU/HALEU fuel that were identified during the ATF Severe Accident PIRT This analysis is constrained within the following scope:

• Develop representative in-containment source terms in a manner consistent with NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) release phases and risk significant radionuclide speciation groupings.

• Account for a range of LWR and containment types

• Account for diverse set of postulated accident sequences that include principle contributors to core damage frequency

• Use radionuclide inventories for extended enrichment HBU fuels based on ORNL studies to evaluate the effects of extended enrichment HBU fuels on depletion characteristics of BWR and PWR fuel assemblies This report presents the process and evaluation of representative radiological releases to containment (i.e., in-containment source terms) for LWRs utilizing HBU/HALEU fuels. To avoid distracting from the primary analysis objectives, the authors did not report RCS retention data. It should be acknowledged however, that simulations performed in this analysis are consistent with SOARCA evaluations, which did discuss radionuclide retention in the RCS. This 35

analysis focuses on HBU fuel utilizing at enrichments of 5% to as much as 8% and 10 % for PWRs and BWRs, respectively.

The process, which follows the overall approach first established in NUREG-1465, begins with the simulation of a set of representative BWR and PWR severe accidents using the MELCOR severe accident analysis code to quantitatively characterize the magnitudes and timings of key radionuclide classes to containment across a finite set of postulated severe accidents. A statistical methodology is then employed to evaluate, from this distribution of radionuclide release magnitudes and timings, average, representative fission product releases to containment during the four characteristic LWR severe accident release phases first identified in NUREG-1465.

The source terms presented in this report supplement the originally developed ASTs from NUREG-1465, extending the NUREG-1465 AST to address LWRs with cores designed to utilize HBU/HALEU fuel. The sequences utilized in this analysis to develop representative and generic source terms for PWRs and BWRs in operation in the United States are intended to be those that are significant contributors to the overall core damage frequency (CDF) for PWR and BWR plants. The relative contribution of different accident sequences to core damage has previously been determined in the study documented in NUREG-1560.

While the core design in this report is assumed to be different from the lower burnup cores originally considered in the NUREG-1560 study, it is reasonable to assume that a higher burnup core will not change the relative contribution of different accident sequences to the total CDF of a PWR or BWR plant.

Each accident prescribed accident sequence is simulated four times, once with each fission product inventory considered. Fission product inventories correspond to core averaged burnups of 60 and 80 GWd/MTU with fuel enrichments of 5% and 8% (peak of 10% for BWR fuel) to demonstrate burnup in excess of the NUREG-1465 40 GWd/MTU and extended fuel enrichment.

The current analysis is performed using core-averaged representative HBU fuel assemblies. This assumption can be refined to use radionuclide inventories obtained from whole core analyses for 60 and 80 GWd/MTU peak assembly burnup with extended enrichments.

Each plant model has been taken from the most recent NRC study to incorporate that model. The models have been modified to perform selected accident sequences that reflect principal contributors to core damage frequency estimates. The Peach Bottom, Surry, and Sequoyah models were used in the prior SOARCA studies (Bixler, Gauntt, Jones, & Leonard, 2013)

(Sandia National Laboratories, 2013) (Ross, Bixler, Weber, Sallaberry, & Jones) (Sandia National Laboratories) (Sandia National Laboratories), while the Grand Gulf model was used in the prior source term analyses (Powers, Leonard, Gauntt, Lee, & Salay, 2011) (Leonard, Gauntt,

& Powers, 2007). The results of the analyses provide characterizations of the in-containment source terms presenting the timing and magnitude of fission product groups entering containment with respect to the NUREG-1465 phase releases.

2.1. Alternative Source Term The in-containment source terms characterized within this analysis are patterned after NUREG-1465. This is purposeful as the AST acceptance criteria provided in Regulatory Guide 1.183, Alternative Radiological Source Terms For Evaluating Design Basis Accidents At Nuclear Power Reactors (U.S. Nuclear Regulatory Commission, 2000), promote commensurate rigor in 36

developing an in-containment source term as NUREG-1465. The following criteria are necessary as per Regulatory Guide 1.183:

The AST must be based on major accidents, hypothesized for the purposes of design analyses or consideration of possible accidental events, that could result in hazards not exceeded by those from other accidents considered credible. The AST must address events that involve a substantial meltdown of the core with the subsequent release of appreciable quantities of fission products.

The AST must be expressed in terms of times and rates of appearance of radioactive fission products released into containment, the types and quantities of the radioactive species released, and the chemical forms of iodine released.

The AST must not be based upon a single accident scenario but instead must represent a spectrum of credible severe accident events. Risk insights may be used, not to select a single risk-significant accident, but rather to establish the range of events to be considered. Relevant insights from applicable severe accident research on the phenomenology of fission product release and transport behavior may be considered.

The AST must have a defensible technical basis supported by sufficient experimental and empirical data, be verified and validated, and be documented in a scrutable form that facilitates public review and discourse.

The AST must be peer-reviewed by appropriately qualified subject matter experts.

The peer-review comments and their resolution should be part of the documentation supporting the AST.

2.2. Selected Reactors for Analysis The overall purpose of the AST methodology introduced in NUREG-1465 is the development of a generic source term that is representative of a broad range of power plants operating across the United States. As this analysis is an update to the in-containment source terms reported in SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011), the same set of reactors are used in this analysis. Plant models used in this analysis include:

• PWR: Large-dry containment (Surry) and Ice Condenser containment (Sequoyah)

• BWR: Mark I containment (Peach Bottom) and Mark III containment (Grand Gulf) 2.2.1. PWR with Large-Dry Containment - Surry The Surry Power Station has two PWR units. It is located at Gravel Neck and is on the James River in Surry County, VA. Both units at Surry are Westinghouse three-loop reactors enclosed in a large-dry containment that is maintained sub-atmospheric during operation. Each unit is assumed to operate at 2546 MWth.

Both Surry units are equipped with pilot-operated relief valves (PORVs), high-pressure safety injection (HPSI), low-pressure safety injection (LPSI), and residual heat removal (RHR) systems to address heat removal and pressure regulation during accident transients. The large-dry (subatmospheric) containments consist of a large reinforced concrete cylinder and steel liner that enclose the reactor vessel, steam generator, and other primary side components. Under accident 37

conditions, containment integrity is maintained by the large free volume and high design pressures. Fission product scrubbing and heat removal is achieved by containment spray (CS) systems.

The Surry MELCOR model for these studies is based on that developed to support the Surry SOARCA study in References (Sandia National Laboratories, 2013), and (Ross, Bixler, Weber, Sallaberry, & Jones). Additional information on important features of the MELCOR model are provided in Section 3.2.

2.2.2. PWR with Ice Condenser Containment - Sequoyah The Sequoyah Nuclear Plant has two PWR units located in east Tennessee, 18 miles north of Chattanooga on the Chickamauga Reservoir. Both units at Sequoyah are Westinghouse four-loop reactors enclosed within an ice condenser containment. Each unit is assumed to operate at a nominal power level of 3455 MWth in this analysis.

Westinghouse four-loop reactors have a number of systems designed to inject coolant or otherwise manage accident conditions including: HPSI, LPSI, RHR, and PORVs. The ice condenser containment operates at low design pressure (approximately 12 psig), small free volume (relative to other PWR containments, and uses ice beds to suppress containment pressure during accident conditions. Pressure suppression is achieved in ice condenser containments through heat transfer to the ice beds; ice beds also offer some fission product scrubbing capability. An air return system is available for generating forced circulation of the containment atmosphere through the ice chests. Further fission product scrubbing and heat removal is accomplished through CS systems.

The Sequoyah MELCOR model utilized for these studies is based on that developed to support the Sequoyah SOARCA study (Sandia National Laboratories). Additional information on important features of the MELCOR model are provided in Section 3.2.

2.2.3. BWR Mark I - Peach Bottom The Peach Bottom Atomic Power Station has two, unit 2 and unit 3, operating General Electric designed boiling water reactors (BWRs), type 4, with Mark I containment systems. Unit 1 is a decommissioned high temperature gas reactor. Unit-2 is analyzed within this analysis and has been a reference power plant for numerous NRC studies historically, including the NUREG-1150 study. The accident model of the facility is derived from the Peach Bottom SOARCA study in References (Bixler, Gauntt, Jones, & Leonard, 2013) and (Sandia National Laboratories).

Unit-2 has an operating thermal capacity of 3951 MWth. As a BWR-4, the safety systems include the high-pressure coolant injection system, reactor core isolation cooling system (RCIC),

low-pressure coolant injection (LPCI) with RHR system operation, and an automatic depressurization system (ADS). The Mark I primary containment is enclosed within a reactor building, referred to as a secondary containment. The primary containment is a steel lined concrete reinforced structure, commonly referred to as the drywell, which employs a torus shaped pressure suppression system referred to as the suppression chamber or wetwell. The suppression chamber is connected to the drywell through a vent system. Steam released to the containment due to LOCAs or steam released from the vessel due to SRV or emergency core cooling system (ECCS) operations to the containment, to the drywell atmosphere or the SRV tail 38

pipe through a pipe-rupture or to the wet well due to safety relief valve operations, are managed passively by steam condensation within the suppression pool.

The Peach Bottom MELCOR model utilized for these studies is based on that developed to support the Peach Bottom SOARCA study (Bixler, Gauntt, Jones, & Leonard, 2013; Sandia National Laboratories). Additional information on important features of the MELCOR model are provided in Section 3.2.

2.2.4. BWR Mark III - Grand Gulf The Grand Gulf Nuclear Station has a single General Electric BWR-6 housed in Mark III containment. The Grand Gulf reactor has undergone a number of power uprates and is currently rated to operate at 4408 MWth.

Safety systems in a BWR-6 include high-pressure coolant injection system, reactor core isolation system (RCIC), LPCI with RHR system operation, and an ADS. The Mark III primary containment consists of both a drywell and suppression pool contained in a cylindrical steel containment vessel. The inner boundary of the suppression pool, the weir wall, is inside of the drywell and exposed to the drywell atmosphere. The weir wall forms an annulus with the external wall of the drywell, which is connected to the rest of the suppression pool by 3 rows of horizontal vent pipes. The suppression pool functions to decrease drywell pressure under accident conditions through the transport of drywell atmosphere into the suppression pool. As pressure increases in the drywell, the water level in the annulus drops. The decrease in water level progressively uncovers the vents and the drywell atmosphere is vented into the suppression pool. The suppression pool can also function as a heat sink during SRV operation or as a heat sink and water source for ECCS.

The Grand Gulf MELCOR model used for these studies is based on the model developed to support the previous severe accident study on HBU fuels (SAND2011-0128) (Leonard, Gauntt,

& Powers, 2007; Powers, Leonard, Gauntt, Lee, & Salay, 2011). Additional information on important features of the MELCOR model are provided in Section 3.2.

2.3. Alternative Source Term Reactor Cores Analyzed 2.3.1. BWR Core Specification The different reactor cores assessed in this analysis are based on work recently performed by Oak Ridge National Laboratory (ORNL) assessing the effects of extended enrichment and HBU fuels on lattice physics parameters and used fuel fission product isotopic compositions for BWRs (Cumberland, Sweet, Mertyurek, Hall, & Wieselquist, 2021). The ORNL Phase 1 study assessed a conventional GNF-2 10x10 BWR fuel lattice developed with GE14 lattice parameters and GNF-2 vanished rod positions. The analysis was conducted utilizing the SCALE Polaris lattice physics code and the ORIGEN depletion and decay code. The study analyzed the following 235U enrichment levels and lattice-averaged burnups

• Enrichment of 5 wt%, reflecting the current enrichment limit for operating LWRs, for burnups as high as 80 GWd/MTU

• Enrichment of 8.5 wt% for burnups as high as 80 GWd/MTU 39

• Enrichment of 10 wt% to capture the maximum extended enrichment increase considered in the near-term, with burnups as high as 80 GWd/MTU The specific results of the ORNL study of relevance to this source term study are

• Decay heat with a focus on the short-term of relevance to reactor accidents initiated from full power The decay heat is applicable to a single fuel assembly. It is expressed as a time-dependent fraction of the assembly power at the time of shutdown. The total power at shutdown for all fuel assemblies in the core is specified by the nominal operating power for the power being analyzed. However, the local power is assumed to follow the same local power established for current reactor cores, Low Burnup (LBU), for the plants being analyzed. The assembly decay heat fraction is thus scaled using radial and axial power profiles used in current MELCOR reactor core models of the plants analyzed in this analysis. These radial and axial power profiles were established from prior ORIGEN analyses.

• Isotopic inventory The isotopic inventory determined for a single fuel assembly is applied uniformly across the reactor cores of the plants analyzed. It is expected that there will be a spatial variation of the isotopic inventory reflecting the distribution of assembly burnup across a reactor core (under realistic fuel loadings). This is common MELCOR modeling practice. The impact of assuming a uniform distribution of isotopic inventories is expected to be small, since the most important condition influencing accident progression and fission product release is the decay heat. The spatial variation of decay heat, as noted above, is reflected in this analysis.

The ORNL study (Cumberland, Sweet, Mertyurek, Hall, & Wieselquist, 2021) assumed a GNF-2 10x10 BWR fuel assembly design as representative for the purposes of assessing the impact of extended enrichment and increased burnup on lattice physics parameters and isotopic inventory.

The assumed fuel design in the ORNL study (Cumberland, Sweet, Mertyurek, Hall, &

Wieselquist, 2021) is summarized in Table 2-1.

40

Table 2-1 Fuel Design Parameters from ORNL BWR Study (Cumberland, Sweet, Mertyurek, Hall, & Wieselquist, 2021)

Parameter Value Assembly lattice 10x10 Assembly pitch 15.24 cm Fuel rod pitch 1.295 cm Clad material Zirc-2 UO2 pellet radius 0.4380 cm Clad inner radius 0.4470 cm Clad outer radius 0.5130 cm Water tube inner radius 1.20 cm Water tube outer radius 1.28 cm Channel width (inside) 13.406 cm Channel box thickness 0.2032 cm Channel radius 0.9652 cm Fuel temperature 1100 K Coolant temperature 580 K Clad temperature 600 K UO2 effective density 10.64 g/cm3 Coolant density 0.7048 g/cm3 The variation of fractional decay heat as a function of cooling time obtained from the ORNL study (Cumberland, Sweet, Mertyurek, Hall, & Wieselquist, 2021) is presented in Figure 2-1.

This illustrates the relatively minor effect on decay heat as a function of cooling time with the introduction of extended enrichment and increased burnup fuels.

41

Figure 2-1 BWR Fractional Decay Heat Variation with Cooling Time (Cumberland, Sweet, Mertyurek, Hall, & Wieselquist, 2021)

The effect of extended enrichment and increased burnup on decay heat can be seen more clearly in the relative difference in decay heat with respect to the reference enrichment and burnup for current cores. Figure 2-2 shows the difference in decay heat relative to the reference fuel. For the purposes of this comparison, the reference fuel is assumed to have an enrichment of 5 max-4.5 avg wt%, a burnup of 60 GWd/MTU at 40% void. During the near-term cooling times of relevance to an accident initiated from full power operation, differences in decay heat relative to the reference fuel do not exceed 10%. The most significant differences for cooling times of 3 days are around 5% to 7%. It is only in the long-term, for cooling times of most relevance to spent fuel pool accidents, that Figure 2-2 identifies significant differences in decay heat relative to the reference fuel.

42

Figure 2-2 BWR Decay Heat Relative to Reference Fuel (Cumberland, Sweet, Mertyurek, Hall, &

Wieselquist, 2021)

The relatively minor change in decay heat observed in Figure 2-2 indicates that there may be relatively minor differences in accident progression and in-containment source terms relative to a reference scenario consistent with NUREG-1465, with burnups not exceeding 60 GWd/MTU.

2.3.2. PWR Core Specification The different reactors cores assessed in this analysis are based on work recently performed by ORNL to assess the effects of extended enrichment and HBU fuels on lattice physics parameters and used fuel isotopic compositions (Hall, Cumberland, Sweet, & Wieselquist, 2021). The ORNL Phase 1 study assessed a conventional Westinghouse 17x17 PWR fuel design utilizing the SCALE Polaris lattice physics code and the ORIGEN depletion and decay code. The study analyzed the following 235U enrichment levels and lattice-averaged burnups

• Enrichment of 5 wt% for burnups as high as 80 GWd/MTU

• Enrichment of 6.5 wt% for burnups as high as 80 GWd/MTU

• Enrichment of 8 wt% for burnups as high as 80 GWd/MTU The specific results of the study of relevance to this source term study are

• Decay heat with a focus on the short-term of relevance to reactor accidents initiated from full power 43

The decay heat is applicable to a single fuel assembly. It is expressed as a time-dependent fraction of the assembly power at the time of shutdown. The total power at shutdown for all fuel assemblies in the core is specified by the nominal operating power for the power being analyzed. However, the local power is assumed to follow the same local power profile established for current reactor cores, Low Burnup (LBU), for the plants being analyzed. The assembly decay heat fraction is thus scaled using radial and axial power profiles used in current MELCOR reactor core models of the plants analyzed in this analysis. These radial and axial power profiles were established from prior ORIGEN analyses.

• Isotopic inventory The isotopic inventory determined for a single fuel assembly is applied uniformly across the reactor cores of the plants analyzed. It is expected that there will be a spatial variation of the isotopic inventory reflecting the distribution of assembly burnup across a reactor core (under realistic fuel loadings). This is common MELCOR modeling practice. The impact of assuming a uniform distribution of isotopic inventories is expected to be small, since the most important condition influencing accident progression and fission product release is the decay heat. The spatial variation of decay heat, as noted above, is reflected in this analysis.

A more detailed evaluation that incorporates isotopic distribution throughout the reactor core corresponding more realistic fuel loading patterns (i.e., accounting for a distribution in burnup) will be possible in Phase 2 of this analysis. Phase 2 will incorporate core level assessments utilizing the core simulator PARCS.

The ORNL study (Hall, Cumberland, Sweet, & Wieselquist, 2021) assumed a Westinghouse 17x17 PWR fuel assembly as representative for the purposes of assessing the impact of extended enrichment and increased burnup on lattice physics parameters and isotopic inventory. The assumed fuel design in the ORNL study (Hall, Cumberland, Sweet, & Wieselquist, 2021) is summarized in Table 2-2.

The variation of fractional decay heat as a function of cooling time obtained from the ORNL study (Hall, Cumberland, Sweet, & Wieselquist, 2021) is presented in Figure 2-3. This illustrates the relatively minor effect on decay heat as a function of cooling time with the introduction of extended enrichment and increased burnup fuels.

The effect of extended enrichment and increased burnup on decay heat can be seen more clearly in the relative difference in decay heat with respect to the reference enrichment and burnup for current cores. Figure 2-4 shows the difference in decay heat relative to the reference fuel. For the purposes of this comparison, the reference fuel is assumed to have an enrichment of 5 wt% at a burnup of 60 GWd/MTU. During the near-term cooling times of relevance to an accident initiated from full power operation, differences in decay heat relative to the reference fuel do not exceed 10%. The most significant differences for cooling times of 3 days are around 5% to 7%.

It is only in the long-term, for cooling times of most relevance to spent fuel pool accidents, that Figure 2-4 identifies significant differences in decay heat relative to the reference fuel.

The relatively minor change in decay heat observed in Figure 2-4 indicates that there may be relatively minor differences in accident progression and in-containment source terms relative to a reference scenario consistent with NUREG-1465, with burnups not exceeding 60 GWd/MTU.

44

Figure 2-3 PWR Decay Heat Relative to Reference Fuel (Hall, Cumberland, Sweet, &

Wieselquist, 2021) 45

Figure 2-4 PWR Fractional Decay Heat Variation with Cooling Time (Hall, Cumberland, Sweet, &

Wieselquist, 2021) 46

Table 2-2 Fuel Design Parameters from ORNL PWR Study (Hall, Cumberland, Sweet, &

Wieselquist, 2021)

Parameter Value Assembly lattice 17x17 Assembly pitch 21.5 cm Fuel rods 264 IFBA rods 104 Guide tubes 24 Instrument tubes 1 Fuel rod pitch 1.26 cm Clad material Zirc-4 UO2 pellet radius 0.4096 cm UO2 model depletion rings 3 equal volumes UO2 effective density 10.26 g/cm3 IFBA radius 0.4106 cm IFBA B-10 loading 0.927 mg/cm/rod Clad inner radius 0.418 cm Clad outer radius 0.475 cm Guide tube inner radius 0.561 cm Guide tube outer radius 0.602 cm Instrument tube inner radius 0.559 cm Instrument tube outer radius 0.605 cm Fuel temperature 900 K Coolant temperature 583.15 K Coolant density 0.7048 g/cm3 Clad temperature 700 K Depletion power 40 MW/MTU 2.3.3. Isotopic Inventories In this source term study, core-averaged isotopic inventories are generated for each reactor considered. Core-averaged isotopic inventories are generated by scaling the isotopic inventories calculated by ORNL for the appropriate reference assembly described in Sections 2.3.1 and 2.3.2. Table 2-3 presents the BWR isotopic inventories for the different radionuclide classes represented by MELCOR in this analysis. Table 2-4 presents the PWR isotopic inventories for the different radionuclide classes. The relative inventory difference compared to the 60 gwd/mtu

- 5 wt% Enrichment case is presented in parentheses for each reactor.

47

• Core-average burnup of 60 GWd/MTU at 5 wt% enrichment

• Core-average burnup of 80 GWd/MTU at 5 wt% enrichment

• Core-average burnup of 60 GWd/MTU at 8 wt% enrichment (peak at 10 wt% for BWRs)

• Core-average burnup of 80 GWd/MTU at 8 wt% enrichment (peak at 10 wt% for BWRs)

Table 2-3 BWR Isotopic Inventory across MELCOR Radionuclide Classes 60 GWd/MTU - 80 GWd/MTU - 60 GWd/MTU - 80 GWd/MTU -

5 wt% 5 wt% 10 wt% 10 wt%

Class Enrichment Enrichment Enrichment Enrichment BWR Mark I - Peach Bottom Noble Gases (kg) 1323.99 1848.13 (+40%) 1280.34 (-3%) 1790.12 (+35%)

Halogens (kg) 52.83 73.70 (+40%) 49.41 (-6%) 69.53 (+32%)

Alkali Metals (kg) 748.78 980.11 (+31%) 817.97 (+9%) 1082.33 (+45%)

Te Group (kg) 142.94 195.01 (+36%) 139.99 (-2%) 190.51 (+33%)

Ba/Sr Group (kg) 551.99 763.09 (+38%) 586.41 (+6%) 814.05 (+47%)

Ru Group (kg) 1058.01 1598.56 (+51%) 919.02 (-13%) 1374.61 (+30%)

Mo Group (kg) 973.05 1305.64 (+34%) 1007.92 (+4%) 1364.59 (+40%)

Lanthanides (kg) 2943.70 3702.34 (+26%) 2922.84 (-1%) 3686.46 (+25%)

Ce Group (kg) 2469.33 2916.84 (18%) 2559.90 (+4%) 3107.02 (+26%)

BWR Mark III - Grand Gulf Noble Gases (kg) 1387.47 1936.75 (+40%) 1341.74 (-3%) 1875.96 (+35%)

Halogens (kg) 55.36 77.23 (+40%) 51.77 (-6%) 72.87 (+32%)

Alkali Metals (kg) 784.69 1027.10 (+31%) 857.19 (+9%) 1134.23 (+45%)

Te Group (kg) 149.79 204.36 (+36%) 146.70 (-2%) 199.65 (+33%)

Ba/Sr Group (kg) 578.46 799.68 (+38%) 614.53 (+6%) 853.08 (+47%)

Ru Group (kg) 1108.74 1675.22 (+51%) 963.09 (-13%) 1440.52 (+30%)

Mo Group (kg) 1019.71 1368.25 (+34%) 1056.25 (+4%) 1430.03 (+40%)

Lanthanides (kg) 3084.85 3879.87 (+26%) 3063.00 (-1%) 3863.23 (+25%)

Ce Group (kg) 2587.74 3056.71 (+18%) 2682.65 (+4%) 3256.01 (+26%)

• Representative in-containment source terms are reported as a fraction of the radionuclide mass inventory shown in the table above. It should be noted that changes in the mass inventory for a given radionuclide class is not directly correlated to an increase in activity or by extension dose.

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Table 2-4 PWR Isotopic Inventory across MELCOR Radionuclide Classes 60 GWd/MTU - 80 GWd/MTU - 60 GWd/MTU - 80 GWd/MTU -

5 wt% 5 wt% 8 wt% 8 wt%

Class Enrichment Enrichment Enrichment Enrichment PWR with Large-Dry Containment - Surry Noble Gases (kg) 740.20 987.15 (+33%) 717.66 (-3%) 959.00 (+30%)

Halogens (kg) 29.31 39.35 (+34%) 27.44 (-6%) 37.06 (+26%)

Alkali Metals (kg) 421.27 537.41 (+28%) 455.26 (+8%) 584.21 (+39%)

Te Group (kg) 74.62 99.01 (+33%) 73.02 (-2%) 96.81 (+30%)

Ba/Sr Group (kg) 305.28 401.76 (+32%) 323.92 (+6%) 428.01 (+40%)

Ru Group (kg) 559.35 807.23 (+44%) 487.92 (-13%) 701.18 (+25%)

Mo Group (kg) 530.59 689.06 (+30%) 546.71 (+3%) 714.95 (+35%)

Lanthanides (kg) 1035.01 1396.16 (+35%) 1048.46 (+1%) 1409.24 (+36%)

Ce Group (kg) 1535.14 1780.67 (+16%) 1599.41 (+4%) 1903.19 (+24%)

PWR with Ice Condenser Containment - Sequoyah Noble Gases (kg) 918.04 1224.33 (+33%) 890.08 (-3%) 1189.41 (+30%)

Halogens (kg) 36.35 48.80 (+34%) 34.03 (-6%) 45.97 (+26%)

Alkali Metals (kg) 522.48 666.52 (+28%) 564.65 (+8%) 724.57 (+39%)

Te Group (kg) 92.54 122.79 (+33%) 90.57 (-2%) 120.07 (+30%)

Ba/Sr Group (kg) 378.63 498.29 (+32%) 401.74 (+6%) 530.85 (+40%)

Ru Group (kg) 693.74 1001.18 (+44%) 605.14 (-13%) 869.65 (+25%)

Mo Group (kg) 658.07 854.61 (+30%) 678.07 (+3%) 886.72 (+35%)

Lanthanides (kg) 1283.69 1731.61 (+35%) 1300.36 (+1%) 1747.83 (+36%)

Ce Group (kg) 1903.98 2208.50 (+16%) 1983.68 (+4%) 2360.46 (+24%)

• Representative in-containment source terms are reported as a fraction of the radionuclide mass inventory shown in the table above. It should be noted that changes in the mass inventory for a given radionuclide class is not directly correlated to an increase in activity or by extension dose.

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3. METHODOLOGY AND ASSUMPTIONS 3.1. Overall Study Methodology The development of representative source terms for BWRs and PWRs has been summarized in SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011) and SAND2016-12954 (Gauntt, Goldman, Kalanich, & Powers, 2016). The methodology was originally developed to determine representative BWR and PWR source terms, respectively, to be used as an Alternate Source Term (AST) in NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995).

The AST was intended to replace the originally bounding source terms derived from TID-14844 (DiNunno, Baker, Anderson, & Waterfield, 1962). It is important to note that the source terms derived from TID-14844 (DiNunno, Baker, Anderson, & Waterfield, 1962) are specific to LWR fuel at much lower burnups than being considered in this analysis. This section provides a summary of the process involved in developing these representative source terms.

As noted above, these representative source terms are intended to support several regulatory activities that require a quantitative estimate of the release of radioactive material into a BWR or PWR containment in the event of a severe accident. The overall approach is designed to provide a single estimate for the in-containment source term at different stages of an accident. Given that there are a range of uncertainties that induce variation in the source term estimate, a statistical approach was developed to establish the original NUREG-1465 source terms (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995).

The statistical approach developed in NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) accounted for the most significant uncertainties that influence fission product release to containment magnitudes and durations during different phases of an accident. At the time of NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) development, the dominant uncertainties were recognized to be the range of different accident scenarios that could be realized, which is often termed aleatory (random) uncertainty. While phenomenological uncertainty, also known as epistemic uncertainty, can be quite important when considering variation in accident progression and source terms for a specific accident, multiple studies have highlighted the more significant variation that occurs in accident progression and source term estimates across different accident scenarios. Thus, aleatory uncertainty remains the primary uncertainty assessed in this analysis to identify a reasonable variability in the in-containment source terms that could occur for BWR and PWR plants. This aleatory uncertainty is probed in this analysis through MELCOR simulation of accident progression and in-containment source term for a range of accident scenarios that lead to significant core damage and in-containment source terms. The selection of these accident scenarios is discussed in more detail in Section 4.

With the variations in the in-containment source term estimates, a statistical bootstrap methodology described in Section 3.4 is utilized to determine a single median in-containment source term magnitude and duration for different accident phases. This approach is used to establish median estimates representative of BWR and PWR plants that can be applied generically across the fleet of operating BWR and PWR plants in the United States.

The overall steps listed below to generate these representative in-containment source terms are depicted in Figure 3-1.

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• Utilize available BWR and PWR plant PRAs to identify BWR and PWR accident scenarios that lead to significant core damage and in-containment source terms

• Develop radionuclide inventories and core decay heats for assumed core loading patterns (i.e., spatial fuel burnup and enrichment distributions) using the SCALE code package10

• Perform accident progression and source term analyses using a best-estimate LWR severe accident analysis code (i.e., MELCOR) for each of the considered accident scenarios

• Develop an empirical distribution of containment fission product release magnitudes and durations at different phases of an accident using a non-parametric, statistical bootstrap method (described in greater detail in Section 3.4)

RN Inventory Base Plant Specification Model MELCOR Input Sequence Specification MELCOR Evaluation of Statistical output QoIs Bootstrap Analysis Figure 3-1 Stages of in-containment source term Analysis 3.2. General MELCOR Modeling Approach 3.2.1. Reactor Coolant System or Nuclear Steam Supply System Failure Modes Failures in the PWR Reactor Coolant System or BWR Nuclear Steam Supply System can be induced during the progression of a non-LOCA severe accident. Since such failures can introduce flow paths for fission products into containment, it is critical to assess the potential during the in-vessel phase for occurrence of these failures.

There has been significant evolution in the understanding of these failure modes through the course of the Peach Bottom, Surry and Sequoyah SOARCA studies (Bixler, Gauntt, Jones, &

Leonard, 2013), (Sandia National Laboratories, 2013), (Ross, Bixler, Weber, Sallaberry, &

Jones), and (Sandia National Laboratories). This enhancement in the severe accident progression 10 Note that these quantities are input to MELCOR simulations. As noted above, the development of HALEU/HBU radionuclide inventories and decay heats is performed in separate studies that provide input to the MELCOR accident progression and source term analyses documented in this report.

52

state-of-knowledge is incorporated into this source term study. It represents an area of significant evolution relative to representation of severe accident progression in the NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) study as well as the more recent BWR and PWR source term studies described in SAND2011-0128 (Powers, Leonard, Gauntt, Lee, &

Salay, 2011) and SAND2016-12954 (Gauntt, Goldman, Kalanich, & Powers, 2016). The subsequent discussion summarizes the aspects of the MELCOR plant models specific to resolving PWR Reactor Coolant System or BWR Nuclear Steam Supply System integrity, prior to RPV lower head breach, under severe accident conditions.

3.2.1.1. Safety Relief Valve Failure of the BWR Nuclear Steam Supply System in Severe Accidents The Peach Bottom SOARCA study (Bixler, Gauntt, Jones, & Leonard, 2013) developed a technical basis for representing the potential for Safety Relief Valve (SRV) seizure under severe accident conditions. Through the SOARCA study, the understanding of thermal conditions under which a cycling SRV could seize open or closed, and hence cease to cycle, experienced significant evolution.

Initially, the criterion for high-temperature valve failure was based on manufacturers information describing the strength of stainless steel, published by the Stainless-Steel Information Center. Softening or loss of strength of stainless steel (300 series) was described as about 1000 oF (811 K). The same reference also suggested the maximum service temperature for intermittent exposure of stainless-steel components is 1600 oF (1100 K). In analytical models for simulation accident progression, this was often assumed to be the temperature of the thermal environment in which steel components were operating. For a cycling SRV, this corresponds to the internal gas temperature to which the SRV is exposed. In initial analyses exploring the effect of severe thermal transient under severe accident conditions on performance of SRVs, failure to operate was assumed to occur when the internal gas temperature exceeded a temperature between 811 K and 1100 K. Specifically, valve seizure was assumed to occur when gas temperatures to which the SRV was exposed exceed about 1000 K over several valve cycles.

However, further study identified a strong sensitivity to temperature across different SRV designs. For example, different thermal failure criteria were developed for the Peach Bottom and Grand Gulf MELCOR models because the two plants have different types of SRVs. Three-stage Target Rock SRVs are installed at Peach Bottom, and Dikkers SRVs are installed at Grand Gulf.

Target Rock SRVs are pilot-operated valves, which lift to a full-open position when pressure within the SRV exceeds a setpoint. When pressure decreases below another setpoint, the valve fully re-seats. Movement of the main valve disc is controlled by a pilot valve, which is distinct from, but integral to, the main SRV valve body. Movement of the pilot valve re-aligns gas flow through small ports and vent lines within the valve body, allowing RPV pressure to maintain the valve fully seated when pressure is within the desired range, and to promote lifting of the valve if pressure becomes too high. In contrast, the Dikkers/Crosby SRV design is a spring- loaded valve that pops open to relieve RPV pressure. It then gradually recloses as internal pressure decreases. The variable valve stem position (or valve open fraction) allows RPV pressure to be maintained close to a target value until RPV pressure reduces below a minimum setpoint when the valve recloses.

The number of times a Target Rock valve could cycle at temperatures at or above 1000 K is not well-known. A value of 10 cycles was chosen to represent the expectation that several cycles 53

would be necessary to transfer enough heat to valve internal components to deform or expand valve components, causing failure. It was recognized that although convective heat transfer from the gas would only occur when the valve was open, heat transfer within the valve body would continue after the valve stem reseated. Therefore, it was postulated that non-uniform thermal expansion could reduce clearances of valve components and cause the valve to seize in the closed position. Alternatively, material softening and deformation could cause the valve to fail in the open position. Uncertainties in both failure mechanisms (i.e., thermal expansion and material softening/deformation) were also reflected in the number of cycles used to characterize failure to reclose. That is, if the SRV having the lowest set point closed after 2 or 3 cycles with gas temperatures above 1000 K, the next lowest set point SRV would pick up the load and begin cycling without significant pre-heating. A nominal valve failure (seize open) criterion was defined as: 10 cycles with gas temperatures, prior to opening, above 1000 K. This was judged to be a reasonable approximation of conditions under which one of the 11 Target Rock valves at Peach Bottom would seize in the open position.

Tracking the number of valve cycles was not judged to be appropriate for characterizing heat up and seizure of a Dikkers/Crosby valve. Instead, seizure is assumed to occur if the valve discharges high temperature gas for a sufficiently long period of time. A failure criterion was developed based on the concept of a cumulative damage function, which tracked the amount of time high temperature gas was discharged through the valve and compared it to a time limit. The time limit was assumed to be inversely proportional to temperature, and the valve was assumed to seize in the position it held at the time the failure limit was reached. This was often a few percent of full open. The specific values used in this time-at-temperature criterion were:

• 60 minutes at temperatures of 1000 K

• 30 minutes above 1500 K 3.2.1.2. Main Steam Line Creep Rupture of the BWR Nuclear Steam Supply System in Severe Accidents Significant investigation of the potential for BWR main steam line rupture was performed for the Peach Bottom SOARCA (Bixler, Gauntt, Jones, & Leonard, 2013). This effort indicated that a potential exists for main steam line creep rupture due to the hot gases exiting the RPV under severe accident conditions.

3.2.1.3. Safety Valve Failure of the PWR Reactor Coolant System in Severe Accidents Special models have been included to simulate the failure of the PWR primary system relief valves. Each valve was individually modeled to accurately characterize its operational characteristics. The potential for failure under normal operating conditions and failure at high temperature severe accident conditions was considered.

The MELCOR model represents each of the three SRVs on the pressurizer separately. The valves are individually sized to allow steam flow of 293,330 lb/hr at 2,485 psig. The opening pressures are staggered by 14.50 psi. The lowest opening pressure is set to 2,485 psig. The valves close when pressure drops below 96% of their opening pressure. The SRV with the lowest opening pressure is configured to fail open using the following criteria:

• Stochastic failure after 256 cycles, or 54

• 10 cycles above 1000 K The MELCOR model represents the two PORVs on the pressurizer separately. The valves are individually sized to allow steam flow of 210,000 lb/hr at 2,335 psig. The opening pressures are staggered by 14.50 psi. The lowest opening pressure is set to 2,335 psig. The valves close when pressure drops below 96% of their opening pressure. The PORV with the lowest opening pressure is configured to fail open using the following criteria:

• Stochastic failure after 247 cycles, or

• 10 cycles above 1000 K The Surry SOARCA uncertainty study (Ross, Bixler, Weber, Sallaberry, & Jones) determined that in most realizations a safety valve on the Reactor Coolant System primary side (on the pressurizer) failed to close. Out of the nearly 1003 realizations studied in the Surry SOARCA uncertainty study (Ross, Bixler, Weber, Sallaberry, & Jones), 686 realizations exhibited a failure to close of a primary system safety valvei.e., 68% of realizations.

3.2.1.4. Induced Failures of the PWR Reactor Coolant System in Severe Accidents Core overheating commences in a PWR once significant loss of water inventory has occurred.

This leads to the flow of very hot steam and potentially hydrogen (from oxidation of exposed Zircaloy fuel cladding) out of the core and into the upper regions of the RPV. These hot gases are ultimately transported into the hot leg and steam generators. At high system pressures in non-LOCA scenarios, natural circulation, and heat transfer from these hot gases results in heatup of structures above the reactor core as well as system components and piping interfacing with the upper plenum of the RPV. At these high system pressures, severe heatup of system structures and piping interfacing with the RPV upper plenum has the potential to result in material creep.

The heatup of system structures and piping is sensitive to the natural circulation flows that develop in a PWR at this stage of an accident. The expected natural circulation flows in the PWR primary system are illustrated in Figure 3-2.

Should sufficient creep occur, an induced rupture of vulnerable system piping could occur. This has the effect of causing an induced LOCA following core damage, opening a pathway for flow not only of steam and hydrogen from the primary system but also fission products generated under the severe conditions experienced by the reactor core. Depending on the failure location, this pathway for release from the primary system may be either into containment or the secondary side of the steam generators.

Several vulnerable areas have been identified where creep rupture under these severe thermal-mechanical loads following the onset of core damage could occur. From the Surry SOARCA study (Sandia National Laboratories, 2013), the following are the key areas of the Reactor Coolant System most vulnerable to experiencing an induced creep rupture.

• Hot leg nozzle carbon safe zone

• Hot leg piping

• Surge line

• Steam generator inlet tubes 55

To characterize the potential for an induced creep rupture to occur at these different locations, significant analytical investigations of natural circulation and heat transfer occurred initially under the Surry SOARCA study (Sandia National Laboratories, 2013). These detailed models were applied in the Surry SOARCA uncertainty study (Ross, Bixler, Weber, Sallaberry, &

Jones).

• In 930 realizations (93% of realizations), a hot leg nozzle creep rupture was identified as occurring

• In 104 realizations (10% of realizations) a steam generator tube rupture occurred

• For every realization where a steam generator tube rupture occurred, a hot leg nozzle creep rupture also occurred A similar study of induced failure of the Reactor Coolant System for the Sequoyah plant was conducted as part of the Sequoyah SOARCA uncertainty study (Sandia National Laboratories).

Like the Surry study (Ross, Bixler, Weber, Sallaberry, & Jones), most realizations exhibited a hot leg nozzle creep rupture.

These past studies highlight the strong likelihood for an induced hot leg creep rupture to occur.

In this source term study, a hot leg creep rupture is treated as a dominant failure mode, with the induced steam generator tube creep rupture not represented. This is intended to capture the most likely scenarios for the purposes of estimating in-containment source terms. Hot leg creep rupture is represented in the MELCOR model using a standard Larson-Miller creep rupture model.

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Figure 3-2 PWR Reactor Circulation System Natural Circulation Flows It is important to note that this is an evolution in modeling that has occurred since both NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) and the more recent source term studies in SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011) and SAND2016-12954 (Gauntt, Goldman, Kalanich, & Powers, 2016). In these past studies, the likely induced failure of the hot leg nozzle following the onset of core damage was not captured explicitly in the modeling. For non-LOCA scenarios, this resulted in accidents progressing to lower head failure, and ex-vessel debris relocation, with the RPV at high pressure. In these types of high-pressure sequences, the transport of steam and hydrogen, as well as fission products, into containment is generally limited during in-vessel accident progression. By contrast, accident scenarios with an initiating LOCA or a later induced LOCA, have flow paths through which steam and hydrogen, as well as fission products, can be more readily transported into containment.

For these low-pressure scenarios, the in-containment source term is expected to be greater during the in-vessel phase of the accident when compared to high-pressure scenarios.

However, upon RPV lower head breach in high-pressure scenarios, it is expected that a more substantial release of fission products to containment will occur as they are swept out of the vessel into the containment.

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The total in-containment source term across all accident phases for low-pressure and high-pressure accident scenarios is likely similar. For low-pressure scenarios, however, the in-containment source term is likely to be larger during the in-vessel phase of accident progression.

Further discussion will be provided below in the analysis of the in-containment source term estimates derived from MELCOR simulations of Surry and Sequoyah accident scenarios.

3.2.2. Treatment of Fuel Degradation and Relocation Under SOARCA, an additional model was added to characterize the structural integrity of fuel rods under highly degraded conditions. This new modeling is intended to account for the thermal-mechanical weakening of the cladding oxide shell that occurs over time when a fuel rod is at elevated temperature. As the local cladding oxide temperature increases above the Zircaloy melting temperature (i.e., 2098 K in MELCOR) towards 2500 K, a thermal lifetime function accrues increasing damage from 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> until a local thermal-mechanical failure occurs. Similar time-at-temperature failure criteria are applied to oxidized channel boxes in the case of a BWRthe collapse of channel boxes into particulate debris typically occurs at a different time from the collapse of fuel assemblies in the same radial ring of the core.

In addition to this time-at-temperature representation of fuel rod failure under degraded conditions, the MELCOR modeling represents the loss of integrity that can occur due to material interactions between dissimilar fuel rod materials at elevated temperatures. This analysis utilized a material interactions model, termed the interactive model, consistent with application in the previous SOARCA studies. In this modeling approach, material liquefaction temperatures for fuel, cladding, and B4C (in the case of BWR control blades) are modified to represent the lower temperatures at which liquefaction of these materials can occur.

For B4C control blades, the liquefaction temperature of 1500 K is assumed based on the observed melting at this temperature of this control blade material due to eutectic interaction with the stainless-steel sheath. Eutectic interaction of fuel with oxidized cladding has been noted in multiple experimental studies to lead to much lower temperature liquefaction of fuel and cladding than the individual material melting temperatures. In this analysis, a liquefaction temperature of 2479 K was assumed for fuel and cladding. This value is consistent with that assumed in the SOARCA UA study.

3.2.3. Lower Plenum Debris/Coolant Heat Transfer and RPV Lower Head Failure Interaction of water in the lower plenum with overheated core debris that has slumped into the lower plenum is critical to represent as it plays an important role in subsequent in-vessel accident progression and evolution to lower head breach and initiation of ex-vessel damage conditions.

The lower plenum heat transfer settings were updated to reflect the end-state thermal condition of the debris in the deep pool FARO tests (i.e., significant thermal interaction with the water).

The resultant behavior results in debris fragmentation and cooling if there is a pool of water in the lower plenum. Debris temperatures can subsequently increase after the pool of water evaporates, which in turn heats up the reactor vessel lower head.

• Lower plenum particulate debris heat transfer. All particulate debris in the lower plenum is permitted to be in contact with water if present. In previous versions of the code, a restrictive one-dimensional counter-current flooding limitation (CCFL) criterion prevented 58

water from penetrating a deep debris bed. This restriction has been removed, effectively assuming steam rising upward from the debris bed does not totally preclude water from entering lower regions of the bed (e.g., via lateral flow from peripheral regions of the lower plenum). This will not appreciably impact calculations since reflood of the vessel is not being considered for the scenarios treated in this study.

• Lower head failure. The mechanical response of the vessel lower head is modeled using a one-dimensional creep rupture model. A Larson-Miller failure criterion is calculated based on the one-dimensional conduction and stress profile through the lower head. Failure of a lower head penetration prior to gross head failure is not explicitly modeled in this analysis, consistent with SOARCA calculations. Prior analyses of BWR lower head penetration failure mechanisms indicate the drain line is more susceptible to failure than other penetrations (e.g.,

CRD and in-core instrument penetrations), particularly for the depressurized conditions observed in the accident progression analysis of this analysis. While penetration failure is a potential outcome, the modification in vessel breach timing is not expected to be significant enough to modify the overall in-containment source term estimates obtained in this analysis.

3.2.4. Core Plate Failure The timing of core plate failure affects the relocation of the degraded core materials from the core region into the lower plenum. The local thermal-mechanical failure of the lower core plate, the flow mixer plate, and the lower support forging are calculated within MELCOR using the Roark engineering stress formulae. The yield stress is calculated based on the loading and local temperature.

3.2.5. Ex-Vessel Molten Corium-Concrete Interaction (MCCI)

An evaluation of typical MELCOR calculations of ex-vessel debris behavior when debris is submerged in a pool of water concluded the default treatment of heat transfer between debris and an overlying pool of water was not consistent with observations from the MACE tests. The default value for the debris-water interface heat transfer coefficient in MELCOR did not account for multi-dimensional effects of fissures, other surface non-uniformities, and side heat fluxes. An enhancement to the default value was used to more closely replicate heat transfer rates observed in the MACE tests. More recently, mechanistic water ingression and corium spreading models have been implemented in the MELCOR code. These have not been explicitly used as part of this analysis to remain consistent with the current state-of-practice established by SOARCA.

Though they can provide greater resolution of the effectiveness of accident management strategies with water injection, the current best-practice modeling approach utilizes models validated and appropriate for regulatory application to the unmitigated scenarios considered in this study.

3.2.6. Containment Failure Modeling 3.2.6.1. BWR Mark I Containment The Peach Bottom containment nodalization is shown in Figure 3-3. The corresponding nodalization of the reactor building is shown in Figure 3-4. In this nodalization diagram of the 59

reactor building, flow paths representing distinct containment failure modes are illustrated. The following containment failure modes are represented in the Peach Bottom MELCOR model:

• Drywell liner failure due to melt-through induced by relocating ex-vessel debris spreading into contact with the drywell liner

• Drywell head lifting inducing leakage out of the top of the drywell

• Torus over-pressure rupture

• Drywell liner shear that occurs because of significant containment over-temperature causing expansion of the drywell sufficient to shear penetration welds Figure 3-3 BWR Mark I Containment Nodalization (Gauntt, Radel, Salay, & Kalinich, 2008) 60

Figure 3-4 BWR Mark I Reactor Building Nodalization (including containment failure flow pathways) (Leonard, Gauntt, & Powers, 2007) 61

3.2.6.2. BWR Mark III Containment The Grand Gulf drywell nodalization is shown in Figure 3-5. The drywell is subdivided into 9 control volumes across 5 separate axial levels. Notable regions include the pedestal region beneath the RPV (CV202), the wetwell annulus region that connects the drywell to the wetwell through a series of 3 vertical vents (CV203), and the drywell head (CV209).

Figure 3-5 BWR Mark III drywell nodalization (Gauntt, Radel, Salay, & Kalinich, 2008)

The corresponding nodalization of the containment is shown in Figure 3-6. The containment is subdivided into 18 control volumes. A single control volume is used to model enclosure building, and 2 control volumes are used for the environment. Containment failure modes considered in this analysis include late overpressurization of containment and early failure during a hydrogen deflagration event.

62

Figure 3-6 BWR Mark III Containment Nodalization (Gauntt, Radel, Salay, & Kalinich, 2008) 3.2.6.3. PWR Large-Dry Containment The containment is divided into a total of 9 control volumes and 17 flow paths. Figure 3-7 and Figure 3-8 show the hydrodynamic nodalization of the containment. The control volumes represent the basement, the cavity under the reactor, the three separate steam generator cubicles, the pressurizer cubicle, the pressurizer relief tank (PRT) cubicle, the lower dome, and the upper dome. The basement region includes the bottom part of the containment as well as the surrounding cavity that lies between the outer wall and internal crane wall.

Testing has shown that concrete containments start to leak at leak rates much higher than design leakage and well before a large rupture or gross failure would occur. This leakage could preclude the large rupture or failure. The same leakage model as developed for the Surry SOARCA study (Sandia National Laboratories, 2013) has been adopted for this source term study.

The location of the leakage can have a significant effect on the results of the severe accident analysis and dose rates. For instance, if the containment leakage occurs through penetrations that are located inside adjoining plant buildings, the fission product release into atmosphere would be significantly less as compared to direct leakage to the environment. Previously, some of the severe accident analyses assumed that the leakage takes place at the top of the containment dome. A more realistic approach is to consider leakage to occur at the equipment hatch. This was 63

assumed in the Surry SOARCA study (Sandia National Laboratories, 2013). The same assumption is adopted in the current source term study. Leakage through the equipment hatch discharges into the environment from the side of the containment dome.

Containment failure modes considered in this analysis include late overpressurization of containment and early failure at the time of vessel failure.

Figure 3-7 PWR Large-Dry Containment Nodalization 64

Figure 3-8 PWR Large-Dry Containment Nodalization (Plan View) 3.2.6.4. PWR Ice Condenser Containment The Sequoyah containment is a free-standing steel containment consisting of a cylinder topped by a hemispherical dome as shown in the containment nodalization diagram in Figure 3-9. The containment is divided into a total of 28 control volumes and 53 flow paths. 18 of these control volumes represent the ice condensers, and the rest represent the reactor cavity, steam generator rooms, the pressurizer room, the lower containment, lower annulus, and the upper dome.

The mean containment failure pressure developed from an aggregate distribution of expert elicited failure pressures in NUREG/CR-4551 (Sandia National Laboratories, 1990) is 0.549 MPa(a). Rupture of the containment is assumed to result in a failure of area 3 ft2.

Failure modes considered in this analysis are overpressurization and early failure at the time of the first hydrogen deflagration.

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Figure 3-9 PWR Ice Condenser Containment Nodalization 3.2.7. Fission Product Release Kinetics To adequately characterize the source term to containment during a severe accident, it is necessary to represent the kinetics associated with fission product release from overheated fuel.

Of significance to this analysis is the adequacy of models for representing fission product release kinetics for fuel at elevated burnups. SAND2016-12954 (Gauntt, Goldman, Kalanich, & Powers, 2016) provides a discussion of fission product release kinetics for fuel at elevated burnups. The main content presented in SAND2016-12954 (Gauntt, Goldman, Kalanich, & Powers, 2016) is provided below to support the overall approach adopted in this analysis.

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Limited data exists on the behavior of HBU or HALEU/HBU fuel under the severe conditions experienced during an accident with significant core melting. As noted in SAND2016-12954 (Gauntt, Goldman, Kalanich, & Powers, 2016), the VERCORS RT-6 test provides information of relevance to understanding fission product release kinetics from HBU fuel under conditions occurring during a severe accident. This test was performed using fuel with a burnup of about 55 GWd/MTU that originated from the Fessenheim reactor in France.

In this experiment, the release of Cs and several other fission products was measured from small re-irradiated samples. These release tests were conducted with an imposed slow temperature excursion. The fractional release of Cs and the temperature excursion to which a fuel sample was subjected are shown in Figure 3-10. Note that the units in Figure 3-10 have been suppressed due to the proprietary nature of the test data.

Figure 3-10 RT-6 Release of Cesium as a function of test sample temperature (Gauntt, Goldman, Kalanich, & Powers, 2016) 3.2.7.1. Fission Product Release Modeling from LBU and HBU Fuel MELCOR incorporates several fission product release models. As in past studies (SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011) and SAND2016-12954 (Gauntt, Goldman, Kalanich, & Powers, 2016)), this analysis uses the Booth diffusional release model. This fission product release model is based on a mechanistic representation of the diffusional release process. While some MELCOR studies have utilized the simplified CORSOR fractional release rate correlation-based model, the more mechanistic nature of the Booth model is more appropriate for application to release from HBU and HALEU/HBU fuel for which fewer experiments have been conducted.

The MELCOR implementation of the Booth model treats the release of Cs from fuel as a diffusion process combined with a mass transport limitation to the local atmosphere. In the first 67

stage of the process, Cs diffuses through the fuel matrix to the surface of a fuel grain. At the fuel grain, Cs can vaporize into the local atmosphere. The rate of vaporization is limited based on the species vapor pressure.

The Cs fractional release at a time is implemented in MELCOR as the solution to the diffusion assuming diffusion through fuel grains having a spherical geometry.

! 1

6) 3!  ! <

= " (3-1) 1 6 exp( " ! )  ! >

1 where:

! 8" is the definition of the effective diffusion coefficient in terms of the temperature-dependent diffusion coefficient is the equivalent sphere radius for the fuel grain The temperature-dependent diffusion coefficient is assumed to follow an Arrhenius form

= # exp 9 = (3-2) where:

is the universal gas constant is the temperature is the activation energy

1. is a pre-exponential factor that is a function of the fuel burnup Experimental data for fractional release as a function of inverse temperature is typically correlated to the Booth model following the procedure of Lorenz and Osborne (Lorenz &

Osborne, 1995). The procedure is illustrated in Figure 3-11 which shows the correlation of the instantaneous diffusion coefficient ! to the inverse temperature.

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Figure 3-11 Booth Model Fits to RT-6 instantaneous diffusion coefficients (Gauntt, Goldman, Kalanich, & Powers, 2016)

The coefficients of the Arrhenius model for the diffusion coefficient in Equation (3-2) are then adjusted to ensure a best fit for the instantaneous diffusion coefficient in Figure 3-11 but also the fractional release as a function of time shown in Figure 3-12.

With this procedure, the specification of the HBU Booth model is obtained. The parameters for the HBU and LBU Booth models are provided in Table 3-1. The equivalent sphere radius corresponding to these parameters is 6 .

Table 3-1 HBU and LBU Booth Model Parameters Model [ ] [ ]

LBU Booth model (ORNL-Booth) 1x10-6 3.814x105 HBU Booth model 2.3x10-9 2.411x105 The comparison of this specified model with VERCORS RT-6 data and the Booth model specified for LBU fuel (i.e., the ORNL-Booth model) is shown in Figure 3-12.

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Figure 3-12 Comparison of RT-6 Cs Fractional Release Measurements against HBU and LBU Booth Model Predictions (Gauntt, Goldman, Kalanich, & Powers, 2016)

The comparison shown in Figure 3-12 indicates that the HBU Booth model predicts fractional release well at both low and high temperatures. Intermediate temperatures exhibit an overall under-prediction of the fractional release from the fuel. However, the HBU Booth model shows a reasonable prediction especially when contrasted with the significant under-prediction of the ORNL-Booth model for LBU fuel. As noted in SAND2016-12954 (Gauntt, Goldman, Kalanich,

& Powers, 2016), the HBU Booth model predicts the key features of release from HBU fuel

• Initiation of fission product release at lower temperatures than exhibited for LBU fuel

• Near total release of Cs at high temperatures For other fission products, the fractional release rate in the MELCOR implementation of the Booth model is scaled to the Cs fractional release rate to match the experimentally observed scaling.

The fractional release rates estimated by the HBU Booth model do exhibit an earlier and more rapid rate of release from fuel compared with the LBU Booth model (ORNL-Booth). While the release rate may appear to lead to the potential for more severe fission product release from higher burnup fuel, there are a broad range of additional factors that control the extent of fission product release from fuel during a severe accident. These ultimately tend to dominate the degree to which fission products are released from the fuel, but most importantly the degree to which fission products can migrate into containment. Additional factors are discussed below to provide further insights into the in-containment source term results presented in Section 5.

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3.2.8. Cesium and Iodine Chemical Forms NUREG-1465 adopted the assumption that 5% of iodine releases would be gaseous including I2 and volatile organic iodides. The rest of the iodine inventory was assumed to be released as metallic iodide (i.e., CsI). After NUREG-1465, significant insights regarding fission product speciation and volatility were identified by the Phébus FP program (Clement & Haste, 2004).

Specifically, it was hypothesized, based on thermodynamic calculations, that Cs and Mo react during severe accidents to form cesium molybdate (Cs2MoO4), and only a small fraction of I is released as gaseous iodine with the rest being released as metallic iodide such as CsI.

The predominant speciation of cesium was changed based on detailed analysis of the deposition and transport of the volatile fission products in the Phébus facility tests (Clement & Haste, 2004). The analysis revealed molybdenum combined with cesium and formed Cs2MoO4.

Previously, the default predominant chemical form of cesium was cesium hydroxide (CsOH). As consistent with past studies, all the released iodine combines with the cesium. Applications of this information to the MELCOR models used in the SOARCA calculations are described in SAND2010-1633 (Gauntt R. , 2010).

Gaseous iodine remains an uncertain source term issue, especially with respect to long-term radioactive release mitigation issues after the comparatively much larger airborne aerosol radioactivity has settled from the atmosphere. The mechanistic modeling treatment for gaseous iodine behavior is a technology still under development with important international research programs underway to determine the dynamic behavior of iodine chemistry with respect to paints, wetted surfaces, buffered and unbuffered water pools undergoing radiolysis, and gas phase chemistry. Results from the Phébus FPT3 experiment suggest that iodine speciation may be impacted by the presence of B4C control rod (Haste, et al., 2013). In particular, the gaseous iodine species were observed to make up a larger fraction of the total iodine releases than aerosol species. It is important to note that the stainless-steel mass present in the FPT3 experiment may not be representative of prototypic BWR core conditions including the relative ratios of B4C and stainless steel. Other experimental programs, including DF-4 and BECARRE, indicate that the presence of stainless steel may function to reduce the impact of B4C on gaseous iodine release.

(Gauntt R. G., 1989) (Dominguez, 2012). The treatment followed in this analysis is consistent with SOARCA best practices.

In its current implementation, MELCOR does not have an active chemistry model (mass transfer between predefined radionuclide classes by dedicated models is possible, and users can implement their own mass transfer control functions for untreated classwise transfers). The MELCOR code approximates radionuclide releases for the spectrum of radionuclides through representative radionuclide groups such as Alkali Metals (representative species CsOH), Mo Group (representative species Mo), and Cesium Molybdate (representative species Cs2MoO4).

Within the current framework, early release of CsOH occurs at the time of gap release, coincident with CsI gap release. Delayed release of remaining CsI and Cs2MoO4 occurs only as core damage proceeds and those species are predicted to vaporize from core debris according to their vapor pressure curves as shown in Figure 3-13.

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Figure 3-13 MELCOR Vapor Pressure curves for RN classes CsM (includes Cs2MoO4), CsI, and Mo This analysis follows the practices outlined in the SOARCA project which meet the findings of NUREG/CR-7110 (Bixler, Gauntt, Jones, & Leonard, 2013). This analysis assumes that

• 100% of the total Iodine inventory reacts with Cesium to form CsI

• 5% of the total Iodine inventory is present as gap inventory o All Iodine present in the gap inventory is reacted CsI

• 5% of the total Cesium inventory is present as gap inventory o Cesium already present in the gap inventory in the form of CsI contributes to this fraction o Any remaining Cesium fraction (5% total Cesium minus % Cs in CsI gap inventory) is reacted CsOH

• All remaining Cesium (95%) is assumed to react with Molybdenum and be released as Cs2MoO4, consistent with past studies SAND2011-0128 (Powers, Leonard, Gauntt, Lee, &

Salay, 2011) and SAND2016-12954 (Gauntt, Goldman, Kalanich, & Powers, 2016) but distinct from NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) 3.3. Accident Tolerant/High Burnup Fuel Severe Accident Phenomena Identification and Ranking Table Based on the PIRT for HBU and/or HBU/HALEU fuel designs (Table 3.42 of NUREG/CR-7283 (Khatib-Rahbar, 2021)), and the discussions of the findings in NUREG/CR-7283 (Khatib-Rahbar, 2021), additional MELCOR sensitivity calculations have been identified.

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In general, there are a small number of issues identified in NUREG/CR-7283 (Khatib-Rahbar, 2021) that differentiate conventional and HBU/HALEU fuel designs with respect to severe accidents and radiological release. Despite some differences between conventional and HBU/HALEU fuel designs being identified in NUREG/CR-7283 (Khatib-Rahbar, 2021),

analytical capabilities exist to assess the impact of these differences on regulatory source terms.

The areas where differences between conventional and HBU/HALEU fuel designs could potentially impact regulatory source terms can be evaluated through MELCOR sensitivity studies. These sensitivity studies assess how specific uncertainties of these fuel designs could impact the overall in-containment source term. They do not represent a characterization of generic severe accident uncertainties (e.g., limitations in the current-state-of-knowledge regarding fission product speciation) as these have been extensively investigated in the context of other studies such as SOARCA. The following MELCOR sensitivity studies have been identified to resolve the impact of these differences:

• Alteration of thermophysical properties of both fuel and cladding.

Issue: Stored heat at the start of a transient is directly affected by fuel and cladding thermal conductivity and specific heat. The effects of fuel thermal conductivity and stored heat have been found to influence the thermal response of the fuel and associated cladding under the conditions of design basis loss of coolant accidents (LOCA). The EPRI Accident Tolerant Fuel (ATF) Safety Benefits Study (Electric Power Research Institute, 2019. 3002015091) did not find a significant impact of thermophysical properties on the course of accident progression for the ATF concepts investigated. This analysis, however, did not investigate how fission product release for ATF concepts may have been altered relative to conventional fuels. To resolve this issue identified in the ATF Severe Accident PIRT, MELCOR analytical capabilities exist to assess the effect on fuel response and fission product release under severe accident conditions.

Sensitivity Study: Alteration of fuel thermal conductivity (reduction to approximately 40%

of nominal value) to reflect higher burnup levels.

o Fuel thermal conductivity tends to decrease with increasing burnup. Fuel and cladding thermal conductivity can be treated parametrically to assess the effects on accident progression and resulting release and transport of radionuclides to containment. A correlation that is available in the FAST fuel performance code illustrates the degradation of the fuel thermal conductivity as a function of fuel burnup.

• HBU fuel is expected to have a different amount of fragmentation or sintering.

Issue: It is postulated that the increased fragmentation or sintering of HBU fuel could lead to failure of the fuel under somewhat less severe conditions. This would result in earlier onset of fuel failure.

Sensitivity Study: Different sensitivities are considered to evaluate the impact of both earlier fuel failures and collapse into particulate debris beds that have more limited porosity and are comprised of smaller rubble particles.

(1). Sensitivity to particulate debris bed porosity. This can be directly evaluated through adjustment of the user-specified debris bed porosity.

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(2). Sensitivity to particulate debris bed particle sizes. This can be directly evaluated through adjustment of debris particle diameters in the core-region and lower plenum.

(3). Adjustment of fuel failure temperature due to liquefaction associated with material interactions. This can be readily adjusted through modification of the effective liquefaction temperatures of fuel and cladding.

(4). Sensitivity to the time-at-temperature failure model introduced to capture the observed collapse of degraded fuel rods prior to substantial melting. This can be directly adjusted through modification of the time-at-temperature lifetime table in MELCOR plant input models.

• Increased cladding embrittlement.

Issue: Increased cladding embrittlement will affect core coolability under core reflood conditions. For increased cladding embrittlement, it is possible that the mechanical loading on fuel when it is quenched during a core reflood could lead to fuel rod collapse into a particle bed. This configuration is generally accepted to have greater uncertainty regarding the degree to which it can be cooled relative to an intact fuel rod geometry. The coolability of a particle bed is determined by the extent to which it has sufficient surface area exposed to cooling water relative to the volume of heat generating fuelthe ratio of heat transfer surface area to debris volume. At high ratios, the debris bed will be more coolable than at low ratios.

The porosity and particle sizes characterizing these debris beds control this surface area to volume ratio.

However, given the present analyses are not focused on post-accident mitigation strategies (e.g., reflood), an evaluation of the impact of increased cladding embrittlement is not considered applicable to the present assessment of accident source term for HBU fuels.

Sensitivity Study: Should studies of accident mitigation be considered in future studies, the impact of cladding embrittlement on accident progression and source term could be resolved through reduction of the porosity and increase of the particle size in MELCOR simulations.

3.4. Non-parametric Approach This analysis employed a non-parametric bootstrap methodology to estimate source terms from a relatively small number of simulations. The method was previously applied for source terms from high burnup or MOX fuel in 2011 (Powers, Leonard, Gauntt, Lee, & Salay, 2011) and detailed in 2020 (Brooks, 2020). The method is non-parametric because it can be applied for simulation results that follow any distribution, standard or otherwise. Instead, it uses repeated re-sampling (bootstrapping) to describe uncertainty around the empirical cumulative distribution function (ECDF) defined by the simulations. The ECDF at any point is just the fraction of the sample less than or equal to that point in value, so the smallest and largest points in the set of simulations define the ends of the distribution. This means that the tail behavior of the distribution estimated using this method can be sensitive to sample size.

The method, as described in (Brooks, 2020), includes the following basic steps:

1. Run -many simulations to get source term estimates.

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2. Sample from the empirical distribution defined by these source term estimates using bootstrap sampling with a large sample size. This involves drawing samples, each of size .

3. For each of the bootstrap sample, calculate the 5th, 10th,,95th percentiles of the sample.

This results in samples of each of the 5th, 10th,,95th percentiles of the distribution.

4. At each percentile, calculate the mean and standard deviation over the N samples. These define uncertainty around the estimate of each percentile.

5. Define the distribution estimate by plotting the means calculated in the previous step versus the cumulative probabilities 0.05, 0.10,,0.95.

6. Define the uncertainty bounds on the distribution by adding and subtracting the standard deviation from the mean at each percentile. If higher coverage is desired, multiples of the standard deviation can be used.

7. Apply an interpolation method to estimate the cumulative probabilities for the simulation data points except for the minimum and maximum simulation data points.

8. Use either an assumption or an extrapolation method to define cumulative probabilities for the minimum and maximum simulation data points.

This analysis considered many scenarios, sites, and radionuclide inventory types. Simulations were grouped to constitute the simulations for a case of interest. For example, seven scenarios were simulated for Grand Gulf and nine scenarios were simulated for Peach Bottom, giving a total of 16 BWR scenarios. Each scenario was simulated for two HBU inventories, so there were

= 32 simulations used to predict BWR HBU source terms using the process defined above.

Regardless of the value of for each case, = 10,000 bootstrap samples were used.

Most steps in the method are straightforward but there is flexibility in application of the second step. For this analysis, sampling of the ECDF was performed by sampling from a uniform distribution on [0,1] and then interpolating the ECDF using a smoothing spline to obtain the corresponding source term for that cumulative probability sample. This can be thought of as smoothing the ECDF and sampling new data points between the existing simulations.

The result of this process is a smoothed mean ECDF on the simulation points with uncertainty bands. The ECDF itself characterizes uncertainty that can be attributed to differences between sites and scenarios, which can be thought of as aleatory or stochastic (random) uncertainty. The bounds on the ECDF characterize sampling uncertainty.

Though the method provides significant information on uncertainty, individual points can be used for comparison. This analysis uses the 50th percentile of the mean ECDF for comparison to previous source term analyses (see Section 5).

3.5. Additional Methodological Assumptions The following additional assumptions are important to note as part of this analysis, assumptions listed in other sections of this report are repeated here for clarity:

• The in-containment source term presented in this analysis does not examine (1) start time of the radionuclide release, (2) variation in the gap inventory at the start of the accident, (3) 75

fraction of aerosolized iodine in containment, or (4) radionuclide removal and retention in containment.

• In general, the MELCOR plant model, modeling approach, and best practices follow the approach established in SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011) with enhancements developed during the SOARCA uncertainty project, as well as post 9/11, and post-Fukushima Daiichi regulatory studies.

• Source term analyses are based on the current state-of-the-art as captured by MELCOR 2.2 and modeling best-practices established under SOARCA with the following exceptions, details are provided in Appendix A:

o The time-at-temperature fuel rod failure model employed in this analysis uses the MELCOR default time-at-temperature fuel rod lifetime curve.

o UO2 and ZrO2 liquefaction temperatures have been reduced to 2479.0 K to account for material interactions.

o Failure temperature of oxidized fuel rods have been reduced to 2479.0 K.

• Extended enrichment HBU cores do not change the relative contribution of different accident sequences to the total CDF of BWRs or PWRs.

• The only source of uncertainty considered is associated with the range of possible accidents that could be realized (i.e., aleatory uncertainty).

o Phenomenological (or epistemic) uncertainty is not considered in determining representative BWR and PWR source terms11.

• Containment scrubbing mechanisms (e.g., the suppression pool, containment sprays, etc.) are not credited in the presented results.

• Release fractions (source terms) between 0.0 and 1.0x10-6 are considered negligibly small for the purpose of this analysis and are truncated.

• Source terms are classified into distinct radiological release phases originally identified in the NUREG-1465 study and shown in Table 3-2.

Table 3-2 Phase Timing Criteria Phase Onset Criteria End Criteria Gap Release RPV water level below top of active fuel Release of 5% of initial, total Xe inventory from fuel Early In-Vessel Release of 5% of initial, total Xe inventory from fuel Lower Head Failure Ex-Vessel Lower Head Failure 95% of total ex-vessel Cs releases Late In-Vessel Lower Head Failure 95% of total late in-vessel Cs releases

• The radionuclide chemical classes considered in this analysis are like that originally considered in the NUREG-1465 study.

11 Key phenomena identified in a PIRT study are investigated through sensitivity studies 76

o In this analysis, however, platinoids (Ru) and early transition elements (Mo) are classified into distinct radionuclide classes consistent with the state-of-the-art in fission product modeling that developed following the Phébus tests (Clement &

Haste, 2004).

• The source term release phase timings and magnitudes for the different chemical classes are assumed to be at the level of fidelity appropriate for both BWRs and PWRs.

• The source term release phase timings and magnitudes for the different chemical classes are reported as median values drawn from the distribution of possible source terms that could be realized across the range of different accident sequences that lead to significant core damage and in-containment source terms.

o The median value has the benefit of weighting the source terms across all accident types equally.

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4. ACCIDENT SELECTION 4.1. Selection of Representative Accident Scenarios A spectrum of accident sequences that lead to significant core damage and in-containment source terms is presented to support the development of a source term for HBU fuel applications. Every scenario produces an accident progression characterized by a substantial core-melt accident and radionuclide releases to the containment structure commensurate with each accident release phase. The accident sequences are comprised of an accident initiating event, defined system availability/functionality, and prescribed component failures to address uncertainties in component failure. SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011) identified a set of accident sequences that represents the major contributors to core damage frequency at the selected plants, as described in references (Leonard, Gauntt, & Powers, 2007) and (Ashbaugh, Leonard, Longmire, Gauntt, & Powers, 2010). Since this work updates the source terms presented in SAND2011-0128, the same set of accident sequences are used in this evaluation.

The accident initiating events shown to have significant risk contribution based on the Individual Plant Examinations (IPE) informed the scenarios identified for each reactor type. Several accident initiating events from the IPE are excluded from the analysis. These exclusions are due to similar scenarios among the BWR or PWR types or due to the scenario, if credible, not supporting a substantial meltdown accident capable of resulting in an ex-vessel accident phase.

These criteria are most likely unique to the analysis of this analysis, given the distinction of site-specific accident events being obfuscated after a representative source term is calculated for the given reactor type. Additionally, bypass and air ingression scenarios are not prescribed in this analysis. It is assumed that the selected accident scenarios and their respective contribution to representative in-containment source terms are only negligibly impacted by HBU fuel and any corresponding changes to plant operations.

Beyond accident initiating events, uncertainty in accident progression is addressed by prescribing system failures consistent with modern expectations and recent findings. These prescriptions are therefore part of the accident scenario. Although they are not expected to significantly affect fission product release to containment, these prescriptions are maintained for consistency with SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011). The analyses prescribe reactor pressure boundary and containment pressure boundary failures to account for the effects of stochastic pressure relief device failures on the primary system as well as differing containment failure mechanism, respectively. While stuck open relief device allowance is enabled or disabled, temperature induced failure of the RCS is always permitted, and may still result. The following containment failure modes are captured in the selected accident scenarios where appropriate based on specific reactor/containment types:

• Direct debris interaction (e.g., drywell liner melt)

• Over-pressurization

• Head flange leakage

• Thermally induced failures

• Hydrogen explosions (assumed deflagration at time of ex-vessel) 79

Although containment failures are specified in the selected accident scenarios for consistency with previous practices, their impact is expected to predominately affect the late in-vessel and ex-vessel accident phases; minimal impact is expected for the gap release or early in-vessel accident phases.

In summary, 16 BWR and 12 PWR accident scenarios are considered. BWR scenarios include one anticipated transient without scram (ATWS) with low RPV pressure at vessel breach, 3 LOCA with low RPV pressure at vessel breach, and 12 station blackout (SBO) scenarios 10 of which exhibit low RPV pressure at vessel breach due to SRV seizure and 2 of which exhibit high RPV pressure at vessel breach. PWR scenarios include eight LOCA scenarios with low RPV pressure at vessel breach and four SBO scenarios that do not specify RPV pressure at the time of failure (modeling approach enables HLCR if predicted). The limited number of high-pressure scenarios reflects the currently understood predominance of low-pressure scenarios in the spectrum of credible severe accidents.

Brief descriptions of the selected accident scenarios and simplified event trees are provided for each reactor in the following sections.

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Table 4-1 Boiling Water Reactor High Burnup Accident Sequence Matrix Initiating DC Coolant RPV Pressure at Case Plant Event Power Injection Vessel Breach Containment Failure Other Peach Low Drywell liner melt OR drywell head 1A Bottom STSBO No None Stuck-open SRV flange leakage*

Peach Low 1B Bottom STSBO No None Stuck-open SRV Drywell liner melt Peach Low 1C Bottom STSBO No None Stuck-open SRV Drywell head flange leakage Basaltic concrete*

Peach 1D Bottom STSBO No None High Drywell liner melt Peach Low 2A Bottom LTSBO 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> RCIC only Stuck-open SRV Drywell liner melt Peach Low 2B Bottom LTSBO 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> RCIC only Stuck-open SRV Drywell head flange leakage Peach Low 2C Bottom LTSBO 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> RCIC only Stuck-open SRV Overpressure of torus Peach Low 3 Bottom SBLOCA Yes None LOCA Drywell head flange leakage Small break in main steam line Peach Low 4 Bottom LBLOCA Yes None LOCA Drywell liner melt Recirculation suction line LOCA Low Early 5A Grand Gulf STSBO No None Stuck-open SRV H2 burn at Vessel Breach H2 burn also causes failure of drywell wall Early 5B Grand Gulf STSBO No None High H2 burn at Vessel Breach H2 burn also causes failure of drywell wall Low Late 5C Grand Gulf STSBO No None Stuck-open SRV Overpressure Low Early 6A Grand Gulf LTSBO 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> RCIC only Stuck-open SRV H2 burn at Vessel Breach H2 burn also causes failure of drywell wall Low Late 6B Grand Gulf LTSBO 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> RCIC only Stuck-open SRV Overpressure Low 7 Grand Gulf ATWS Yes Yes ADS Prior to Onset of Core Damage MSIV closure Low Late 8 Grand Gulf LBLOCA Yes RCIC only LOCA Overpressure Recirculation suction line LOCA

• Different from (Leonard, Gauntt, & Powers, 2007).

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Table 4-2 Pressurized Water Reactor High Burnup Accident Sequence Matrix Containment Fan Cavity RCP Seal Containment Case Plant Initiating Event ECCS AFW Spray Cooler Flood Leak Failure 1A Surry STSBO No No No No 1B Surry SBLOCA No Yes No* No*

Injection-mode 1C Surry LBLOCA available Yes No* No 1D Surry STSBO No No No No No 1F Surry SBLOCA No Yes No* No* Early TDAFW-4A Sequoyah RCP Seal LOCA No Controlled Inj. Yes Yes Yes TDAFW-4B Sequoyah RCP Seal LOCA No Controlled Inj. Yes No Yes MDAFW 4C Sequoyah RCP Seal LOCA Inj. (requires ECCS) Yes Yes Yes TDAFW-4D Sequoyah STSBO No Uncontrolled No No 4E Sequoyah STSBO No No No No Early TDAFW-4F Sequoyah LBLOCA No Controlled No Yes 4G Sequoyah SBLOCA No No No No

• Different from (Ashbaugh, Leonard, Longmire, Gauntt, & Powers, 2010).

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4.2. BWR Accident Scenarios 4.2.1. BWR Mark I Containment Accident Scenarios Figure 4-1 shows a simplified event tree for the Peach Bottom accident sequences considered in this analysis. A high-level description of each sequence is provided in the paragraphs below.

Sequence PB1a is short-term station blackout (STSBO). No onsite direct current (DC) power is available during the duration of the accident. Coolant injection is also unavailable. The accident sequence characterizes the vessel as being at low pressure during vessel breach due to prior seizure of the safety/relief valve in the open position. Containment failure is assumed to occur initially due to either drywell liner melt-through or drywell head flange leakage. Drywell liner penetration due to shear failure is permissible as a subsequent containment failure.

Sequence PB1b is STSBO. No onsite DC power is available during the duration of the accident.

Coolant injection is also unavailable. The accident sequence characterizes the vessel as being at low pressure during vessel breach due to prior seizure of the safety/relief valve in the open position. Containment failure is assumed to occur initially due to drywell liner melt-through.

Drywell liner penetration due to shear failure is permissible as a subsequent containment failure.

Sequence PB1c is initiated by a STSBO. No alternating current (AC) or DC power is available during the duration of the accident. Coolant injection is also unavailable. This sequence is unique as the drywell floor uses non-site-specific Basaltic concrete. The accident sequence characterizes the vessel as being at low pressure during vessel breach due to prior seizure of the safety/relief valve in the open position. Containment failure is assumed to occur initially due to drywell head flange leakage. Drywell liner penetration due to shear failure is permissible as a subsequent containment failure.

Sequence PB1d is initiated by a STSBO. No AC or DC power is available during the duration of the accident. Coolant injection is also unavailable. The accident sequence characterizes the vessel as being at high pressure during vessel breach; all safety/relief values operate as designed in safety relief mode. Containment failure is assumed to occur initially due to drywell liner melt-through. Drywell liner penetration due to shear failure is permissible as a subsequent containment failure.

Sequence PB2a is initiated by a long-term station blackout (LTSBO). No AC power is available during the duration of the accident and battery power duration is eight hours. RCIC is operable during the early accident sequence. The sequence characterizes the vessel as being at low pressure during vessel breach due to prior seizure of the safety/relief valve in the open position.

Containment failure is assumed to occur initially due to drywell liner melt-through. Drywell liner penetration due to shear failure is permissible as a subsequent containment failure.

Sequence PB2b is initiated by a LTSBO. No AC power is available during the duration of the accident and battery power duration is eight hours. RCIC is operable during the early accident sequence. The sequence characterizes the vessel as being at low pressure during vessel breach due to prior seizure of the safety/relief valve in the open position. Containment failure is assumed to occur initially due to drywell head flange leakage. Drywell liner penetration due to shear failure is permissible as a subsequent containment failure.

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Sequence PB2c is initiated by a LTSBO. No AC power is available during the duration of the accident and battery power duration is eight hours. RCIC is operable during the early accident sequence. The sequence characterizes the vessel as being at low pressure during vessel breach due to prior seizure of the safety/relief valve in the open position. Containment failure is assumed to occur due to an overpressure of the wetwell torus. All high-temperature failure mechanisms of the containment are neglected.

Sequence PB3 is initiated by a small break LOCA in the form of a small break at the main steam line. AC power is available throughout the duration of the accident. It is, however, assumed coolant injection is unavailable. The sequence characterizes the vessel as being at low pressure during vessel breach due to the LOCA depressurizing the RPV. Containment failure is assumed to occur initially due to drywell head flange leakage. Drywell liner penetration due to shear failure is permissible as a subsequent containment failure.

Sequence PB4 is initiated by a large break LOCA in the form of a recirculation suction line break. AC power is available during the duration of the accident; however, it is assumed coolant injection is unavailable. The sequence characterizes the vessel as being at low pressure during vessel breach due to the LOCA depressurizing the RPV. Containment failure is assumed to occur initially due to drywell liner melt-through. Drywell liner penetration due to shear failure is permissible as a subsequent containment failure.

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DC Power RCIC DC Power RPV Sequence Accident Initiator Available Containment Failure Operation Lost Late Depressurized Label Early Liner Melt-through + Late Penetration Failure 2a SRV Stuck- Drywell Head Flange Leakage Y Y Y open + Late Penetration Failure 2b Torus Overpressure 2c Drywell Head Flange Leakage

+ Liner Melt-through + Late SBO Penetration Failure 1a SRV Stuck- Liner Melt-through + Late open Penetration Failure 1b Drywell Head Flange Leakage N N N/A + Late Penetration Failure 1c RPV not Liner Melt-through + Late depressurized Penetration Failure 1d Drywell Head Flange Leakage Small Main Steam Line Break Y N N LOCA + Late Penetration Failure 3 Liner Melt-through + Late Recirculation Suction Line Break Y N N LOCA Penetration Failure 4 Figure 4-1 Peach Bottom Event Progression for Characteristic Accident Scenarios 4.2.2. BWR Mark III Containment Accident Scenarios Figure 4-2 shows a simplified event tree for the Grand Gulf accident sequences considered in this analysis. A high-level description of each sequence is provided in the paragraphs below.

Sequence GG5a is initiated by a STSBO. No AC or DC power is available during the duration of the accident. Coolant injection is also unavailable. The accident sequence characterizes the vessel as being at low pressure during vessel breach due to prior seizure of the safety/relief valve in the open position. Containment failure is assumed to occur due to a deflagration event at the time of vessel failure.

Sequence GG5b is initiated by a STSBO. No AC or DC power is available during the duration of the accident. Coolant injection is also unavailable. The accident sequence characterizes the vessel as being at high pressure during vessel breach; all safety/relief values operate as designed in safety relief mode. Containment failure is assumed to occur due to a deflagration event at the time of vessel failure.

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Sequence GG5c is initiated by a STSBO. No AC or DC power is available during the duration of the accident. Coolant injection is also unavailable. The accident sequence characterizes the vessel as being at low pressure during vessel breach due to prior seizure of the safety/relief valve in the open position. Containment failure is assumed to occur due to late over-pressurization.

Sequence GG6a is initiated by a LTSBO. No AC power is available during the duration of the accident and battery power duration is eight hours. RCIC is operable during the early accident sequence. The accident sequence characterizes the vessel as being at low pressure during vessel breach due to prior seizure of the safety/relief valve in the open position. Containment failure is permitted to occur due to a deflagration event at the time of vessel failure.

Sequence GG6b is initiated by a LTSBO. No AC power is available during the duration of the accident and battery power duration is eight hours. RCIC is operable during the early accident sequence. The accident sequence characterizes the vessel as being at low pressure during vessel breach due to prior seizure of the safety/relief valve in the open position. Containment failure is assumed to occur due to late over-pressurization.

Sequence GG7 is initiated by an inadvertent closure of one MSIV; however, without a scram event. Power is available during the duration of the accident. Coolant injection is also available.

ADS actuation is assumed to result. Containment failure is to occur initially due to over-pressurization prior to onset of core damage.

Sequence GG8 is initiated large break loss of coolant accident due to the recirculation line breaking. Power is available during the duration of the accident and RCIC is operable throughout the transient, but no other coolant injection is available. Reactor vessel depressurization results from the break. Containment failure is to occur initially due to late over-pressurization.

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DC Power RCIC DC Power Sequence Accident Initiator RPV Depressurized Containment Failure Available Early Operation Lost Late Label H2 deflagration or Late over-pressurization 6a Y Y Y SRV Stuck-open Late over-pressurization 6b SBO H2 deflagration 5a SRV Stuck-open N N N/A Late over-pressurization 5c RPV not depressurized H2 deflagration 5b ATWS Y Y N ADS actuation Over-pressurization 7 Recirculation Suction Line Break Y Y N LOCA Late over-pressurization 8 Figure 4-2 Grand Gulf Event Progression for Characteristic Accident Scenarios 4.3. PWR Accident Scenarios 4.3.1. PWR Large-Dry Containment Accident Scenarios Figure 4-3 shows a simplified event tree for the Surry accident sequences considered in this analysis. A high-level description of each sequence is provided in the paragraphs below.

Sequence SU1a is initiated by a STSBO. No AC or DC power is available during the duration of the accident. The ECCS is also unavailable. Component cooling water is unavailable, leading to damage of the reactor coolant pump (RCP) seals which leak for the remainder of the scenario.

The containment sprays, fan coolers, and auxiliary feedwater systems are all unavailable.

Containment failure is assumed to occur at the time of vessel breach.

Sequence SU1b is initiated by a small break LOCA in the cold leg. AC and DC power are available throughout the accident. Component cooling water is available, preventing the reactor pump seals from being damaged. The ECCS fails, along with the containment spray system and fan coolers. The auxiliary feedwater system is available. Containment failure is assumed to only occur due to overpressurization.

Sequence SU1c is initiated by a large break LOCA in the hot leg. AC and DC power are available throughout the accident. Component cooling water is available, preventing the reactor pump seals from being damaged. The ECCS operates in injection mode only, as do the 87

containment spray and fan cooler systems. The auxiliary feedwater system is available. The containment is assumed to fail at the time of vessel breach.

Sequence SU1d is initiated by a STSBO. No AC or DC power is available during the duration of the accident. The ECCS is also unavailable. Component cooling water is available, preventing the reactor pump seals from being damaged. The containment sprays, fan coolers, and auxiliary feedwater systems are all unavailable. The containment is assumed to fail at the time of vessel breach.

Sequence SU1f is initiated by a small break LOCA in the cold leg. AC and DC power are available throughout the accident. Component cooling water is available, preventing the reactor pump seals from being damaged. The ECCS fails, along with the containment spray system and fan coolers. The auxiliary feedwater system is available. This sequence prescribes an early containment failure at the time of vessel breach.

RCP Containment Fan Containment Sequence Accident Initiator ECCS AFW seal Spray Cooler Failure Label leakage Y @VF 1A SBO N N N N N @VF 1D LBLOCA(10 hot leg break) Injection Mode Y N N N @VF 1C Overpressurization 1B SBLOCA (1 cold leg break) N Y N N N

@VF 1F Figure 4-3 Surry Event Progression for Characteristic Accident Scenarios 4.3.2. PWR Ice Condenser Containment Accident Scenarios Figure 4-4 shows a simplified event tree for the Sequoyah accident sequences considered in this analysis. A high-level description of each sequence is provided in the paragraphs below.

Sequence SQN4a is initiated by a RCP seal LOCA. AC and DC power availability is maintained through the transient. Coolant injection is achieved by controlled TDAFW operation, ECCS is unavailable. In containment, air return system fans are operational and containment sprays operate in injection mode only. Cavity flooding occurs when the lower containment is overfilled.

Early containment failure is not enforced.

Sequence SQN4b is initiated by a RCP seal LOCA. AC and DC power availability is maintained through the transient. Coolant injection is achieved by controlled TDAFW operation, ECCS is unavailable. In containment, the air return system is operational and containment sprays operate 88

by injection mode only. Cavity flooding does not occur. Early containment failure is not enforced.

Sequence SQN4c is initiated by an RCP seal LOCA. AC and DC power availability is maintained through the transient. Coolant injection is achieved by MDAFW operation, ECCS operates in injection mode only. In containment, the air return system is operational and containment sprays operate by both injection and recirculation modes. Cavity flooding does not occur. Early containment failure is not enforced.

Sequence SQN4d is initiated by a station blackout (SBO). AC and DC power are lost at the start of the transient. Some degree of coolant injection is achieved through uncontrolled TDAFW operation. In containment, the air return system and containment sprays are not available. Cavity flooding does not occur. Early containment failure is not enforced Sequence SQN4e is initiated by an SBO. AC and DC power are lost at the start of the transient.

Coolant injection is unavailable. In containment, the air return system and containment sprays are not available. Cavity flooding does not occur. Containment fails early at the time of the first hydrogen deflagration.

Sequence SQN4f is initiated by a large break LOCA. AC and DC power availability is maintained through the transient. Coolant injection is achieved by controlled TDAFW operation, ECCS is unavailable. In containment, the air return system is available, but containment sprays are not. Cavity flooding does not occur. Early containment failure is not enforced.

Sequence SQN4g is initiated by a small break LOCA. AC and DC power availability is maintained through the transient. Coolant injection is unavailable. In containment, the air return system and containment sprays are not available. Cavity flooding does not occur. Early containment failure is not enforced.

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Accident Initiator ECCS AFW Containment Air Cavity Early Sequence Sprays Return Flooding Containment Label System Failure Injection Mode MDAFW Y Y N N 4C RCP Seal LOCA Y N 4A N Controlled TDAFW Injection Mode Y N N 4B Uncontrolled TDAFW N N N N 4D SBO N N N N N Y 4E LBLOCA N Controlled TDAFW N Y N N 4F SBLOCA N N N N N N 4G Figure 4-4 Sequoyah Event Progression for Characteristic Accident Scenarios 4.4. NUREG-1465 Accident Scenarios Accident scenarios employed in the NUREG-1465 study are presented below for completeness, but the scenarios are limited to the nuclear power facilities analyzed within this analysis. These scenarios are provided as a matter of convenience and to relate the risk-significance as indicated NUREG-1560 Volume 2. While NUREG-1465 incorporated additional reactor and containment types, a reduced set of reactor-containment combinations was considered reasonable to produce a NUREG-1465 stylized source term. Further details on the NUREG-1465 accident sequences are presented in (Nourbakhsh, 1993).

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4.4.1. Boiling Water Reactors The relevant accident sequences presented in NUREG-1465 are provided in Table 4-3 for the BWRs.

Table 4-3 Risk-Significance of NUREG-1465 BWR Accident Sequences (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995)

NUREG- 1465 Risk Significance Plant Sequence Description Other Comments Sequence (NUREG-1560)

TC1 ATWS, reactor depressurized Low TC2 ATWS, reactor pressurized Low TC3 TC2, wetwell venting Low TB1 SBO, battery depletion High TB2 Same as TB1 except CF at VF High Pressure @ VF or shell Peach Bottom melt through S2E1 2-equivalent diameter LOCA, no ECCS, no ADS Low S2E2 Same as S2E1except Peach Bottom concrete Low replaced by basaltic concrete V ISLOCA Low TBUX SBO, loss of all DC power High TC ATWS, early CF fails ECCS Low TB1 SBO, battery depletion High Grand Gulf TB2 Same as TB1 except H2 burn- induced CF High TBS SBO, no ECCS, reactor depressurized High TBR Same as TBS except AC recovery after VF High 91

4.4.2. Pressurized Water Reactors The relevant accident sequences presented in NUREG-1465 are provided in Table 4-4 for the PWRs.

Table 4-4 Risk-Significance of NUREG-1465 Pressurized-Water Reactor Accident Sequences (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995)

NUREG- Risk Sequence Plant 1465 Significance Other Comments Description Sequence (NUREG-1560)

AG Hot leg LOCA, no CS, no FC Moderate LLOCAs currently thought to be minimally risk significant. However, induced creep-rupture failure (e.g.,

during SBO) more important.

TMLB LOOP, no PCS, no AFW High V ISLOCA Low IPE identification of potential bypass path led to operator training to Surry minimize risk.

S3B SBO, RCP seal LOCA High S2D-d SBLOCA, no ECCS, H2 High combustion S2D-b SBLOCA, 6-in. initial CF Not discussed S3HF1 RCP seal LOCA, no ECCS, no CS High Transient causes loss of pump seal recirc, cavity flooded cooling.

S3HF2 S3HF1, hot leg creep rupture Not discussed Transient causes loss of pump seal before VF cooling.

S3HF3 S3HF1, dry cavity Not discussed Some large-dry and ice condenser designs limit flow of water to cavity.

S3B 1/2-in.-equivalent diameter Low LOCA, SBO, no AFW Sequoyah TBA SBO, hot leg creep-rupture High before VF, H2 burn-induced CF ACD Hot leg LOCA, no ECCS, no CS Moderate S3B1 SBO, delayed RCP seal failure High TD-AFW likely fails upon battery (4), turbine-driven AFW depletion.

S3HF RCP seal LOCA, no ECCS, no CS High Transient causes loss of pump seal Recirculation cooling.

S3H RCP seal LOCA, no ECCS High Transient causes loss of pump seal recirculation cooling.

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5. RESULTS The analysis of accident progression and in-containment source terms is presented in this section.

The results are organized in a manner to support direct comparison with the format of the in-containment source term release phase magnitudes and durations presented in NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995). Phase durations and release fractions to containment for each radionuclide class are presented for each accident phase. Phase timings are calculated according to criteria described in Table 3-2. Release fractions have been calculated for a variety of accident sequences for four prototypic reactors described in Sections 4.2 and 4.3

• PWR: Large-dry containment (Surry) and Ice Condenser containment (Sequoyah)

• BWR: Mark I containment (Peach Bottom) and Mark III containment (Grand Gulf) for four initial radionuclide inventories presented in Section 2.3.3:

• LEU: 60 GWd/MTU and 80 GWd/MTU

• HALEU: 60 GWd/MTU and 80 GWd/MTU For consistency with the definition of in-containment source terms in RG 1.183 and previous in-containment source term analyses (NUREG-1465 and SAND2011-0128), the in-containment source term reported in this analysis include the entire radionuclide inventory within containment. In other words, results presented do not reflect radionuclide removal and retention mechanisms in containment (e.g., gravitational settling, suppression pool scrubbing, etc.).

Finally, release fractions of less than 1.0x10-6 are considered negligible for the purpose of this analysis and are truncated.

To produce the release fractions presented in this section, a non-parametric bootstrap analysis, described in Section 3.4, was employed. As described in Section 3.4, the non-parametric bootstrap analysis is used to develop a smoothed, mean ECDF. Example smoothed, mean ECDFs for BWRs and PWRs, obtained using the non-parametric bootstrap analysis, are shown in Figure 5-1. Results are shown for percentiles from 5 to 95, each point corresponds to a given early in-vessel phase duration and the percentage of simulations that exhibit phase durations less than or equal to that value. Reported results are taken as the 50th percentile to equally weight the set of simulations considered. Results presented in this section are intended to update the HBU source term recommendations presented in SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011).

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50th Percentile Figure 5-1 Non-parametric bootstrap of early in-vessel phase duration Overall in-containment source terms for BWRs and PWRs (accounting for all core types considered) with percent uncertainty are presented in 5-1 and 5-2. Table entries present the 50th percentile value of the smoothed, mean-ECDF calculated using the non-parametric bootstrap methodology for each QoI for each reactor type as well as the standard deviation at that percentile expressed as percent uncertainty.12 Limited uncertainty in the in-containment source terms for both reactor types indicate only slight variations in the in-containment source terms between cores with increased burnup and extended enrichment.

• Evolution of severe accident state-of-knowledge. Since the drafting of NUREG-1465, several insights into severe accident phenomena and progression have been identified. In particular, advancements have been made in understanding radionuclide chemistry and transport under severe accident conditions. Such advancements are discussed in the text as they become pertinent.

• Evolution of severe accident state-of-practice. Numerous improvements to the MELCOR code and modeling practices have taken place since NUREG-1465 was published. In particular, heterogeneous modeling of the 2D MELCOR representation of the reactor core and improved fuel failure modeling practices have been developed.

• Semi-volatile and non-volatile radionuclide releases during the gap release phase.

Heterogeneous modeling of the reactor core in MELCOR allows overlap between gap release 12 Medians (50 percentiles) of the distributions are considered here to be the best estimates of the uncertain quantities because it avoids any weighting of specific simulation.

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and early in-vessel phase phenomena during each phase. That overlap permits quantifiable semi-volatile and non-volatile radionuclide releases to occur during the gap release phase.

• Largest fractional releases during the early in-vessel phase. The majority of fractional radionuclide releases to containment occur during the early in-vessel phase for both reactor types. This follows from the fact that the majority of core degradation and liberation of the fission products from both the fuel matrix and cladding fission product barriers occurs during the early in-vessel phase. Volatile and semi-volatile fission products are the primary contributors to containment releases during the early in-vessel phase. This follows from the behavior of primary depressurization mechanisms for both reactor types. A larger contribution by semi-volatile and some non-volatiles is observed for PWRs, which exhibited higher incidences of primary pressure boundary breaches that allowed liquid coolant to escape into containment; PWR simulations generally predicted hot leg creep rupture while BWR simulations generally predicted thermal SRV seizure. Low pressure scenarios lead to larger releases to containment.

• Restricted radionuclide release during the late in-vessel and ex-vessel phases.

Radionuclide release during the late in-vessel and ex-vessel phases are limited. Two notable exceptions are observed in the halogen and Te group releases during the late in-vessel phase for BWRs and are highlighted in Table 5-1. Both releases are addressed in greater detail in the following sections.

Table 5-1 Recommended BWR phase durations (in hours) and release fractions with uncertainties for all core variations (60 GWd/MTU, 80 GWd/MTU, LEU and HALEU).

Late in-vessel releases for halogen and Te group radionuclide releases are large relative to other radionuclide groups during that accident phase.

Gap Release Early In-vessel Late In-vessel Ex-vessel Phase Duration [h] 0.70 (5%) 6.7 (3%) 44.6 (3%) 3.1 (24%)

Noble Gases 0.016 (26%) 0.95 (1%) 0.005 (23%) 0.011 (17%)

Halogens 0.005 (69%) 0.71 (2%) 0.16 (12%) 0.017 (13%)

Alkali Metals 0.005 (73%) 0.32 (10%) 0.021 (11%) 0.009 (19%)

Te Group 0.003 (85%) 0.56 (7%) 0.19 (12%) 0.003 (9%)

Ba/Sr Group 0.0006 (81%) 0.005 (6%) 0.002 (7%) 0.038 (20%)

Ru Group <1.0E-6 0.006 (2%) 7.9E-05 (30%) <1.0E-6 Mo Group 1.9E-05 (160%) 0.12 (8%) 0.002 (18%) 2.3E-05 (136%)

Lanthanides <1.0E-6 <1.0E-6 <1.0E-6 3.6E-05 (24%)

Ce Group <1.0E-6 <1.0E-6 0.0 (0.0%) 0.003 (51%)

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Table 5-2 Recommended PWR phase durations (in hours) and release fractions with uncertainties for all core variations (60 GWd/MTU, 80 GWd/MTU, LEU and HALEU).

Gap Release Early In-vessel Late In-vessel Ex-vessel Phase Duration [h] 1.3 (3%) 4.0 (6%) 24.0 (42%) 1.9 (13%)

Noble Gases 0.026 (6%) 0.93 (1%) 0.010 (24%) 0.018 (12%)

Halogens 0.007 (15%) 0.58 (12%) 0.031 (53%) 0.020 (10%)

Alkali Metals 0.003 (43%) 0.50 (25%) 0.013 (15%) 0.015 (12%)

Te Group 0.006 (9%) 0.55 (14%) 0.019 (37%) 0.005 (8%)

Ba/Sr Group 0.001 (10%) 0.002 (8%) 0.0001 (83%) 0.011 (20%)

Ru Group <1.0E-6 0.008 (19%) 5.4E-05 (32%) <1.0E-6 Mo Group 2.0E-05 (81%) 0.15 (23%) 0.002 (23%) 0.002 (94%)

Lanthanides <1.0E-6 <1.0E-6 <1.0E-6 1.4E-05 (25%)

Ce Group <1.0E-6 <1.0E-6 <1.0E-6 0.0006 (47%)

5.1. High Burnup, High Assay Low-Enriched Uranium Fuel A comparison of phase durations and in-containment source terms (fractional releases) for each core type is presented in 5-3 and 5-4 for BWRs and PWRs, respectively. Table entries present the 50th percentile value of the smoothed, mean-ECDF calculated using the non-parametric bootstrap methodology for each QoI for each core type. Across the set of core types, little variation is observed in the in-containment source terms in terms of release fractions with increased burnup and extended enrichment. This result is unsurprising when considering the small variations in decay heat observed with increased burnup and extended enrichment observed in [ORNL PWR and BWR references]. In fact, the most significant variations in the in-containment source term come from accident scenario differences. Changes to radionuclide inventories as a result of increased burnup and enrichment, however, may impact subsequent dose calculations and feasibility of HBU/HALEU implementation.

• Accident sequence differences dominate variation in source term to containment. The agreement between results for the different fuel burnup and extended enrichments (i.e. core types) strongly indicates that in-containment source terms are less sensitive to core type variation than accident initiators and the sequence of events that follow. In other words, variation in the in-containment source terms is larger for different accident sequences than for different core types.

• Similar accident phase durations. For BWRs, results show almost no discernable differences between core types in phase duration for the gap release and early in-vessel phases. Both late in-vessel and ex-vessel phase durations exhibit larger variations between 2 h and 5 h. For PWRs, accident phase durations exhibit minimal variation for the gap release, early in-vessel, and also ex-vessel phases. Greater variation is observed for the late in-vessel phase across core types.

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• Similar release fraction distributions. Distribution of BWR release fractions, across the radionuclide classes, is essentially unchanged for each core type despite differences in phase durations. The largest observed difference is ~10% during the early in-vessel phase for the Te group radionuclides; a discrepancy believed to arise from prolonged peripheral core structure lifetimes in the 80 GWd/MTU core case, which is supported by a subsequently larger Te group release during the late in-vessel phase. As in the case of BWRs, the distribution of release fractions is similar for PWRs regardless of core type. The largest discrepancy of 2.1%

occurs for the halogen radionuclides during the late in-vessel phase.

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Table 5-3 Comparison of recommended BWR phase durations (in hours) and release fractions for all core types. (1) - 60 GWd/MTU LEU, (2) - 80 GWd/MTU LEU, (3) - 60 GWd/MTU HALEU, and (4) - 80 GWd/MTU HALEU.

Gap Release Early In-vessel Late In-vessel Ex-vessel Core (1) (2) (3) (4) (1) (2) (3) (4) (1) (2) (3) (4) (1) (2) (3) (4)

Type Phase Duration

[h] 0.67 0.64 0.67 0.65 6.7 6.3 6.5 6.3 44.8 46.7 41 42.4 2.9 3.6 3 3 Noble Gases 0.016 0.016 0.017 0.016 0.94 0.96 0.94 0.94 0.007 0.004 0.007 0.004 0.014 0.006 0.013 0.01 Halogens 0.008 0.006 0.012 0.007 0.71 0.71 0.76 0.71 0.15 0.17 0.12 0.14 0.02 0.012 0.02 0.017 Alkali Metals 0.008 0.006 0.011 0.006 0.31 0.31 0.31 0.26 0.017 0.026 0.019 0.024 0.013 0.005 0.011 0.009 Te Group 0.006 0.005 0.009 0.005 0.54 0.59 0.56 0.49 0.21 0.17 0.19 0.16 0.004 0.003 0.004 0.003 Ba/Sr Group 0.001 0.0009 0.002 0.001 0.005 0.005 0.006 0.005 0.003 0.002 0.002 0.002 0.038 0.047 0.033 0.032 Ru 4.60E- 6.80E-Group <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 0.006 0.006 0.005 0.006 05 05 0.0001 0.0001 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 Mo 7.90E- 2.90E- 2.70E- 9.50E- 5.40E- 2.40E- 6.40E-Group 05 0.0001 05 05 0.12 0.12 0.11 0.12 0.002 0.004 0.002 0.003 06 05 05 05 Lanthani 3.50E- 5.40E- 2.70E- 3.60E-des <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 05 05 05 05 Ce Group <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 0 0 0 0 0.003 0.005 0.002 0.003 98

Table 5-4 Comparison of recommended PWR phase durations (in hours) and release fractions for all core types. (1) - 60 GWd/MTU LEU, (2) - 80 GWd/MTU LEU, (3) - 60 GWd/MTU HALEU, and (4) - 80 GWd/MTU HALEU.

Gap Release Early In-vessel Late In-vessel Ex-vessel Core Type (1) (2) (3) (4) (1) (2) (3) (4) (1) (2) (3) (4) (1) (2) (3) (4)

Phase Duration

[h] 1.3 1.3 1.3 1.3 4 3.8 4.2 3.8 27 21.1 14.1 28.2 2.2 1.8 1.9 2.1 Noble Gases 0.024 0.024 0.027 0.027 0.93 0.92 0.91 0.92 0.012 0.011 0.009 0.011 0.014 0.016 0.018 0.018 Halogens 0.007 0.006 0.008 0.007 0.57 0.56 0.57 0.58 0.05 0.039 0.026 0.025 0.018 0.019 0.02 0.021 Alkali Metals 0.003 0.002 0.003 0.003 0.5 0.5 0.5 0.51 0.018 0.014 0.014 0.01 0.012 0.014 0.016 0.016 Te Group 0.006 0.005 0.007 0.005 0.55 0.54 0.55 0.56 0.029 0.024 0.016 0.014 0.005 0.005 0.005 0.005 7.20E-Ba/Sr Group 0.001 0.0009 0.001 0.0009 0.002 0.002 0.002 0.002 0.0005 0.0008 0.0004 05 0.01 0.013 0.009 0.01 5.60E- 6.90E- 3.70E- 4.30E-Ru Group

<1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 0.007 0.008 0.008 0.008 05 05 05 05 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 1.60E- 2.10E- 3.90E- 4.60E-Mo Group 05 05 05 05 0.15 0.15 0.15 0.15 0.002 0.002 0.002 0.002 0.003 0.002 0.001 0.002 1.70E- 2.20E- 1.20E- 1.30E-Lanthanides

<1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 05 05 05 05 1.70E-Ce Group

<1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 05 <1.0e-6 <1.0e-6 0.0007 0.001 0.0005 0.0006 99

5.1.1. Gap Release Phase 5.1.1.1. Results As discussed in SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011), evaluating the end of a gap release phase is made ambiguous by the heterogeneity of core damage progression exhibited in modern, full-core MELCOR analyses (Powers, Leonard, Gauntt, Lee, &

Salay, 2011). The gap release phase, as observed in single rod/assembly experiments, captures the period of localized heatup in which the fuel cladding balloons and ruptures. At the time of rupture, radioactive vapors contained in the void between the fuel and cladding and other voids in the fuel pellet matrix, the gap inventory, are released from the fuel into the primary circuit.

Radionuclides considered in the gap inventory include noble gases (Kr and Xe) and volatile compounds such as Cs- and I-based compounds. The gap release phase ends (and early in-vessel phase begins) at high core temperatures, when rod-like geometry is lost allowing rapid diffusion of radionuclides previously contained in the fuel pellet matrix.

In this analysis the authors have employed the same gap release phase timing metrics as those employed in SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011). The gap release phase is defined as follows:

• Start of phase: core water level reaches the top of active fuel

• End of phase: 5% of Xe release from fuel By this approximation, some in-vessel accident progression as prescribed in NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) can take place.

In MELCOR the reactor core is discretized into axial levels and radial rings. This 2D representation allows local gap inventory release in individual axial levels and radial rings. For this type of discretization of the core, degradation tends to proceed in hotter, central core cells before either cooler, peripheral core cells experience severe enough conditions for fission product gap inventory to begin releasing or the gap inventory has been completely released. Thus full-core MELCOR calculations tend not to exhibit a distinct gap release phase of an accident because of the prolonged time over which fuel degradation commences across the core.

Radionuclide release fractions during the gap phase are shown in Figure 5-2. BWR release during the gap phase are lower than PWR releases for all radionuclide classes considered.

Releases for noble gases, halogens, and alkali metals are below NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) recommendations in all simulations. As discussed above, due to the criteria selected to signal the end of the gap release phase, early in-vessel accident phenomena do occur during the gap release phase in both PWR and BWR simulations.

The occurrence of early in-vessel phenomena is evidenced by releases of the Te, Ba/Sr, Ru, and Mo groups, which are considered in the gap inventory. The variance in BWR results is notably smaller than the variance observed in PWR results for nearly all radionuclide classes.

Figure 5-3 shows the outcomes of MELCOR simulation gap phase durations and radionuclide class release fractions prior to applying the non-parametric bootstrap analysis. Each point on the plot corresponds to a given MELCOR result (small circles overlaying the boxplots) or regulatory recommendation (large circles), while the boxplots provide insight into the distribution of MELCOR outcomes prior to statistical processing including the 25th, 50th, and 75th percentile 100

outcomes and distribution bounds/outliers. In Figure 5-3 gap release phases are generally shorter for BWRs than for PWRs. Gap release phase durations are, however, longer for both reactor types than recommended in NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995).

The results of a bootstrap analysis, discussed in Section 3.4, performed on the data shown in 5-2 and 5-3, and previous related values for comparison, are presented in Table 5-5 and Table 5-6 for BWRs and PWRs, respectively. The bootstrap analysis confirms that the gap release phase duration, and release fractions for each radionuclide are smaller for BWRs than PWRs except for alkali metals, which is marginally larger for BWRs. This result follows the median behavior for BWRs and PWRs shown in Figure 5-2.

5.1.1.2. Discussion Gap release phase release fractions for the noble gases, halogen, and alkali metals radionuclide classes are less than NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) recommendations, while Te, Ba/Sr, Ru, and Mo groups releases are larger than NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) recommendations. Results do not deviate significantly from those reported in SAND2011-0128 (Powers, Leonard, Gauntt, Lee, &

Salay, 2011) for either reactor type outside of phase duration, which is longer in the current study.

• Long gap release phase. Phase durations for the gap release phase are longer, relative to NUREG-1465 and SAND2011-1028, for both reactor types. Gap release phase criteria can only coarsely approximate the release of the gap inventory from fuel which does not occur over a distinct time frame in a heterogenous core. By its definition, the approximated gap release phase may overlap with the core heat-up/boiloff and early in-vessel phases, prolonging the observed gap release phase duration. Furthermore, modern MELCOR models the distribution of heat in the core and heat removal from the core more efficiently during the core boiloff transient than the STCP.

• Lower volatile releases. Gap release phase source terms to containment are lower in this analysis compared to NUREG-1465 because of a time delay between fission product release from fuel and fission product transport through the primary system into containment.

• Semi-volatile and non-volatile releases. Semi-volatile and non-volatile releases are observed during the gap release phase because the gap phase timing criteria must be approximated when modeling a heterogeneous reactor core. In approximating the gap release phase timings, the front end of early in-vessel accident progression is captured in some cells of the heterogeneous core model.

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Figure 5-2 MELCOR calculated fractional release to containment during the gap release phase.

Releases for noble gases, halogens, and alkali metals are all less than NUREG-1465 recommendations. Releases for Te group, Ba/Sr group, and Mo group are greater than NUREG-1465 recommendations.13 13 Diamonds represent outliers (beyond the quartiles) and circles show MELCOR results for each simulation in the data category.

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Figure 5-3 MELCOR calculated accident phase durations. Phase durations for the gap release, early in-vessel, and late in-vessel phases are generally greater than NUREG-1465 recommendations. Ex-vessel phase durations are consistent with or less than NUREG-1465 recommendations.14 14 Diamonds represent outliers (beyond the quartiles) and circles show MELCOR results for each simulation in the data category.

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Table 5-5 Comparison of Recommended BWR phase durations (in hours) and release fractions with those recommended in SAND2011-0128 and NUREG-1465. Darker shades - larger than recommendations from both previous studies. Lighter shades - larger than NUREG-1465 or SAND2011-0128 release recommendations.

Gap Release Early In-vessel Late In-vessel Ex-vessel 2021 2011 NUREG- 2021 2011 NUREG- 2021 2011 NUREG- 2021 2011 NUREG-Report 1465 1465 1465 1465 Phase Duration 0.70 0.16 0.50 6.7 8.0 1.5 44.6 12.0 10.0 3.1 2.9 3.0 Noble Gases 0.016 0.008 0.050 0.95 0.96 0.95 0.005 0.016 0.0 0.011 0.009 0.0 Halogens 0.005 0.002 0.050 0.71 0.47 0.25 0.16 0.39 0.010 0.017 0.013 0.30 Alkali Metals 0.005 0.002 0.050 0.32 0.13 0.20 0.021 0.050 0.010 0.009 0.010 0.35 Te Group 0.003 0.002 0.0 0.56 0.39 0.050 0.19 0.33 0.005 0.003 0.002 0.25 Ba/Sr Group 0.0006 0.0 0.0 0.005 0.005 0.020 0.002 0.005 0.0 0.038 0.029 0.10 Ru Group <1.0e-6 0.0 0.0 0.006 0.003 0.003 7.9E-05 0.0001 0.0 <1.0e-6 0.003 0.003 Mo Group 1.9E-05 0.0 0.0 0.12 0.020 0.003 0.002 0.005 0.0 2.3E-05 0.003 0.003 Lanthanides <1.0e-6 0.0 0.0 <1.0e-6 <1.0e-6 0.0002 <1.0e-6 0.0 0.0 3.6E-05 5.0E-05 0.005 Ce Group <1.0e-6 0.0 0.0 <1.0e-6 <1.0e-6 0.0005 0.0 0.0 0.0 0.003 0.002 0.005 104

Table 5-6 Comparison of recommended PWR phase durations (in hours) and release fractions with those recommended in SAND2011-0128 and NUREG-1465. Darker shades - larger than recommendations from both previous studies. Lighter shades - larger than NUREG-1465 or SAND2011-0128 release recommendations.

Gap Release Early In-vessel Late In-vessel Ex-vessel 2021 2011 NUREG- 2021 2011 NUREG- 2021 2011 NUREG- 2021 2011 NUREG-Report 1465 1465 1465 1465 Phase Duration [h] 1.3 0.22 0.50 4.0 4.5 1.3 24.0 143.0 10.0 1.9 4.8 2.0 Noble Gases 0.026 0.017 0.050 0.93 0.94 0.95 0.010 0.003 0.0 0.018 0.011 0.0 Halogens 0.007 0.004 0.050 0.58 0.37 0.35 0.031 0.21 0.10 0.020 0.011 0.25 Alkali Metals 0.003 0.003 0.050 0.50 0.23 0.25 0.013 0.060 0.10 0.015 0.020 0.35 Te Group 0.006 0.004 0.0 0.55 0.30 0.050 0.019 0.10 0.005 0.005 0.003 0.25 Ba/Sr Group 0.001 0.0006 0.0 0.002 0.004 0.020 0.0001 0.0 0.0 0.011 0.003 0.10 Ru Group <1.0e-6 0.0 0.0 0.008 0.006 0.003 5.4E-05 0.0 0.0 <1.0e-6 0.003 0.003 Mo Group 2.0E-05 0.0 0.0 0.15 0.080 0.003 0.002 0.030 0.0 0.002 0.010 0.003 Lanthanides <1.0e-6 0.0 0.0 <1.0e-6 <1.0e-6 0.0002 <1.0e-6 0.0 0.0 1.4E-05 1.3E-05 0.005 Ce Group <1.0e-6 0.0 0.0 <1.0e-6 <1.0e-6 0.0005 <1.0e-6 0.0 0.0 0.0006 0.0002 0.005 105

5.1.2. Early In-Vessel Phase 5.1.2.1. Results During the early in-vessel phase, the reactor core temperature continues to increase. Further elevated core temperatures cause the loss of core geometry and downward relocation of core materials in the reactor pressure vessel either as falling solid particulate debris, or molten flows.

Volatile radionuclides previously retained in the fuel matrix are liberated upon the loss of core geometry. At the same time, elevated core debris temperatures allow semi-volatile radionuclides to escape fuel debris. The early in-vessel phase is defined as follows:

• Start of phase: 5% of Xe release from fuel (end of gap release phase)

• End of phase: reactor pressure vessel failure Simulation results for the duration of the early in-vessel phase and release of each radionuclide class are shown in 5-3 and 5-4, respectively. All simulations exhibit longer early in-vessel phase durations than NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) recommendations. As discussed in SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011), the early in-vessel phase is longer in modern MELCOR studies because of heat transfer modeling improvements that predict more efficient heat removal from the reactor core, limiting the core heatup transient. Releases during the early in-vessel phase are not as simply characterized as they were for the gap phase between BWRs and PWRs or between this analysis and the recommendations from NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995).

The variance in BWR results and PWR results, respectively, is similar for the early in-vessel phase across all radionuclide classes as shown in Figure 5-4. BWR releases tend to cluster about slightly higher releases fractions than PWRs for noble gases, halogens, Te group, and Ba/Sr group radionuclides. The opposite is true for alkali metals, Ru group, and Mo Group radionuclides. Both PWRs and BWRs generally release the majority (about 95%) of the noble gas inventory during the early in-vessel phase, which is consistent with NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) recommendations. NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) recommendations fall inside of the variation of results for alkali metals, and Ru group releases. Ba/Sr group releases are notably less that NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) recommendations, in part due to releases characterized as occurring during the gap release phase.

Bootstrap analysis results shown in Table 5-5 and Table 5-6 confirms that the early in-vessel phase is longer for BWRs, and that BWR releases trend higher for all radionuclide classes except for the alkali metals, Ru, and Mo group radionuclides (Lanthanides and Ce group release are below the release fraction reporting threshold). Early in-vessel release fractions for the noble gases, Ba/Sr, Ru, Lanthanides, and Ce groups presented in Table 5-5 and Table 5-6 are consistent with or less than NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) and SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011). Releases for halogens, alkali metals, Te, and Mo groups, however, are all higher than both NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) and SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011) reported values. As discussed above, this is due to 106

advancements in the state-of-knowledge regarding the chemical behavior of fission products and the increased number of low-pressure scenarios predicted.

Figure 5-4 MELCOR calculated fractional release to containment during early in-vessel phase.

Releases for noble gases are comparable to NUREG-1465 recommendations.

Releases for Ba/Sr group, lanthanides, and Ce group radionuclides are less than NUREG-1465 recommendations. Releases for halogens, alkali metals, Te group, Ru group and Mo group radionuclides are generally greater than NUREG-1465 recommendations.15 5.1.2.2. Discussion Nearly all simulations for this analysis exhibit larger releases for the alkali metals, halogen, Te, Ru, and Mo groups than recommended by NUREG-1465 (Soffer, Burson, Ferrell, Lee, &

Ridgely, February 1995) for both PWRs and BWRs. There are several reasons for this behavior.

• Long early in-vessel phase. As observed for the gap release phase, modern MELCOR models the distribution of heat in the core and heat removal from the core more efficiently during the core boiloff transient than the STCP. Larger releases during the early in-vessel phase are observed, in part, because of the longer early in-vessel phase duration, which allows for more complete core degradation and release of radioactive compounds in both reactor types. In other words, releases that would have been characterized as late in-vessel or 15 Diamonds represent outliers (beyond the quartiles) and circles show MELCOR results for each simulation in the data category.

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ex-vessel releases in NUREG-1465 (Powers, Leonard, Gauntt, Lee, & Salay, 2011) are included in this phase.

• Occurrence of Safety Relief Valve (SRV) seizure for BWRs. Unlike many of the SBO sequences considered in NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995), the SBO sequences considered in SAND2011-0128 (Powers, Leonard, Gauntt, Lee, &

Salay, 2011) and this present study for BWRs were predominately ones with a thermally induced seizure of the SRV prior to lower head breach. This resulted in the vessel achieving a low-pressure condition in the majority of the SBO scenarios considered in the SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011) and present studies. This has a notable effect on the ability of fission products generated during the early in-vessel phase to be released from the vessel into containment. Across the scenarios for which SRV seizure resulted in the vessel depressurizing prior to lower head breach, the early in-vessel CsI releases exhibited relatively higher releases to containment from the time the gap release phase ended to the time of lower head failure. By contrast, the much smaller set of SBO scenarios in which the vessel remained at high-pressure prior to lower head breach exhibited lower CsI releases over the time from cessation of the gap phase to lower head breachthe magnitude of CsI release for a condition with the vessel remaining at high-pressure was consistent with that reported for halogen releases during the early in-vessel phase of the NUREG-1465 study (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995).

• Hot Leg Creep Rupture (HLCR) occurrence for PWRs. Releases of the halogen, alkali metals, Te, and Mo group radionuclide classes are notably larger for PWRs compared to NUREG-1465 recommendations because of a high incidence of hot leg creep rupture events during the early in-vessel phase. HLCR causes rapid depressurization of the primary system and low-pressure conditions prior to vessel breach. Both vaporized and unvaporized fission products travelling through the primary circuit are released to containment through HLCR sites increasing in-containment source terms during the early in-vessel phase. In the Surry SOARCA UA, HLCR was also observed in the majority of cases. The subset of simulations that do not exhibit HLCR and remain at high-pressure until vessel failure exhibit lower in-containment source terms for halogen, alkali metal, Te, and Mo group radionuclides as well as the other, non-volatile radionuclide classes - the magnitude of release among these classes is consistent with the NUREG-1465 recommendations except where the state-of-knowledge has advanced (i.e., Te and Mo group releases are in-line with the current state-of-knowledge).

• Enhancement in radiochemistry state-of-knowledge. The state-of-knowledge of radiochemistry impacting radionuclide releases to containment has also advanced since NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995). This analysis assumes negligible gaseous halogen inventories and the formation of fission product compounds not considered during the original drafting of NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995). For example, 100% of the I inventory is assumed to combine with Cs to form CsI; any remaining Cs mass is assumed to form Cs2MoO4. This is consistent with the findings of the Phébus-FP program (Clement & Haste, 2004). Thus, the dominant forms of I and Mo releases are associated with chemical compounds that include Cs. Larger Te releases are also consistent with the findings of the Phébus-FP program (Clement & Haste, 2004), which observed more efficient transport of Te group materials to containment than NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) 108

recommendations suggested. Lanthanide and Ce group radionuclide classes remain below the threshold (1.0x10-6) for all simulations performed in this analysis.

5.1.3. Late In-Vessel Phase 5.1.3.1. Results The late in-vessel phase, as it is defined, occurs in parallel with the ex-vessel phase and encompasses the release of radionuclides from the primary circuit after lower head failure.

Radionuclides remaining in the primary circuit after lower head failure include inventories retained in intact fuel and fuel debris that has not yet relocated ex-vessel as well as any radionuclides that were deposited in the primary circuit during the gap release and early in-vessel phases. Thus, radionuclide release phenomena during the late in-vessel phase include release of radionuclides from remaining intact fuel as it degrades and revaporizes radionuclides deposited in the RPV or primary coolant system before escaping to the containment. The late in-vessel phase is defined as follows:

• Start of phase: reactor pressure vessel failure (end of early in-vessel phase)

• End of phase: 95% the of total late in-vessel Cs releases have occurred Late in-vessel phase durations in this analysis are longer than NUREG-1465 recommendations, as shown in Figure 5-3. Longer phase durations observed in this analysis are a result of prolonged period of release of a small amount of cesium after RPV failure, the phase criterion metric. Table 5-5 and Table 5-6 show the late in-vessel phase results of the bootstrap analysis, which confirms that the phase duration for both PWRs and BWRs exceed NUREG-1465 recommendations (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995). And comparison of phase durations indicates that the late in-vessel phase is longer for BWRs than PWRs. The late in-vessel phase duration for BWRs in this analysis is also longer than the results of the SAND2011-0128 study (Powers, Leonard, Gauntt, Lee, & Salay, 2011), while the opposite is true for PWRs. It should be noted that simulations were only run for 3 days (72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />) in this analysis.

In Figure 5-5, late in-vessel release fractions are shown for both reactor types. In general, late in-vessel release fractions are larger for BWRs for the halogen, Te, Ba/Sr, and Ru groups, while they greater agreement is observed for the noble gases, alkali metals, and Mo groups - the Lanthanides and Ce group releases remain negligibly small for both reactor types. As shown in Table 5-5, compared to NUREG-1465 recommendations, the current study shows show larger releases for halogen, alkali metal, and Te groups in BWRs. Meanwhile in Table 5-6, PWRs show lower releases for halogens and alkali metals and larger releases for the Te group in comparison to NUREG-1465 recommendations (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995).

Both BWRs and PWRs show larger release than NUREG-1465 recommendations (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) of zero for noble gases, Ba/Sr, Ru, Mo, lanthanides, and Ce groups. However, apart from the Mo group, the release fractions determined in this analysis for these groups are relatively small. Noble gases, Ba/Sr, Ru, lanthanides, and Ce groups for both reactor types exhibit release fractions of about 0.5% or less of their initial total core inventory during the late in-vessel phase.

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Figure 5-5 MELCOR calculated fractional release to containment during late in-vessel phase.

Releases for noble gases, Te group, Ba/Sr Group, Ru group, and Mo group radionuclides are generally greater than NUREG-1465 recommendations. Releases for halogens and alkali metals are generally greater than NUREG-1465 recommendations for BWRs and less than NUREG-1465 recommendations for PWRs.16 5.1.3.2. Discussion Some differences between the results of this analysis, specific to the late in-vessel phase, and the results reported in NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) have been observed and are discussed in this section.

• Longer late in-vessel phase duration. By the definition of the late in-vessel phase described in SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011) and employed here, the late in-vessel phase duration is longer for both BWRs and PWRs relative to NUREG-1465.

Compared to SAND2011-0128, BWR late in-vessel phase durations are larger, while PWR phase durations are shorter.

• Continued core degradation. Portions of the reactor core can remain intact beyond vessel failure, particularly peripheral core structures including fuel assemblies. Persistent, partially intact fuel rods retain a proportion of the total radionuclide inventory that is ultimately 16 Diamonds represent outliers (beyond the quartiles) and circles show MELCOR results for each simulation in the data category.

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released during the late in-vessel phase upon failure of those fuel rods failure. These late fuel failures result in containment releases for all radionuclide classes during the late in-vessel phase.

5.1.4. Ex-Vessel Phase 5.1.4.1. Results The ex-vessel phase encompasses radionuclide releases from ex-vessel core debris. Hot debris ejected from the reactor pressure vessel after its failure accumulates on the floor of the reactor cavity where it interacts with and ablates the concrete floor. The corium concrete interaction liberates water trapped in the concrete further oxidizing the corium mixture. In the case of further zirconium oxidation, this oxidation can lead to a temperature increase transient that releases the remaining volatile and semi-volatile radionuclides. Furthermore, after the preferential oxidation of structural metals (i.e., Zr and Fe), volatile Mo oxides will begin to form increasing radioactive Mo releases. The ex-vessel phase is defined as follows:

• Start of phase: reactor pressure vessel failure (end of early in-vessel phase)

• End of phase: 95% of total ex-vessel Cs releases have occurred Contrary to the gap release and early in-vessel phases, ex-vessel phase durations shown in Figure 5-3 are in good agreement between this analysis and the NUREG-1465 recommendations (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995). BWR simulations exhibit longer ex-vessel phase durations than PWR simulations. Release fractions calculated for the ex-vessel phase in this analysis are lower than those recommended in NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) in almost all cases as can be seen in Figure 5-6. This is because of the larger releases observed during the gap release and early in-vessel phase, which reduces the available inventory of radionuclides to release during the ex-vessel and late in-vessel accident phases. The primary exception being Ce group releases, which are consistent with NUREG-1465 recommendations. A small amount of remaining noble gas inventory, nonetheless larger than the NUREG-1465 recommendation (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) of 0.0, is released during the ex-vessel phase in nearly all simulations. Calculated release fractions for halogens, alkali metals, Te, Ba/Sr, Ru, Mo groups, and Lanthanides are all lower than NUREG-1465 recommendations (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) in most cases.

The bootstrap analysis results in Table 5-5 and Table 5-6 confirm the same findings. Namely, that both PWR and BWR releases are consistent with or less than the recommendations of NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) and results presented in SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011).

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Figure 5-6 MELCOR calculated fractional release to containment during ex-vessel phase.

Releases for noble gases are greater than NUREG-1465 recommendations. Releases for halogens, alkali metals, Te group, Ba/Sr Group, Ru group, Mo group, lanthanides, and Ce group radionuclides are generally less than NUREG-1465 recommendations.

17 5.1.4.2. Discussion Differences between the results of this analysis, specific to the ex-vessel phase, and the results reported in NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) have been observed and are discussed in this section.

• Reduced ex-vessel releases. The contribution of the ex-vessel release phase to the overall in-containment source term falls between NUREG-1465 and SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011) results. Large releases during the earlier, prolonged accident phases depleted most of the volatile and semi-volatile fission products from core debris.

5.2. Additional Sensitivity Analyses Additional sensitivity analyses have been conducted to address issues specific to HBU/HALEU fuel that were identified during the ATF Severe Accident PIRT. The overall approach and 17 Diamonds represent outliers (beyond the quartiles) and circles show MELCOR results for each simulation in the data category.

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motivation for these sensitivity analyses are discussed in more detail in Section 3.3. The following discussion provides the results of MELCOR sensitivity studies targeted to address the specific issues identified by the ATF Severe Accident PIRT. These studies assess how uncertainty in specific HBU/HALEU fuel design parameters and performance could affect the in-containment source term, but do not constitute a characterization of generic severe accident uncertainties. The following sensitivity studies are presented below:

3. Alteration of thermophysical properties of the fuel and cladding at HBU - at increased burnup, the fuel thermal conductivity will decrease. Since the burnup dependence of fuel thermal conductivity is typically not represented in MELCOR simulations, this additional sensitivity is used to identify the impact of this assumption on in-containment source terms given the relatively HBUs under consideration (i.e., as high as 80 GWd/MTU).

o Fuel thermal conductivity - A sensitivity calculation is performed at lower fuel thermal conductivity:

• Reference fuel thermal conductivity - 4.92 W/m-k

• Reduced fuel thermal conductivity - 2.02 W/m-K

4. Fuel fragmentation and sintering at HBU - at increased burnup, fuel fragmentation and sintering is postulated. Increased fragmentation and sintering of fuel rods could result in earlier fuel degradation and failure. Various sensitivities are considered to evaluate the impacts of modified fuel degradation and debris behaviors that might occur as a result of increased fuel fragmentation and sintering.

o Porosity of in-vessel particulate debris - Sensitivity calculations are performed with varied in-vessel particulate debris porosities:

• Reference in-vessel particulate debris porosity - 0.4.

• Increased in-vessel particulate debris porosity - 0.6.

• Reduced in-vessel particulate debris porosity - 0.2.

o Diameter of in-vessel particulate debris - Sensitivity calculations are performed with varied in-vessel particulate debris diameters:

• Reference in-vessel particulate debris diameter -core region diameter: 0.01 m, lower plenum region diameter 0.002 m.

• Increased in-vessel particulate debris diameter - core region diameter: 0.015 m, lower plenum region diameter 0.005 m.

• Reduced in-vessel particulate debris diameter - core region diameter: 0.005 m, lower plenum region diameter 0.001 m.

o Particulate debris falling velocity - A sensitivity calculation is performed at lower particulate debris falling velocity:

• Reference in-vessel particulate debris falling velocity - Peach Bottom: 0.94 m/s, Surry: 0.64 m/s.

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• Reduced in-vessel particulate debris falling velocity - Peach Bottom: 0.094 m/s, Surry: 0.064 m/s.

o Fuel relocation temperature - Sensitivity calculations are performed with varied fuel relocation temperatures. A sensitivity study was also performed with the eutectics model activated.

• Reference fuel relocation temperature - 2479.0 K.

• Increased fuel relocation temperature - 2728.0 K.

• Reduced fuel relocation temperature - 2230.0 K18.

• Eutectics model - default eutectics model activated (interactive materials model deactivated).

o Fuel rod lifetime - Sensitivity calculations are performed with varied fuel rod lifetimes.

• Reference fuel rod lifetime - default time-at-temperature model activated.

• Increased fuel rod lifetime - Lifetime function that accrues damage from 22.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 20.0 minutes at temperatures from 2100.0 K - 2600.0 K.

• Reduced fuel rod lifetime - Lifetime function that accrues damage from 1.67 hours7.75463e-4 days <br />0.0186 hours <br />1.107804e-4 weeks <br />2.54935e-5 months <br /> to 3.3 minutes at temperatures from 2100.0 K - 2600.0 K.

• SOARCA fuel rod lifetime - Lifetime function that accrues damage from 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> to 5 minutes at temperatures from 2100.0 K - 2600.0 K.

Hot leg creep rupture - An additional sensitivity, not identified as part of the ATF Severe Accident PIRT, is performed to investigate in-containment source term sensitivity to early primary pressure boundary failure by HLCR. HLCR is suppressed in this sensitivity calculation.

5.2.1. Fuel Thermal Conductivity To evaluate the effect of decreasing fuel thermal conductivity with increasing burnup, a sensitivity calculation was performed in which the thermal conductivity of fuel material in the MELCOR model was adjusted using the temperature-dependent correlation for fuel thermal conductivity implemented in the FAST computer code (Geelhood, Kyriazidis, Corson, Luscher,

& Goodson, 2021). To illustrate the effect of burnup on fuel thermal conductivity, consider that the typical value for fuel thermal conductivity is 4.92 W/m-K at 757 K according to the base correlation implemented in MELCOR. According to the FAST correlation, at 80 GWd/MTU, the fuel thermal conductivity is reduced to 2.02 W/m-K at 757K.

However, a full MELCOR analysis indicates that the fuel thermal conductivity does not fundamentally alter the transient response of fuel material throughout the course of an accident in either BWR or PWRs. This is demonstrated by the behavior of peak fuel temperatures throughout the course of the accident simulation, shown in Figure 5-7 for Peach Bottom and Surry, respectively. The degradation of cooling that defines a short-term SBO, dominates the effect of thermal conductivitythat is, without sufficient cooling the decay energy is primarily 18 To maintain stylized progression of degradation phenomena, the zircaloy melt breakout temperature was also reduced from 2350.0 K to 2100.0 K in this simulation.

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rejected as stored energy in the fuel. The rejection of decay energy into stored energy is not governed by thermal conductivitythe fuel temperature transient due to decay heat converted into stored energy is primarily influenced by the specific heat of the fuel material. Since the specific heat has not been found to exhibit significant dependence on burnup, there is limited influence on the fuel temperature transient through the progression of a severe accident driven primarily by a loss of decay heat removal.

Figure 5-7 Peak Fuel Temperature Transient in Peach Bottom Short-Term (left) and Surry (right)

SBO reference simulation and reduced fuel conductivity simulation.

Given that fuel temperature is a strong signature of the progression of core damage and is critical for establishing the conditions for release of fission products from fuel, the largely similar temperature transients for each plant type shown in Figure 5-7 indicate that there is likely a limited effect of fuel thermal conductivity on fission product release to containment. Comparison of fission product release transients to containment, as shown in Figure 5-8 for Peach Bottom and Figure 5-9 for Surry, confirm that there is no appreciable impact, in excess of the uncertainties considered in Section 5.1, of fuel thermal conductivity on accident progression and in-containment source term. Thus, variation of fuel thermal conductivity with burnup does not impact the key results of this source term study.

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Figure 5-8 Fission Product Release to Containment for Peach Bottom Short-Term SBO Fuel Thermal Conductivity Sensitivity 116

Figure 5-9 Fission Product Release to Containment for Surry SBO Fuel Thermal Conductivity Sensitivity 5.2.2. Porosity of In-Vessel Particulate Debris The significant disruption to the fuel that occurs during operation through to very high burnups has been postulated to lead to degradation of fuel in a manner that promotes much greater disintegration of material. One means of representing this effect in MELCOR is to adjust the porosity of in-vessel particulate debris beds. The reference MELCOR analyses assume a porosity of 0.40. Two sensitivities, utilizing both a Peach Bottom short-term SBO scenario and a Surry SBO scenario, were conducted to assess the impact of in-vessel particulate debris bed porosity.

• Sensitivity #1, low debris porosity: assumed particulate debris porosity of 0.20

• Sensitivity #2, high debris porosity: assumed particulate debris porosity of 0.60 117

Between the two sensitivities, there is generally a close agreement in the overall accident progression for both plant types as shown in Table 5-7. With respect to lower plenum dryout, the largest discrepancy in time is approximately 0.70 hours8.101852e-4 days <br />0.0194 hours <br />1.157407e-4 weeks <br />2.6635e-5 months <br />. The sensitivity to particulate debris bed porosity was somewhat greater in the case of lower head failure time, with a maximum discrepancy of 1.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Table 5-7 In-vessel debris bed porosity sensitivity accident timings Peach Bottom [h] Surry [h]

Lower Plenum Dryout Lower Head Failure Lower Plenum Dryout Lower Head Failure Reference Case 3.75 6.75 5.10 6.50 Low debris porosity 3.95 6.70 5.30 6.55 High debris porosity 4.45 7.95 5.58 7.18 This variation is overall inconsequential given the greater variation in lower head failure times observed when considering only aleatory uncertainty due to variability in the realized accident scenario with the results presented in Section 5.1.

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Figure 5-10 Fission Product Release to Containment for Peach Bottom Short-Term SBO In-Vessel Debris Bed Porosity Sensitivity 119

Figure 5-11 Fission Product Release to Containment for Surry SBO In-Vessel Debris Bed Porosity Sensitivity Comparison between Peach Bottom sensitivity simulations shown in Figure 5-10 indicates that in-vessel debris porosity has a limited effect on containment fission product releases throughout the first 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of the accident scenario being simulated. The same observation is made for the Surry sensitivity simulations, which are shown in Figure 5-11. The variability introduced by in-vessel particulate debris bed porosity on the times of key accident scenario events and the magnitude of fission product release into containment is not appreciable compared to the variability observed across accident scenarios considered in Section 5.1. Thus, variation of porosity of in-vessel particulate debris does not impact the key results of this source term study relative to other considered uncertainties.

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5.2.3. Diameter of In-Vessel Particulate Debris The diameter of in-vessel particulate debris is assumed in the reference MELCOR analysis to be

• 10 mm particle diameters for debris that has formed inside the core region

• 2 mm particle diameters for debris that has accumulated in the lower plenum given the likely disintegration of debris that occurs when slumping into a lower plenum water pool To assess the impact of a various degrees of fuel breakup that could occur when fuel at higher burnups degrades (i.e., forms a rubble bed), additional sensitivities were conducted assuming

• Sensitivity #1, low particle diameter: core region diameter: 0.005 m, lower plenum region diameter 0.001 m

• Sensitivity #2, high particle diameter: core region diameter: 0.015 m, lower plenum region diameter 0.005 m The impact of this variation on key event times for the Peach Bottom short-term SBO simulated are shown in Table 5-8. Overall, the simulations exhibit less sensitivity to particle diameter than particulate debris porosity. As such, the variations in key event timings introduced by changes to the particle diameters are not appreciable relative to the variation observed across accident scenarios.

Table 5-8 In-vessel particle diameter sensitivity accident timings Peach Bottom [h] Surry [h]

Lower Plenum Dryout Lower Head Failure Lower Plenum Dryout Lower Head Failure Reference Case 3.75 6.75 5.10 6.50 Low particle diameter 3.75 6.75 5.23 6.77 High particle diameter 3.85 7.15 5.54 6.67 The impact on in-containment source term is similarly small as in the evaluation of the sensitivity to particulate debris bed porosity. This is illustrated in Figure 5-12 for Peach Bottom and Figure 5-13 for Surry. Thus, variation of particle diameters does not impact the key results of this source term study relative to other considered uncertainties.

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Figure 5-12 Fission Product Release to Containment for Peach Bottom Short-Term SBO In-Vessel Particle Diameter Sensitivity 122

Figure 5-13 Fission Product Release to Containment for Surry SBO In-Vessel Particle Diameter Sensitivity 5.2.4. Particulate Debris Falling Velocity Varying particulate debris sizes that may arise in HBU cores under severe accident conditions reasonably impact the falling velocity of particulate debris in the lower plenum. In the Peach Bottom reference simulation, the falling velocity of particulate debris in the lower plenum is 0.94 m/s. For Surry, the reference simulation used a falling velocity of 0.64 m/s. A sensitivity calculation was performed for each plant to investigate the effect of reduced falling velocities of particulate debris in the lower plenum. In both sensitivity calculations, the falling velocity of particulate debris in the lower plenum is reduced by a factor of 10 (Peach Bottom: 0.094 m/s, Surry: 0.064 m/s). The impact of reducing the falling velocity of particulate debris in the lower plenum on in-containment source terms in Peach Bottom is shown in Figure 5-14 and was found 123

to be negligible on both core damage progression and in-containment source terms. For the Surry sensitivity calculation, the impact of reducing the falling velocity of particulate debris in the lower plenum on in-containment source terms is larger as shown in Figure 5-15. However, it is still found that the falling velocity of particulate debris in the lower plenum does not impact the key results of this source term study relative to other considered uncertainties.

Figure 5-14 Fission Product Release to Containment for Peach Bottom Short-Term SBO Particle Debris Falling Velocity Sensitivity 124

Figure 5-15 Fission Product Release to Containment for Surry SBO Particle Debris Falling Velocity Sensitivity 5.2.5. Fuel Rod Lifetime Physio-chemical processes that occur in fuel rods at high temperatures are known to cause fuel rod failure at temperatures below the melting point of constituent materials. These physio-chemical processes are represented in MELCOR by a lifetime function, by which fuel rods accrue damage at a given temperature-dependent rate until the fuel rod has sustained sufficient damage to cause its failure. The fuel rod lifetime model approximates fuel rod collapse as a result of the integral accumulation of damage that occurs when fuel rods experience extended periods of high temperature; the model imposes a maximum length of time (lifetime) that a fuel component may remain intact based on its unique temperature history. According to MELCOR best practices established during the SOARCA project, fuel rod damage due to high temperature 125

effects is considered insignificant below 2090 K. The impact of HBU fuels on the speed at which damage is accrued in fuel rods is investigated by using four representative fuel rod lifetimes.

• Reference case: default fuel rod lifetime - MELCOR default time-at-temperature model activated.

• Sensitivity #1: reduced fuel rod lifetime - Lifetime function that accrues damage predicting fuel failure from 1.67 hours7.75463e-4 days <br />0.0186 hours <br />1.107804e-4 weeks <br />2.54935e-5 months <br /> to 3.3 minutes at temperatures from 2100.0 K - 2600.0 K as shown in Table 5-9.

Table 5-9 Reduced fuel rod lifetime function Temperature [K] Lifetime [s]

< 2090.0 Infinite 2100.0 6000.0 2500.0 600.0 2600.0 200.0

• Sensitivity #2: increased fuel rod lifetime - Lifetime function that accrues damage predicting fuel failure from 22.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to 20.0 minutes at temperatures from 2100.0 K to 2600.0 K, shown in Table 5-10.

Table 5-10 Increased fuel rod lifetime function Temperature [K] Lifetime [s]

< 2090.0 Infinite 2100.0 80000.0 2500.0 2000.0 2600.0 1200.0

• Sensitivity #3: SOARCA fuel rod lifetime - Lifetime function that accrues damage predicting fuel failure from 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> to 5 minutes at temperatures from 2100.0 K - 2600.0 K, shown in Table 5-11.

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Table 5-11 SOARCA fuel rod lifetime function Temperature [K] Lifetime [s]

< 2090.0 Infinite 2100.0 36000.0 2500.0 3600.0 2600.0 300.0 Table 5-12 shows the impact of fuel rod lifetime variation on key event times, which are generally minimal for both plants. The largest variation is observed in lower head failure timing, which occurs most rapidly (~0.1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> earlier) with reduced fuel rod lifetimes.

Table 5-12 Fuel rod lifetime sensitivity accident timings Peach Bottom [h] Surry [h]

Lower Plenum Dryout Lower Head Failure Lower Plenum Dryout Lower Head Failure Reference Case 3.75 6.75 5.10 6.50 10th percentile lifetime 3.75 6.65 5.10 6.56 90th percentile lifetime 3.75 6.75 5.10 6.50 SOARCA lifetime 3.75 6.75 5.10 6.50 Minimal variation observed in key event timings across the longer fuel rod lifetime simulations (reference, sensitivity #2, and sensitivity #3) supports that fuel rod failure is dominated by alternate fuel failure models in this sequence, including the fuel relocation temperature. Some variation in key event timings are observed when the lifetime function is reduced. Similarly, small variation is observed in the in-containment source term, shown in Figure 5-16 for Peach Bottom and Figure 5-17 for Surry, when reduced fuel rod lifetimes are considered. Variations observed in both the time of lower head failure and in-containment source terms as a result of modified fuel lifetime functions, however, are small relative the variability observed across the accident scenario set. Thus, variations of the fuel rod lifetime function considered in this analysis does not impact the key results of this source term study relative to other considered uncertainties.

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Figure 5-16 Fission Product Release to Containment for Peach Bottom Short-Term SBO Fuel Rod Lifetime Sensitivity 128

Figure 5-17 Fission Product Release to Containment for Surry SBO Fuel Rod Lifetime Sensitivity 5.2.6. Fuel Relocation Temperature Material interactions between the fuel and cladding materials cause the liquefaction and early failure of fuel rods. It is postulated that these material interactions will proceed at different rates for HBU cores leading to fuel failure according to different temperature criteria. Two methods are considered in this analysis to investigate the effect of material interactions on HBU core fuel rod failure and subsequent in-containment source terms. The first approach entails modification of the effective liquefaction temperature of UO2 and ZrO2 to approximate the effects of material interactions between fuel and cladding materials on fuel rod failure19. The second approach 19 In general, this approach also requires modification of the failure temperature of fuel rods with fully oxidized cladding to match the UO2/ZrO2 liquefaction temperature.

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entails activating the eutectics model, which treats material interactions explicitly through composition dependent mixture properties for selected material pairs including UO2/ZrO2.

• Reference case: assumes 2479.0 K fuel relocation temperature

• Sensitivity #1: assumes increased fuel relocation temperature 2728.0 K

• Sensitivity #2: assumes reduced fuel relocation temperature 2230.0 K

• Sensitivity #3: assumes default eutectics model activated (interactive materials model deactivated)20 The impact of fuel relocation temperature variation on key event times for the Peach Bottom short-term SBO simulated are more significant relative to other sensitivity studies considered.

Table 5-13 Fuel relocation temperature sensitivity accident timings Peach Bottom [h] Surry [h]

Lower Plenum Dryout Lower Head Failure Lower Plenum Dryout Lower Head Failure Reference Case 3.75 6.75 5.10 6.50 Low Fuel Relocation Temperature 3.60 6.30 5.38 7.17 High Fuel Relocation Temperature 4.90 7.95 5.60 6.85 Eutectics model 3.15 6.45 4.85 6.17 Figure 5-18 shows the significant impact that fuel relocation temperature has on peak fuel temperatures for both the Peach Bottom and Surry simulations; higher fuel relocation temperatures result in higher peak fuel temperatures and reduced fuel relocation temperatures result in reduced peak fuel temperatures. Differences between the peak core temperatures observed in the reference and eutectics model cases appear only after significant core degradation.

20 The failure temperature of fuel rods with fully oxidized cladding is not assumed to change relative to the reference calculation. No modifications were made to eutectics model parameters.

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Figure 5-18 Peak Fuel Temperature Transient in Peach Bottom Short-Term (left) and Surry (right)

SBO Fuel Relocation Temperature Sensitivity In-containment source terms shown in Figure 5-19 and Figure 5-20 follow much the same patterns. For Peach Bottom, nearly all radioactive noble gases are released to the containment in the first 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of each simulation. Halogen releases over the first 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> roughly correlate to the fuel relocation temperature, with the largest releases being observed for the highest fuel relocation temperature and the smallest releases being observed for the lowest fuel relocation temperature and eutectics model sensitivity, which show the strongest agreement. At the end of the 12-hour period, the reference simulation lies in the center of the distribution. Alkali metals follow suit, with the exception of the low fuel relocation temperature case, which exhibits larger in-containment source terms than both the reference and eutectics model cases; The reference case and eutectics model sensitivity show the strongest agreement for Alkali Metal releases. Te group releases follow similar trends observed for radioactive halogens, with the largest releases observed for the highest fuel relocation temperature followed by the reference calculation. The lowest fuel relocation temperature calculation initially shows the lowest releases, but at the end of the 12-hour period exhibits larger releases than the eutectics model calculation.

The Surry calculations exhibit somewhat different characteristics. In general, the high fuel relocation temperature, low fuel relocation temperature, and reference calculations agree, particularly at the end of the first 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of the accident. The eutectics model, an outlier in this sensitivity, exhibits the largest releases earlier in the transient for noble gases, halogens, and Te group releases because of more significant early degradation of the core between 3 and 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. At the end of the 12-hour transient, however, the eutectics model exhibits the lowest releases for halogens, alkali metals, and Te group radionuclides. Another notable difference is observed for the lowest fuel relocation temperature sensitivity, which only exhibits approximately 90% of the noble gases inventory released to containment after 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The reference, eutectics model, and highest fuel relocation temperature sensitivity calculations all indicate that nearly 100% of noble gases are released to containment after 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

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Overall, fuel relocation exhibits the greatest impact on in-containment source terms of the group of sensitivities investigated in response to ATF Severe Accident PIRT findings. In-containment source term variability observed for variations in fuel relocation temperatures are still generally smaller relative to the variability observed across the accident scenario set, however, the strong impact of fuel relocation temperatures on accident progression and in-containment source terms may warrant further investigation. Further analysis of simulations utilizing the eutectics models are of particular interest.

Figure 5-19 Fission Product Release to Containment for Peach Bottom Short-Term SBO Fuel Relocation Temperature Sensitivity 132

Figure 5-20 Fission Product Release to Containment for Surry SBO Fuel Relocation Temperature Sensitivity 5.2.7. Hot Leg Creep Rupture As discussed in previous sections, the in-containment source terms are sensitive to high/low pressure scenarios and primary pressure boundary failure during critical phases of the accident progression. To highlight this sensitivity, a Surry calculation was performed with hot leg creep rupture disabled to investigate the impact of delayed hot leg creep rupture beyond key release phases at during the first 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of the accident transient. The impact of hot leg creep rupture on in-containment source terms is shown in Figure 5-21. When hot leg creep rupture is disabled, both core damage progression and radionuclide releases are impacted.

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• For the reference case, hot leg creep rupture occurred at 3.29 hours3.356481e-4 days <br />0.00806 hours <br />4.794974e-5 weeks <br />1.10345e-5 months <br />, lower plenum dryout occurred at 5.10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />, and lower head failure occurred at 6.50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br />

• For sensitivity #1, hot leg creep rupture did not occur, lower plenum dryout occurred at 3.85 hours9.837963e-4 days <br />0.0236 hours <br />1.405423e-4 weeks <br />3.23425e-5 months <br />, and lower head failure occurred at 5.27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br /> Noble gas releases are slightly delayed when hot leg creep rupture does not occur, and the quantity of radioactive noble gases that reach containment during the first 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of the scenario is also smaller (approximately 90% without hot leg creep rupture versus approximately 100% with hot leg creep rupture). Without hot leg creep rupture, Halogens, Alkali Metals, and Te group radionuclide releases to containment are significantly reduced during the first 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> period. Halogen releases are approximately 10% without hot leg creep rupture and almost 60%

with hot leg creep rupture. Less than 5% of the radioactive Alkali Metals inventory is released to containment without hot leg creep rupture while nearly 40% is released with hot leg creep rupture. Te group releases are approximately 10% without hot leg creep rupture and almost 50%

with hot leg creep rupture.

Thus, the impact of hot leg creep rupture timing, and more broadly the timing of vessel depressurization, during critical accident phases is confirmed to be a significant factor for both accident progression and in-containment source term. Were hot leg creep rupture to be delayed relative to reference calculation predictions, a moderate reduction of radionuclide inventories reaching containment could reasonably be expected. It should be noted that the reduction the in-in-containment source terms resulting from preservation of the primary pressure boundary is due to greater retention of radionuclides in the primary system, which includes the pressure relief tank, and not a reduction in core damage observed or a less progressed accident. Improved modeling of hot leg creep rupture in this analysis leads to the difference between predicted halogen releases in SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011) and this analysis.

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Figure 5-21 Fission Product Release to Containment for Surry SBO Hot Leg Creep Rupture Sensitivity 5.3. Release Rates Previous, related source term studies such as NUREG-1465 (Soffer, Burson, Ferrell, Lee, &

Ridgely, February 1995), SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011), and SAND2016-12954 (Gauntt, Goldman, Kalanich, & Powers, 2016) have employed an approximation of constant release rates. Constant release rates are obtained by dividing the release fraction of a selected radionuclide class during a given phase by the phase duration. This assumption is understood to be a simplified approximation of the true radionuclide release rates adopted to facilitate the regulatory process.

Assuming constant release rates, the containment releases in this analysis are in greater agreement with past studies. In fact, the prolonged accident phases tend to reduce containment 135

release rates for key risk-significant radionuclides including noble gasses, halogens, and alkali metals to values less than the release rates calculated using NUREG-1465 source terms and phase durations (assuming uniform releases), as shown in 5-14 and 5-15. Two notable exceptions to this reduction in containment releases exist for the Mo group and Te group releases. As discussed above, these radionuclide groups transport more efficiently to containment than estimated in NUREG-1465 in light of experimental evidence from the Phébus programme (Clement & Haste, 2004).

As discussed in SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011), this assumption is particularly significant for early in-vessel phase releases. The early in vessel phase encompasses the transition from an intact reactor core through states of core degradation across a variety of distinct pathways over a large temperature range. Thus, significant variations in release rate over time during the early in-vessel phase occur due to the complexity and broad range of core degradation scenarios that can occur, given unique accident-specific boundary conditions.

There are several factors that promote periods of increased fission product release from fuel-bearing material, either in the original reactor core or in debris beds formed because of core degradation.

Larger release rates are expected during periods of significant core degradation when (a) regions of the reactor core are at significantly elevated temperatures, and (b) significant amounts of structural failure, material degradation and relocation are occurring. Under these conditions, physical barriers (e.g., fuel cladding) preventing release of fission products from fuel structures, while the high fuel temperatures promote effective diffusion of fission products out of fuel grains to vaporize and be released into the local atmosphere.

Fission product release rates from fuel, however, are typically lower during periods of slower/progressive heat-up or cooldown of core structures and debris, during which thermal hydraulic transport phenomena would dominate radionuclide mass transfer.

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Table 5-14 Comparison of recommended BWR constant release rates with those of SAND2011-0128 and NUREG-1465. Darker shades - larger than recommendations from both previous studies. Lighter shades - larger than NUREG-1465 or SAND2011-0128 release recommendations.

Gap Release Early In-vessel Late In-vessel Ex-vessel NUREG- NUREG- NUREG- NUREG-2021 2011 2021 2011 2021 2011 2021 2011 Report 1465 1465 1465 1465 Noble Gases 0.023 0.050 0.10 0.14 0.12 0.63 0.0001 0.001 0.0 0.003 0.003 0.0 Halogens 0.007 0.013 0.10 0.11 0.059 0.17 0.004 0.033 0.001 0.006 0.004 0.100 Alkali Metals 0.007 0.013 0.10 0.047 0.016 0.13 0.0006 0.004 0.001 0.003 0.003 0.12 Te Group 0.005 0.013 0.0 0.091 0.049 0.033 0.005 0.028 0.0005 0.001 0.0007 0.083 Ba/Sr Group 0.0010 0.0 0.0 0.0009 0.0006 0.013 5.7E-05 0.0004 0.0 0.013 0.010 0.033 Ru Group <1.0e-6 0.0 0.0 0.0009 0.0003 0.002 2.2E-06 8.3E-06 0.0 <1.0e-6 0.0009 0.0008 Mo Group 2.7E-05 0.0 0.0 0.017 0.003 0.002 5.3E-05 0.0005 0.0 6.0E-06 0.001 0.0008 Lanthanides <1.0e-6 0.0 0.0 <1.0e-6 <1.0e-6 0.0001 <1.0e-6 0.0 0.0 1.2E-05 1.7E-05 0.002 Ce Group <1.0e-6 0.0 0.0 <1.0e-6 <1.0e-6 0.0003 0.0 0.0 0.0 0.0009 0.0007 0.002 137

Table 5-15 Comparison of recommended PWR constant release rates with those of SAND2011-0128 and NUREG-1465. Darker shades - larger than recommendations from both previous studies. Lighter shades - larger than NUREG-1465 or SAND2011-0128 release recommendations.

Gap Release Early In-vessel Late In-vessel Ex-vessel NUREG- NUREG- NUREG- NUREG-2021 2011 2021 2011 2021 2011 2021 2011 Report 1465 1465 1465 1465 Noble Gases 0.019 0.077 0.10 0.21 0.21 0.73 0.0008 2.1E-05 0.0 0.009 0.002 0.0 Halogens 0.003 0.018 0.10 0.16 0.082 0.27 0.001 0.001 0.010 0.009 0.002 0.12 Alkali Metals 0.001 0.014 0.10 0.15 0.051 0.19 0.0005 0.0004 0.010 0.008 0.004 0.17 Te Group 0.003 0.018 0.0 0.15 0.067 0.038 0.0008 0.0007 0.0005 0.002 0.0006 0.12 Ba/Sr Group 0.0006 0.003 0.0 0.0007 0.0009 0.015 3.1E-06 0.0 0.0 0.005 0.0006 0.050 Ru Group <1.0e-6 0.0 0.0 0.002 0.001 0.002 2.1E-06 0.0 0.0 <1.0e-6 0.0005 0.001 Mo Group 1.5E-05 0.0 0.0 0.045 0.018 0.002 6.8E-05 0.0002 0.0 0.0009 0.002 0.001 Lanthanides <1.0e-6 0.0 0.0 <1.0e-6 <1.0e-6 0.0002 <1.0e-6 0.0 0.0 7.2E-06 2.7E-06 0.003 Ce Group <1.0e-6 0.0 0.0 <1.0e-6 <1.0e-6 0.0004 <1.0e-6 0.0 0.0 0.0003 5.0E-05 0.003 138

5.4. Aerosol Retention and Removal Mechanisms Aerosol retention and removal mechanisms function to limit and remove radionuclides from the containment atmosphere decreasing inventory of radionuclides that can be released the environment. Engineered aerosol retention and removal systems include containment sprays, BWR suppression pools, filtration systems, ice condenser beds, and submersion of heat bearing debris (in- or ex-vessel). The efficiency and also relevance of such systems are sensitive to the accident progression and fission product release pathways (e.g., scenarios in which fission product release pathways bypass the suppression pool). While such systems were not directly considered in the release fractions presented in Section 5.1, additional information from the analysis performed enables a demonstration of radionuclide inventory accumulation in the BWR suppression pool.

Table 5-16 shows radionuclide release fractions to containment with and without the radionuclide inventory accumulated in the suppression pool water inventory. The results are included to demonstrate the integral effect of the suppression pool on relevant simulations and facilitate comparison to previous studies including SOARCA. Table results show the difference between total accumulated containment inventory (including the suppression pool inventory) and the containment inventory accumulated outside of the suppression pool (excluding the suppression pool inventory). Data is only available for the gap release and early in-vessel phase, prior to lower head failure. Radionuclide groups that exhibit the greatest accumulation in the suppression pool are Mo, Ru, Alkali Metal, Te, and halogens, in that order when considering total containment releases. Unreactive radionuclides such as noble gases and radionuclides that bypass the suppression pool (e.g. are released to containment with corium or released as vapors through a bypass) are not efficiently retained or removed from the in-containment source term.

Further aerosol retention by containment, in other words further reduction of the potential environmental release, is expected in BWRs through other engineered aerosol removal systems such as containment sprays or by natural processes such as gravitational settling, diffusive deposition, diffusiophoresis, and thermophoresis. Similarly, engineered aerosol removal and retention systems and the same natural processes function to reduce radionuclide releases during severe accidents in PWRs. Though deposition of radionuclides in containment by these engineered systems and natural processes was not considered in the same detail as the BWR suppression pool, their combined effect for select accident scenarios are presented in Appendix B. The results presented in Appendix B indicate that in general, a significant fraction, and even the vast majority in some cases, of radionuclides released to containment are deposited in-containment though one or more of these mechanisms significantly reducing the potential environmental releases.

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Table 5-16 Derived BWR release fractions including and excluding the suppression pool inventory for all core variations (60 GWd/MTU, 80 GWd/MTU, LEU and HALEU).

Gap Release Early In-vessel Total (end of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />)

Release Including Suppression Excluding Suppression Including Suppression Excluding Suppression Including Suppression Excluding Suppression Category Pool Inventory Pool Inventory Pool Inventory Pool Inventory Pool Inventory Pool Inventory Noble Gases 0.016 0.016 0.95 0.95 1 1 Halogens 0.005 1.30E-06 0.71 0.06 0.87 0.2 Alkali Metals 0.005 1.20E-06 0.32 0.006 0.35 0.039 Te Group 0.003 <1.0e-6 0.56 0.038 0.78 0.26 Ba/Sr Group 0.0006 <1.0e-6 0.005 0.0003 0.048 0.042 Ru Group <1.0e-6 <1.0e-6 0.006 7.40E-06 0.006 0.0001 Mo Group 1.90E-05 <1.0e-6 0.12 0.0001 0.13 0.002 Lanthanides <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 3.70E-05 3.60E-05 Ce Group <1.0e-6 <1.0e-6 <1.0e-6 <1.0e-6 0.003 0.003 140

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6.

SUMMARY

Representative in-containment source terms have been developed for BWR and PWR plants with HBU and elevated enrichment fuels to extend the recommendations of NUREG-1465 and address these technologies. Reactor cores considered in this analysis utilize appropriate decay heats and radionuclide inventories that represent core average loadings with fuel at elevated burnups including:

• Core-average burnup of 60 GWd/MTU at 5 wt% enrichment,

• Core-average burnup of 80 GWd/MTU at 5 wt% enrichment,

• Core-average burnup of 60 GWd/MTU at 8 wt% enrichment (peak at 10 wt% for BWRs),

and

• Core-average burnup of 80 GWd/MTU at 8 wt% enrichment (peak at 10 wt% for BWRs).

The in-containment source terms have been generated following the methodology for developing generic, but representative, source terms first demonstrated in NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995). To develop in-containment source terms representative of the nuclear reactor fleet deployed in the United States, multiple reactor types have been modeled including BWR/Mk-I, BWR/Mk-III, PWR/large-dry, and PWR/ice condenser reactors.

To generate representative in-containment source terms, MELCOR simulations are performed for each accident sequence-core type combination; an ensemble of accident scenarios that reflect the principal contributors to historical core damage frequency estimates are considered. A non-parametric statistical analysis is then applied to the quantities of interest derived from MELCOR simulation results to evaluate their respective distributions. Representative quantities of interest are taken as the median value of the derived distributions. For consistency with the definition of in-containment source terms in RG 1.183 and previous in-containment source term analyses (NUREG-1465 and SAND2011-0128), the in-containment source term reported in this analysis include the entire radionuclide inventory within containment.

Results from this analysis suggest that an increase in burnup and enrichment does not strongly impact the in-containment source term in terms of radionuclide release fractions. Changes to radionuclide inventories as a result of increased burnup and enrichment, however, may impact subsequent dose calculations and feasibility of HBU/HALEU implementation.

Relative to previous studies, including NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) and earlier considerations of HBU fuel in SAND2011-0128 (Powers, Leonard, Gauntt, Lee, & Salay, 2011) and SAND2016-12954 (Gauntt, Goldman, Kalanich, & Powers, 2016), there are some important differences identified in the in-containment source terms estimated in this analysis. Nonetheless, in terms of fractional releases, source terms developed for HBU and HBU/HALEU fuel do not differ significantly from NUREG-1465 recommendations except where the state-of-knowledge and/or state-of-practice has evolved.

Differences that have arisen relative to NUREG-1465 (Soffer, Burson, Ferrell, Lee, & Ridgely, February 1995) are attributable to the following factors.

• Chemical form of iodine. Advancements in the understanding of radiochemistry during severe accidents have led to modification of the dominant chemical forms of iodine assumed 142

to occur during a severe accident. In the current state of practice, 100% of iodine is assumed to react with Cesium to form CsI, whereas NUREG-1465 assumed 95% of iodine in the form of CsI and no more than 5% of iodine in the form of volatile molecular iodine (I2) and other volatile organic iodides. Both studies assume 5% of the total iodine inventory is present in the gap inventory.

• Chemical form of cesium. The current understanding of cesium chemistry during severe accidents has also been advanced, altering the dominant chemical forms of cesium assumed to occur during severe accidents. Current best-practice assumes that 5% of cesium is present in the gap inventory, in the form of both CsI and CsOH. All remaining cesium is assumed reacted with Mo to form Cs2MoO4. At the time NUREG-1465 was written, cesium was assumed to be present predominantly in the form of volatile CsOH.

• Mo release. Since cesium release is primarily in the form of Cs2MoO4, Mo releases are higher than in past studies. As a result, Mo releases are now higher than other metallic fission products such as Ru and Pd.

• Te release. Advancement in the understanding of Te chemistry during severe accidents indicates more extensive Te release than reported in NUREG-1465. More efficient transport of Te to containment occurs because Te is no longer assumed to take a chemical form that readily reacts with fuel cladding or other metallic surfaces in the reactor, thereby decreasing Te retention in the reactor vessel and primary circuit.

• Heterogeneous, integrated reactor core modeling. In modern MELCOR analysis, the reactor core is discretized into axial levels and radial rings and stronger coupling between severe accident phenomena are modeled through an integrated approach. These advancement in modeling practice has numerous implications:

o 2D discretization of the reactor core allows modelling of radial and axial power profiles, resulting in more efficient distribution of heat in the core, heat removal from core, and progressive degradation of the core o MELCOR no longer exhibits a distinct gap release phase because localized core degradation can occur prior to release of the entire gap inventory. Instead, MELCOR captures progressive changes of state across the core domain, in which severe accident phenomena are no longer isolated to a given accident phase.

o Gap-release and early in-vessel phase durations are prolonged because heat is more efficiently distributed in and removed from the core during the boiloff transient relative to the STCP.

• Low-pressure accident sequences. Relative to NUREG-1465, the current study considers more low-pressure accident sequences resulting in higher containment releases.

o Thermally induced SRV seizure was exhibited by the majority of BWR accident sequences considered in this analysis, unlike NUREG-1465. Greater sampling of low-pressure scenarios resulted in commensurately greater releases of halogens, alkali metals, Te group, and Mo group radionuclides. Radionuclide releases for accident sequences in which the vessel remained at high-pressure prior to vessel failure, however, exhibited greater consistency with NUREG-1465.

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o Hot leg creep rupture was exhibited by the majority of PWR accident sequences considered in this analysis, unlike NUREG-1465. As with BWRs, greater sampling of low-pressure scenarios resulted in commensurately greater releases of halogens, alkali metals, Te group, Mo, and Ru group radionuclides. Radionuclide releases for accident sequences in which the vessel remained at high-pressure prior to vessel failure, however, exhibited greater consistency with NUREG-1465.

• Increased late in-vessel releases. Radionuclide release during the late in-vessel phase was generally larger than NUREG-1465 recommendations, though still significantly smaller than early in-vessel phase releases. This is in part due to the prolonged duration of the late in-vessel phase, particularly for BWRs. Furthermore, many simulations exhibited some degree of continued core degradation during the late in-vessel phase, which contributed to the observed increase in releases.

• Reduced ex-vessel releases. The contribution of the ex-vessel release phase to the overall in-containment source term is reduced relative to NUREG-1465. This is because core debris was depleted of most of the volatile and semi-volatile fission products during the prolonged early in-vessel phase. Furthermore, a greater fraction of zirconium was oxidized during the prolonged early in-vessel phase, which reduced the metallic zirconium available to interact with structural concrete.

Differences that have arisen relative to SAND2011-0128 (Powers, Leonard, Gauntt, Lee, &

Salay, 2011) and SAND2016-12954 (Gauntt, Goldman, Kalanich, & Powers, 2016), are attributable to the following factors.

• Advanced severe accident modeling state-of-practice.

o Grand Gulf release fractions are consistent with Grand Gulf release fractions used in SAND2011-0128.

o Peach bottom modeling strategies have been improved since the SAND2011-0128, especially during the Peach Bottom SOARCA project. Peach Bottom release fractions are consistent with release fractions reported in SOARCA, which exhibited a limited halogen retention by the primary system during the early-in vessel phase.

o Both Surry and Sequoyah modeling strategies have been improved since the SAND2011-0128, especially during the SOARCA project. Release fractions for both Surry and Sequoyah are consistent with release fractions reported in SOARCA, which exhibited a limited halogen retention by the primary system after hot leg creep rupture.

• Low-pressure accident sequences. Relative to SAND2011-0128 and SAND2016-12954, the current study considers more low-pressure accident sequences resulting in higher containment releases.

o Hot leg creep rupture was exhibited by the majority of PWR accident sequences considered in this analysis, unlike SAND2011-0128 and SAND2016-12954. As with BWRs, greater sampling of low-pressure scenarios resulted in commensurately greater releases of halogens, alkali metals, Te group, Mo, and Ru group radionuclides. This is due to the enhanced modeling of Reactor Coolant System natural circulation under severe accident conditions introduced in the SOARCA 144

project. This modeling results in a very strong tendency for hot leg creep rupture to occur soon after the onset of core degradation in a PWR. This failure of the Reactor Coolant System induces a flow path to containment through which fission products can more readily move during the in-vessel accident phase. In prior studies, the absence of a hot leg creep rupture led to more sequences progressing to lower head failure with the reactor vessel remaining at high pressure. Under high-pressure scenarios, there is more limited flow of volatile fission products, like CsI, into containment.

• HBU/HALEU fuel sensitivities. Sensitivity study results indicate that the separate effects of variations to fuel thermal conductivity, in-vessel particulate debris porosity, in-vessel particulate debris diameter, particulate debris falling velocity in the lower plenum, and fuel rod lifetimes introduce inconsequential variability into in-containment source terms relative to the variance observed across accident scenarios. The fuel relocation temperature sensitivity study, however, exhibited notable variability in both core damage progression (i.e., key event timings) and in-containment source term. It is worth noting that general agreement is observed between the reference case and eutectics sensitivity in-containment source terms. Ultimately, variability observed in the in-containment source terms are smaller than the degree of variability observed across accident scenarios.

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APPENDIX A. PLANT-INDEPENDENT MELCOR MODELING PRACTICES EMPLOYED IN THIS ANALYSIS A.1. CAV Package Modeling Options Default Modified Record Calculations Field Description Value Value CAV_U All calculations boiling (value) 10.0 1.0 Boiling curve modification ingress ON OFF Water ingress model flag A.2. COR Package Modeling Options Default Modified Record Calculations Field Value Value Description COR_ROD All calculations, IRODDAMAGE none Not modified Fuel failure lifetime as a function of except TaT model cladding temperature (tabular sensitivities function name), MELCOR employs a default time-at-temperature fuel failure lifetime function should this record be unspecified.

COR_EUT All calculations, IEUMOD ON OFF Eutectics model activation flag except eutectics model sensitivity COR_LHM All calculations LHMCOR ON OFF Lower head melting model activation flag COR_CR All PWR IAICON NACT - not ACTC - Silver release model activation flag calculations active Vaporization is allowed from candling material only COR_SC All calculations, 1131 (2) 2400.0 2350.0 Zircaloy melt breakout temperature except reduced fuel relocation temperature sensitivity All calculations, 1132 (1) 2500.0 2479.0 Oxidized fuel rod failure temperature, except fuel reduced to UO2-INT/ZRO2-INT relocation liquefaction temperature per best-temperature practice sensitivities All calculations 1141 (2) 1.0 0.2 Maximum molten pool drainage rate 1504 (1) 10x unit 1.0e-5 Maximum relative error permitted in roundoff total volume of any cell 1030 (4) 10.0 0.1 Characteristic time for coupling dT/dz and average control volume temperatures 149

A.3. CVH/FL Package Modeling Options Modified Record Calculations Field Default Value Value Description CVH_SC All calculations 4401 (3) 8 + number of flow 50.0 Number of velocity calculation iterations paths in model before subcycle time step reduction 4414 (1) 0.01 Not Minimum hydrodynamic volume fraction Modified A.4. HS Package Modeling Options Default Modified Record Calculations Field Value Value Description HS_SC All calculations 4055 (2) 5.0e-4 1.0e-3 Conduction convergence criteria A.5. MP Package Modeling Options Default Modified Record Calculations Field Value Value Description MP_PRC: UO2- All calculations, except TMLT 3113.0 2479.0 UO2 liquefaction temperature, reduced INT eutectics model to account for known fuel-cladding sensitivity material interactions MP_PRC: ZRO2- All calculations, except TMLT 2990.0 2479.0 ZrO2 liquefaction temperature, reduced INT eutectics model to account for known fuel-cladding sensitivity material interactions A.6. RN Package Modeling Options Default Modified Record Calculations Field Value Value Description RN1_SC All calculations 7100 (2) 0.0 0.1 Zr release multiplier for CORSOR release model 7100 (3) 0.0 1.0 ZrO2 release multiplier for CORSOR release model 150

APPENDIX B. HBU ACCIDENT PROGRESSION AND SOURCE TERM REFERENCE ANALYSES This appendix provides a discussion of accident progression and source term reference analyses for selected accident scenarios of each of the plants analyzed. For the two PWR plants considered, the following scenarios are discussed in more detail below

• Surry Scenario 1B, which is a small-break loss of coolant accident (SBLOCA) with complete loss of the ECCS. The auxiliary feedwater and containment spray systems operate in injection mode.

• Sequoyah Scenario 4A, which is a RCP seal leak accident with cavity flooding and late containment failure.

For the two BWR plants analyzed in this analysis, the following scenarios are discussed in more detail below

• Peach Bottom Scenario 1A is a STSBO. Thermal SRV seizure induces low-pressure conditions at the time of vessel failure. Drywell linear melt-through and drywell head flange leakage containment failure modes are assumed.

• Grand Gulf Scenario 5A is also a STSBO. In this accident sequence there is no coolant injection, thermal SRV seizure induces low-pressure conditions at the time of vessel failure, which ignites the hydrogen in containment and causes early containment failure.

B.1. PWR Accident Progression and Source Term Reference Analyses B.1.1. Large-Dry Containment: Surry The accident progression for SU1b is discussed in this section. Refer to Section C.1.2 for the corresponding event timing table for exact timing of sequence events.

Sequence SU1b is initiated by a small break LOCA in the cold leg. AC and DC power are available throughout the accident. Component cooling water is available, preventing the reactor pump seals from being damaged. The ECCS fails, along with the containment spray system and fan coolers. The auxiliary feedwater system is available. The containment is assumed to fail due to over-pressurization of the containment.

The following parameters are plotted from the 60 and 80 GWd/MTU for both normal enrichment and HALEU cases:

• Thermal hydraulic response o Core water level o Reactor vessel pressure o Containment pressure o Containment temperature

• Reactor core degradation o Core damage 151

o Corium ejection

• Fission product release fractions (fractions of initial inventory)

The lower head and containment failures are marked on each plot in addition to the model responses for each metric.

B.1.1.1. Thermal Hydraulic Response Accident sequence SU1b begins when a small break LOCA opens in the cold leg. The reactor successfully scrams, and the main feedwater system trips. The ECCS system fails to start, but the motor-driven AFW operates as intended. Containment sprays operate in injection mode only.

The SBLOCA begins at time = 0. Figure B-1 shows the core water level response with the top and bottom of the active fuel indicated with dashed lines. The core water level dropped below the top of the active fuel at ~2.0 hr and the first gap release occurred at ~2.9 hr. The core water level dropped below the bottom of the active fuel for the first time at ~3.0 hr. At ~6-7 hr, slumping of the core material into the lower head momentarily increases the water level in the reactor vessel to above the original top of active fuel, after which water inventory steadily decreases until lower head dryout or vessel failure.

Lower head failure occurred between ~5.5 and ~12.5 hr, depending on the burnup and enrichment considered. The lower head failure is indicated on all plots by the hashed gold vertical filled sections which continue until the grey vertical filled sections which mark containment failure.

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Figure B-1 Surry 1b Pressure Vessel collapsed water level for sequence 1b Figure B-2 shows the vessel pressure response during the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the sequence.

Immediately following the sequence onset, the vessel depressurizes quickly from ~17 MPa to

~8MPa, where it is maintained by auxiliary feedwater injection. As shown in Figure B-3, the motor-driven auxiliary feedwater (MDAFW) stops operating ~2 hr after the start of the sequence.

Shortly thereafter, the reactor vessel pressure begins to subside. Intermittent pressure spikes coincide with core material relocation into the lower head, which momentarily accelerates steam production.

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Figure B-2 Surry 1b Vessel Pressure 154

Figure B-3 Surry 1b motor-driven auxiliary feedwater flowrate Throughout the accident scenario, the pressure and temperature inside containment both steadily increase until very close to vessel failure, when the first hydrogen rapidly increases both the pressure and temperature for a short time. This pressure spike is insufficient to cause containment failure and conditions quickly stabilize. The containment pressure and temperature then continue to rise steadily until containment beings to fail due to over-pressurization.

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Figure B-4 Surry 1b Containment Pressure 156

Figure B-5 Surry 1b Containment Temperature 157

B.1.1.2. Reactor Core Degradation Figure B-6 SU1b Peak Core Temperatures As shown in Figure B-7, the core is damaged extensively during these sequences over the course of about an hour, between 3.5 and 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> from the start of the accident.

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Figure B-7 Surry 1b Core Damage Percentage As shown in Figure B-8, corium is ejected from the reactor vessel immediately following vessel failure. Corium continues to escape the vessel for several hours and slows to a trickle 159

Figure B-8 Surry 1b Corium Ejection B.1.1.3. Fission Product Release Fractions Total fission product release fractions to containment for the noble gases, halogen, alkali metals, Te group, Ba/Sr group, Ru group, Mo group, lanthanides, and Ce group are shown below. Each figure shows the deposited fraction (solid green) and airborne fraction (dash-dotted blue) of the radionuclide inventory in containment as well as the fraction of the radionuclide inventory that has escaped containment (dashed teal). The total fraction of radionuclides released to containment (the sum of the airborne, deposited, and escaped fractions) is also shown in solid black. In this analysis deposited radionuclides include all radionuclides that are not airborne (i.e., scrubbed in the suppression pool, settled, etc.). Radionuclides that have escaped containment include those radionuclides that have reached the environment or been released into the enclosure building. Vertical shaded regions are used to indicate vessel depressurization (by 160

SRV seizure), and containment failure (coincident with vessel failure in this scenario). It is important to note that the majority of the releases to containment are deposited in containment with the exception of the noble gases.

Figure B-9 below shows the release fraction of noble gases to the containment. Essentially all of the noble gas present in the reactor is rapidly released to the containment prior to vessel depressurization. A small portion of the noble gas slowly escapes during this phase, but the majority remains airborne in containment. Once the vessel and containment fail, the fraction of noble gas that escapes rapidly increases until practically all of the noble gas has escaped containment.

Figure B-9 Surry 1b Noble Gas Release Fraction to Containment Figure B-10 shows the release fractions of halogens. Approximately 10%-20% of halogens are released to containment during the early in-vessel phase through open SRVs. Most of these 161

halogens quickly settle and are deposited in the containment. Once the reactor vessel and containment have failed, most of the halogens deposit in containment, while between 5% - 20%

escapes, depending on the case. For case 80 GWd/MTU Haleu case, some of the previously deposited material becomes airborne once more. This coincides with the ejection of core melt into containment which can lead to re-aerosolization.

Figure B-10 Surry 1b Halogen Release Fraction to Containment Figure B-11 shows the release fraction of Alkali metals in containment. The results are similar to those shown for the halogen release fractions, although a larger portion of the alkali metals deposit in containment, allowing approximately 2% of alkali metals to escape containment.

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Figure B-11 Surry 1b Alkali Metals Release Fraction to Containment The release fractions of fission products in the Te group that reach containment are shown in Figure B-12. Approximately 20% of the Te group mass is released to containment prior to vessel failure. Following the failure of the reactor vessel and containment, additional Te group products are released which are quickly deposited. In one case, the same phenomenon as described for the halogens occurs, leading to previously deposited Te group mass being reintroduced into the airspace. Depending on the case, between 5% - 10% of Te group mass is released to the environment 163

Figure B-12 Surry 1b TE group release fraction to containment Release fractions to containment for the Ba/Sr group is shown in Figure B-13. A much smaller fraction of this group is observed to escape the vessel during SRV operation compared to previously discussed groups. In all cases, less than 2% of the Ba/Sr group present in the reactor arrives in the containment. A negligible amount of Ba/Sr escapes from containment prior to containment failure. After containment failure, less than 0.5% is found to escape, while the remainder stays in the reactor vessel or settles in containment.

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Figure B-13 Surry 1b Ba/Sr groups release fraction to containment Figure B-14 shows the release fractions to containment of the Ru group. In all cases less than 1%

of the reactor vessel Ru group inventory is released to containment. Of that small amount, the majority quickly deposits. A very small fraction escapes containment shortly after containment failure, totaling less than 0.05% in all cases 165

Figure B-14 Surry 1b Ru group release fraction to containment The release fractions of Mo that reach containment are shown in Figure B-15. In the worst case, as much as 20% of the reactor Mo inventory is released to the containment where the majority is deposited. A small fraction, less than 1% in all cases escapes to the environment following vessel and containment failure.

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Figure B-15 Surry 1b Mo group release fraction to containment The release fraction of lanthanides is shown in Figure B-16. No significant quantity of lanthanides were observed to escape the reactor vessel prior to vessel failure. Following vessel and containment failure, a small fraction of the total reactor lanthanide inventory is released, where the majority is deposited in containment. In the worst case, the fraction of lanthanides that reach containment is 1'10-4, while the fraction that escapes to the environment is no more than 2'10-5.

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Figure B-16 1b Lanthanides release fraction to containment As shown in Figure B-17, the release fractions of the Ce group to containment are very low, like they were for lanthanides. However, in one case there was a significantly higher fraction of Ce group mass released from the reactor following containment failure. In this case, approximately 0.2% of all the Ce in the reactor reached the containment, with about 0.08% escaping to the environment. In the other three cases, the releases were much smaller, with a fraction less than 2'10-5 reaching containment, and less than 5'10-6 escaping to the environment.

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Figure B-17 Surry 1b Ce group release fraction to containment B.1.2. Ice Condenser Containment: Sequoyah The accident progression for SQN4a is discussed in this section. Refer to Section C.2.1 for the corresponding event timing table for exact timing of sequence events.

Sequence SQN4a is initiated by an RCP seal LOCA on the single, non-lumped loop. AC and DC power are available throughout the transient. AFW is available for coolant injection. In containment, fan coolers are operational and containment sprays operate by injection mode only.

Cavity flooding does occur. Containment fails late by overpressure.

The following parameters are plotted from the 60 and 80 GWd/MTU for both normal enrichment and HALEU cases:

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• Thermal hydraulic response o Core water level o Reactor vessel pressure o Containment pressure o Containment temperature

• Reactor core degradation o Core damage o Corium ejection

• Fission product release fractions (fractions of initial inventory)

The lower head and containment failures are marked on each plot in addition to the model responses for each metric.

B.1.2.1. Thermal Hydraulic Response While none of the individual Sequoyah accident sequences were exactly identical, there were no significant differences in the accident sequences or reactor or containment transient signatures which warranted a separate discussion. As such, the accident progression discussed herein is the typical sequence observed in all of the Sequoyah 4a simulations.

The Sequoyah 4a sequence initial conditions are as follows:

• Single (i.e., non-lumped) RCP seal failure at 0.0 h

• Offsite AC power and onsite DC power sources are available

• TDAFW is available with operator control

• Containment sprays are available in injection-only mode (i.e., no recirculation)

• Air return system fans are available

• Valves between the lower containment and cavity are open to allow for cavity flooding

• No early containment failure conditions are enforced.

The 4a accident sequence initiates when the RCP seal on the single (non-lumped) loop fails. The leak rate at the operating pressure is 8.01 kg/s at the nominal cold leg coolant density of 747.0 kg/m3, or 170 gpm. The volumetric flow rate transient from the failed RCP seal in gpm is shown in Figure B-18. The initial flow rate of 8.0 kg/s quickly decreases as the primary circuit pressure decreases from TDAFW operation. The oscillations between 3.0 h to 9.0 h occur simultaneously with atmospheric relief valve cycling on the secondary circuits. Leakage through the failed RCP seal stops around 12.0 h when hot leg creep rupture depressurizes the primary circuit and equilibrates the containment and primary circuit pressures.

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Figure B-18 Sequoyah 4a mass flow rate through single loop failed RCP seal.

The decrease in reactor pressure created by the seal failure trips the reactor at 0.4. The TDAFW immediately after the reactor SCRAMs. The TDAFW flow rate rates to the single and triple SGs and the single and triple SG levels are shown in Figure B-19 and Figure B-20, respectively.

Following the TDAFW actuation signal, the TDAFW injects coolant into the single and triple SGs at a combined mass flow rate of 55.0 kg/s (880 gpm). While the flow rate into the single SG is constant, the flow into the lumped SG begins to decrease around 1.0 h. The cause for the asymmetrical TDAFW response is unknown. The coolant injected by the TDAFW removes heat from the primary circuit and reduces the primary circuit pressure. TDAFW injection stops when the single SG fills and floods the steam supply line.

171

Figure B-19 Sequoyah 4a TDAFW flow rate to single and lumped (triple) SGs.

172

Figure B-20 Sequoyah 4a single and lumped (triple) SG levels.

The reactor primary and secondary pressures are shown in Figure B-21 and Figure B-22 respectively. Following TDAFW failure, the single and lumped SGs quickly pressurize, and the SG pressure begins to oscillate between the atmospheric relief valve lower and upper setpoints of 6.88 MPa and 7.16 MPa, respectively. The coolant level in the SGs decreases gradually as the ARVs vent, and the SGs dry out between 8.5 h to 9.0 h. After the SGs dry out, the secondary pressure gradually decreases while the primary pressure increases and oscillates between the PORV lower and upper setpoints of 15.55 MPa and 16.02 MPa. The primary circuit depressurizes around 12.0 h from hot leg creep failure resulting from the high pressures and gas temperatures in the primary system.

173

Figure B-21 Sequoyah 4a primary system pressures.

174

Figure B-22 Sequoyah 4a primary system pressures.

The RPV coolant level is shown in Figure B-23. The RPV initially decreases as coolant leaks through the failed seal. Starting when the steam generator ARVs begin to cycle around 3.0, the RPV level remains at a relatively constant level. Once the steam generators dry out and the lowest setpoint PORV begins to cycle, the RPV level quickly decreases as decay heat and oxidation energy vaporize coolant which vents through the PORV. Around 12.0 h, hot leg creep failure depressurizes the primary circuit which initiates accumulator injection. The coolant injected by the accumulators quickly boils off, and the RPV dries out around 15.0 h.

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Figure B-23 Sequoyah 4a RPV Level The RPV level has two atypical transients after LHF. At about 21.0 h, the RPV level swells to BAF, and, similarly, around 60.0 h, the RPV level increases to or above TAF. The level transients are a result of water from the containment sprays and melted ice in the ice chest flooding the cavity above the elevation of the LHF vessel breach which allows liquid coolant to flow back into the RPV before draining back out. This occurs because the combined volume of the water sprayed into containment from the RWST and the water from the melted ice baskets is enough to fill the reactor cavity above the LHF breach. As all fuel has since melted by 21.0 h when the first fill transient occurs, this does not impact the core degradation sequence.

Figure B-24 shows the containment pressure transient. The containment pressure effectively remains constant near 0.1 MPa until about 7.5 h when the containment sprays and circulation fans initiate. The containment fans indirectly circulate air through the ice condensers which reduces the pressure while also melting the ice. Around 9.5 h, the rupture disc on the PRT bursts 176

which creates a small pressure increase in containment and creates a pathway for hydrogen generated in-vessel to transport to containment. When hot leg failure occurs around 11.5 h occurs, the hot gasses from the hot leg rupture ignite the hydrogen in containment which creates a series of deflagrations between 11.5 h to 15.0 h. At about 16.5 h, the lower head fails, and the containment begins to pressurize from core-concrete interactions. The ejected corium acts as a persistent ignition source, and the hydrogen and carbon monoxide produced by core-concrete interactions create a series of deflagrations that exceed the 0.18 MPa (12 psig) containment design pressure between 16.5 h to 24.0 h. However, the 0.46 MPa (60 psig) containment rupture pressure is never reached at any point.

Figure B-24 Sequoyah 4a Containment Pressure Figure B-25 shows the fraction of ice in the ice condensers that has melted. Minimal ice melts until 7.5 h. At 7.5 h, the containment air return system activates. The air return system creates a circulation pattern inside the containment vessel that forces air through the ice condensers. While this suppresses the containment pressure, the heat transfer from the containment atmosphere to 177

the ice condensers melts the ice which subsequently flows into the lower containment volume and floods the reactor cavity.

Figure B-25 Sequoyah 4a Melt Fraction of Ice in Ice Condensers B.1.2.2. Reactor Core Degradation Figure B-26 shows the peak fuel temperature in each COR ring where a 0.0 K temperature indicates all fuel in the COR ring has failed, and Figure B-28 shows the core damage fraction where core damage is quantified as the ratio of fuel mass outside the active core region (i.e., in the form of particulate debris) to the initial fuel loading. Prior to the RPV level reaching TAF around 9.0 h, the fuel temperatures remain constant at about 600.0 K. After the RPV level decreases below TAF, the peak fuel temperatures increase from the combined deposition of decay heat and oxidation energy into the fuel. The onset of core damage occurs around 11.5 h when the peak fuel temperature first reaches 2479.0 K, the best estimate melting point of the UO2-ZrO2 eutectic. The peak fuel temperatures temporarily decrease following accumulator 178

injection around 12.0 h. However, the peak fuel temperatures quickly increase back to 2479.0 K and core damage continues to accumulate as the fuel melts. At about 14.0 h, all of the fuel has melted and relocated to the lower plenum.

Figure B-26 Sequoyah 4a Peak Fuel Temperature in COR Rings 1 through 5 179

Figure B-27 Sequoyah 4a Core Damage Fraction B.1.2.3. Fission Product Releases Total fission product release fractions to containment for the noble gases, halogen, alkali metals, Te group, Ba/Sr group, Ru group, Mo group, lanthanides, and Ce group are shown in Figure 29 to Figure 37. Each figure shows the deposited fraction (solid green) and airborne fraction (dash-dotted blue) of the radionuclide inventory in containment as well as the fraction of the radionuclide inventory that has escaped containment (dashed teal). The total fraction of radionuclides released to containment (the sum of the airborne, deposited, and escaped fractions) is also shown in solid black. In this analysis deposited radionuclides include all radionuclides that are not airborne (i.e., scrubbed in the suppression pool, settled, etc.). Radionuclides that have escaped containment include those radionuclides that have reached the environment or been released into the enclosure building. Vertical shaded regions are used to indicate vessel 180

depressurization. It is important to note that the majority of the releases to containment are deposited in containment with the exception of the noble gases.

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Figure B-28 shows release of the noble gases to containment. The noble gases transport efficiently to the containment upon their release. This results in approximately 1% of the noble gases reaching the containment prior to vessel depressurization. After vessel depressurization, the remainder of the noble gases quickly accumulate in the containment. At the end of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the majority of the noble gases remain airborne in the containment with a much smaller fraction (less than 0.1%) having escaped. Noble gases are non-reactive, which leads to negligible deposition of radionuclides.

Figure B-28 Sequoyah 4a Noble Gases Release to Containment 182

The fraction of halogens released to containment is shown in Figure B-29. The largest fraction (approximately 85%) of halogen radionuclides are deposited in containment at the end of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The next largest fraction of the halogen (less than 1%) remain airborne at the end of the simulation. Since at the end of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the containment has not failed due to overpressure or another mechanism, only a very small portion of the halogens (less than 0.00004%) have escaped containment.

Figure B-29 Sequoyah 4a Halogens Release to Containment 183

The fraction of alkali metals radionuclides released to containment is shown in Figure B-30.

Between approximately 82% to 87% of the alkali metals are deposited in the containment at the end of the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> simulation. A smaller fraction of alkali metals (between 0.03% to 0.05%)

remain airborne at the end of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Only a very small portion of the halogens (less than 0.00003%) have escaped containment by the simulation end time.

Figure B-30 Sequoyah 4a Alkali Metals Release to Containment 184

The fraction of the Te group radionuclides released to containment is shown in Figure B-31.

Similar to the halogens and alkali metals, the majority of Te group radionuclides released are deposited in the containment and a much smaller fraction (less than 0.1%) of the radionuclides remain airborne after containment failure. During the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> duration, a very small amount of the Te group radionuclides (less than 0.00003%) escape containment.

Figure B-31 Sequoyah 4a Te Group Release to Containment 185

The fraction of the Ba/Sr group radionuclides released to containment is shown in Figure B-32Figure B-51. Similar to the alkali metals and halogens group, the Ba/Sr group radionuclides accumulate in the containment, with the largest fraction (between approximately 1% to 3%)

being deposited and less than 0.0003% remaining airborne by the simulation conclusion at 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. A negligible percentage of radionuclides within this group escape containment by the end of the simulation.

Figure B-32 Sequoyah 4a Ba/Sr Group Release to Containment 186

The fraction of Ru group radionuclides released to containment is shown in Figure B-33. Nearly all Ru group radionuclides that are released to containment are deposited. After 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, a negligible fraction of the Ru group remains airborne or has escaped the containment.

Figure B-33 Sequoyah 4a Ru Group Release to Containment 187

The fraction of the Mo group radionuclides released to containment is shown in Figure B-34.

Similar to other Sequoyah release fractions documented in this appendix, the largest fraction of the total Mo group release is deposited in the containment by the end of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, with a much smaller percentage (less than 0.0002%) remaining airborne, and nearly negligible fraction (less than 0.00001%) escaping the containment.

Figure B-34 Sequoyah 4a Mo Group Release to Containment 188

The fraction of Lanthanides group radionuclides released to containment is shown in Figure B-

35. The total release is negligible up until the vessel is depressurized. After vessel depressurization, the largest fraction of released Lanthanides are deposited (between 0.001% to 0.004%), with the next largest fraction being those that are airborne. A negligible amount of the Lanthanide group escapes containment.

Figure B-35 Sequoyah 4a Lanthanides Release to Containment 189

The fraction of Ce group radionuclides released to containment is shown in Figure B-36. Similar to the Lanthanides group, the total release is negligible up until the vessel is depressurized. By the end of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, the majority of the released Ce group has deposited (between approximately 1% to 0.02%), with a much smaller fraction still remaining airborne. Almost none of the Ce group radionuclides escape the containment.

Figure B-36 Sequoyah 4a Ce Group Release to Containment 190

B.2. BWR Accident Progression and Source Term Reference Analyses B.2.1. BWR Mark I Containment: Peach Bottom The accident progression observed in PB1a is briefly described in the following section. Refer to Section C.3.1 for the corresponding event timing table for exact timing of sequence events.

Sequence PB1a is a short term station black out scenario. No AC or DC power is available throughout the sequence. Additionally, no coolant is available during the accident sequence.

PB1a is defined as having low vessel pressure at vessel breach. This is due to a stuck open SRV that depressurizes the RPV prior to lower head failure. Initial containment failure is defined as either drywell liner melt-through or leakage of the drywell head flange.

The following parameters are plotted from the 60 and 80 GWd/MTU for both normal enrichment and HALEU cases:

• Thermal hydraulic response o Core water level o Steam dome pressure o Steam dome temperature o Containment pressure (wetwell and drywell) o Containment temperature (wetwell and drywell) o Peak temperatures

• Reactor core degradation o Core damage o Corium ejection

• Fission product release fractions (fractions of initial inventory)

The lower head and containment failures are marked on each plot in addition to the model responses for each metric.

B.2.1.1. Thermal Hydraulic Response The short-term system blackout began at time 0 s. All power and coolant injection was lost at this time. Figure B-37 shows the core water level response with the top and bottom of the active fuel indicated with dashed lines. Soon after the beginning of the accident sequence, the core water level dropped below the top of the active fuel at ~0.5 hr and the first gap release occurred at ~0.9 hr. The core water level dropped below the bottom of the active fuel near time ~2 hr.

Lower plenum dryout time varied between cases and occurred between ~4.0 to ~4.2 hr.

Figure B-38 shows the vessel pressure response during the first 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of the sequence. Prior to SRV failure, the SRVs were operating as intended in safety relief mode. At time ~2.3 hr, SRV 1 stuck open due to cycling at temperatures exceeding 1000 K. This caused the RPV to depressurize rapidly from pressure ~8 MPa to ~0.6 MPa. A vessel pressure spike is observed at

~4 hr during all cases, which can be attributed to relocation of debris causing a rapid boiloff of water in the vessel.

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Lower head failure subsequently occurred between ~7.3 and ~7.6 hr (depending on the burnup and enrichment). On all plots, the hashed gold vertical filled sections indicate the lower head failure and continue until the grey vertical filled sections which mark containment failure.

In all accident sequences where drywell liner melt is an allowable containment failure, the drywell liner is assumed to melt when it first meets either of the two conditions: (a) the liner exceeds a maximum allowed temperature or (b) the model time is greater than ten minutes after the lower head failure.

All cases have the drywell liner melt-through occurring before drywell head flange leakage. The pressure responses at the drywell (Figure B-40) and wetwell (Figure B-41) show a sharp pressure increase at RPV failure and then a rapid depressurization of the containment at containment failure. A subsequent drywell head flange failure occurred approximately 24 to 36 minutes after the drywell liner melt when the head flange lifted due to overpressure and overheating conditions. (see Appendix C.3.1).

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Figure B-37 PB1a Core Water Level 193

Figure B-38 PB1a Steam Dome Pressure 194

Figure B-39 PB1a Steam Dome Temperature 195

Figure B-40 PB1a Drywell (Upper) Pressure 196

Figure B-41 PB1a Wetwell (Lower) Pressure 197

Figure B-42 PB1a Drywell (Upper) Vapor Temperature 198

Figure B-43 PB1a Wetwell (Lower) Liquid Temperature 199

B.2.1.2. Reactor Core Degradation Core degradation occurs shortly after the core is uncovered, as illustrated in Figure B-44 and Figure B-45, which show the RPV water level and core damage fraction, respectively. Core damage fraction is used as a representation of the core degradation. Core damage begins to accumulate at 1.2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during scenario PB1a. By 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and in all cases for PB1a, the core is severely degraded. Prior to lower head failure, the core damage fraction reaches 1.0.

Core ejection mass is shown in Figure B-46. Between approximately 334 and 378 MT of material is ejected following the lower head failure.

Figure B-44 Peak Temperatures 200

Figure B-45 PB1a Core Damage 201

Figure B-46 PB1a Corium Ejection B.2.1.3. Fission Product Release Fractions Total fission product release fractions to containment for the noble gases, halogen, alkali, Te group, Ba/Sr group, Ru group, Mo group, lanthanides, and Ce group are shown below. Each figure shows the deposited fraction (solid green) and airborne fraction (dash-dotted blue) of the radionuclide inventory in containment as well as the fraction of the radionuclide inventory that has escaped containment (dashed teal). The total fraction of radionuclides released to containment (the sum of the airborne, deposited, and escaped fractions) is also shown in solid black. In this analysis deposited radionuclides include all radionuclides that are not airborne (i.e., scrubbed in the suppression pool, settled, etc.). Radionuclides that have escaped containment include those radionuclides that have reached the environment or been released into the enclosure building. Vertical shaded regions are used to indicate vessel depressurization (by SRV seizure), and containment failure (coincident with vessel failure in this scenario).

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Figure B-47 shows release of the noble gases to containment. As vapors, noble gases transport efficiently to containment upon release from the reactor fuel through the SRVs. As most of the inventory has already reached containment at the time of vessel depressurization, this has minimal effect on the total noble gases release. Starting at containment failure, the noble gases quickly escape containment and the majority are released to the environment. Only a very small fraction of noble gas radionuclides are retained in containment (less than 0.0001%) as airborne radionuclides. Noble gases are non-reactive and have large vapor pressures which leads to negligible deposition.

Figure B-47 PB1a Noble Gases Release to Containment 203

The fraction of halogens released to containment is shown in Figure B-48. The majority of halogen radionuclides are deposited in containment after flowing through the SRVs prior to vessel depressurization. A smaller fraction (approximately 10%) of halogens are released to the environment after containment failure. After 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, a very small amount of iodine (less than 0.001%) remains airborne in the containment.

Figure B-48 PB1a Halogens Release to Containment 204

The fraction of alkali metals radionuclides released to containment is shown in Figure B-49.

Similar to the halogens, the majority of alkali metals that are released to containment are deposited after flowing through the SRVs prior to vessel depressurization. A smaller fraction (between approximately 2% to 5%) of alkali metals are released to the environment after containment failure. After 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, a very small amount of the alkali metal radionuclides (less than 0.005%) remains airborne in the containment.

Figure B-49 PB1a Alkali Metals Release to Containment 205

The fraction of the Te group radionuclides released to containment is shown in Figure B-50.

Similar to the halogens and alkali metals, the majority of the Te group radionuclides released to containment are deposited prior to vessel depressurization and a smaller fraction (between approximately 13% to 16%) are released to the environment after containment failure. After 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, a small amount of the Te group radionuclides (between 0.01% and 0.03%) remains airborne in the containment.

Figure B-50 PB1a Te Group Release to Containment 206

The fraction of the Ba/Sr group radionuclides released to containment is shown in Figure B-51.

The majority of the Ba/Sr group radionuclides escape containment after drywell failure which allows the radionuclides to leave the containment directly with a smaller fraction (between approximately 3% to 5%) being deposited. This is in contrast to previously described groups, which typically show the largest fraction of radionuclides being deposited. After 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, a small amount of the Ba/Sr group radionuclides (between 0.001% and 0.004%) remains airborne in the containment.

Figure B-51 PB1a Ba/Sr Group Release to Containment 207

The fraction of Ru group radionuclides released to containment is shown in Figure B-52. Nearly all Ru group radionuclides that are released to containment are deposited, with a smaller fraction (between 0.004% and 0.01%) escaping after containment failure. After 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, a negligible percentage (less than 0.0001%) of the Ru group remains airborne in the containment.

Figure B-52 PB1a Ru Group Release to Containment 208

The fraction of the Mo group radionuclides released to containment is shown in Figure B-53.

The majority of the Mo group radionuclides escape containment with a smaller fraction (22% to 44% depending on the case) being deposited. Similar to Ba/Sr group, the drywell failure allows the radionuclides to leave the containment directly, which results in a sharp increase of the escaped release fraction at containment failure. Later, at approximately 40 to 52 hours6.018519e-4 days <br />0.0144 hours <br />8.597884e-5 weeks <br />1.9786e-5 months <br /> (depending on the case), the escaped release fraction increases again as a result of the preferential oxidation of the Mo group after the depletion of the other metal inventories. This results in the final escaped release fraction exceeding the deposited release fraction. After 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, a small amount of the Mo group radionuclides (less than ~0.001%) remains airborne in the containment.

Figure B-53 PB1a Mo Group Release to Containment 209

The fraction of Lanthanides group radionuclides released to containment is shown in Figure B-

54. The total release is negligible up until containment failure. After containment failure, the largest fraction of Lanthanides escapes containment (between 0.008% to 0.03%) as a result of drywell failure which allows the radionuclides to leave the containment directly. A smaller fraction (approximately 0.002% to 0.01%) is deposited by 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The Lanthanides group radionuclides that remain airborne is negligible at the end of the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> runtime.

Figure B-54 PB1a Lanthanides Release to Containment 210

The fraction of Ce group radionuclides released to containment is shown in Figure B-55. The total release is negligible up until containment failure. After containment failure, the largest fraction of Ce group escapes containment (approximately 0.4% to 2%) as a result of drywell failure which allows the radionuclides to leave the containment directly. Approximately 0.1% to 0.7% is deposited by 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The amount of Ce group radionuclides that remain airborne is negligible at the end of the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> runtime.

Figure B-55 PB1a Ce Group Release to Containment 211

B.2.2. BWR Mark III Containment: Grand Gulf The accident progression for GG5a is discussed in this section. Refer to Section C.4.1 for the corresponding event timing table for exact timing of sequence events.

Sequence GG5a is initiated by a STSBO. No AC or DC power is available during the duration of the accident. Coolant injection is also unavailable. The accident sequence characterizes the vessel as being at low pressure during vessel breach due to prior seizure of the safety/relief valve in the open position. Containment failure is assumed to occur due to a deflagration event at the time of vessel failure.

The following parameters are plotted from the 60 and 80 GWd/MTU for both normal enrichment and HALEU cases:

• Thermal hydraulic response o Core water level o Reactor vessel pressure o Steam dome temperature o Containment pressure (wetwell and drywell) o Containment temperature (wetwell and drywell)

• Reactor core degradation o Core damage o Corium ejection

• Fission product release fractions (fractions of initial inventory)

The lower head and containment failures are marked on each plot in addition to the model responses for each metric.

B.2.2.1. Thermal Hydraulic Response The STSBO begins at time t=0s. Due to loss of power (including backup DC batteries), coolant injection is unavailable. The loss of core cooling capabilities results in core heat-up and boiloff.

Core boiloff is characterized by decreasing water level - vessel pressure is maintained by SRV cycling - and increased temperatures in the upper regions of the RPV and uncovered regions of the core. The core and downcomer water levels are shown in Figure B-56. At about 2/3 top of active fuel, the top of the jet pumps, the downcomer water level decouples from the core. Core boiloff decelerates as the water level drops to lower core and lower plenum elevations. After failure of the core support plate, large quantities of core debris relocate to the lower plenum, accelerating lower plenum boiloff until lower plenum dryout.

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Figure B-56 GG5a Core Water Level Figure B-57 shows the pressure of the RPV steam dome. As stated, during the initial core boiloff and heat-up transient, vessel pressure is maintained by SRV cycling. The core plate fails around 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> after the loss of power and debris relocation to the lower plenum begins. Shortly after core plate failure, thermal seizure of an SRV occurs. Unabated flow through the seized valve results in rapid depressurization of the primary circuit. A short period of sustained vessel pressure is observed just before it equilibrates with containment pressure due to steam produced during boiloff of the lower plenum water inventory. Vessel remains in equilibrium with containment pressure until containment failure.

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Figure B-57 GG5a Steam Dome Pressure Steam dome temperatures rise above operating temperatures through the core boiloff and heat-up transient as shown in Figure B-58. A significant temperature excursion can be observed after uncovered core materials begin to oxidize. In particular, exothermic oxidation of Zr-based cladding and canister structures causes rapid heat-up of the core and production of non-condensable H2. Oxidation of core materials slows after the core water level reaches lower core elevations and steam production is reduced. When large quantities of core debris relocate to the lower plenum, the water inventory is converted to steam that subsequently rises through the core region, oxidizing remaining unoxidized materials leading to another oxidation driven temperature excursion. Temperatures continue to rise beyond lower plenum dryout until after lower head failure, when massive debris ejection relocates the majority of core debris ex-vessel.

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Figure B-58 GG5a Steam Dome Temperature Figure B-59 shows the wetwell annulus pressure. During the core boiloff and heat-up phase, wetwell pressure increases. Pressure increases in containment are driven by the expulsion of hot, non-condensable gases through the SRV. Steam from the primary circuit that is directed to the wetwell suppression pool through the SRV is condensed in the suppression pool, increasing its temperature. Upon lower head failure, a hydrogen deflagration event occurs in containment causing containment failure. Figure B-60 shows the vapor temperature in the wetwell. A small rise in wetwell temperatures is observed prior to containment failure through the addition of hot gases; the suppression pool does not reach saturation temperatures during the transient.

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Figure B-59 GG5a Wetwell Pressure 216

Figure B-60 GG5a Wetwell Temperature Drywell pressures and temperatures, shown in Figure B-61 and Figure B-62, respectively, follow much the same behavior has that observed in the wetwell. Prior to containment failure, drywell pressures and temperatures increase. After lower head failure, ejected core debris interacts with concrete in the pedestal beneath the RPV. The molten corium and concrete interaction (MCCI),

produces large quantities of vapors as the hot core debris ablates the concrete, heating and liberating previously trapped gases and H2O. The release and transport of hot gases and vaporized fission products increases the temperature of the containment atmosphere.

217

Figure B-61 GG5a Drywell Pressure 218

Figure B-62 GG5a Drywell Temperature B.2.2.2. Reactor Core Degradation Peak temperatures for fuel, cladding, particulate debris, oxidic molten pool, and metallic molten pool are shown in Figure B-63. Temperatures of 0.0 K signify the absence of a particular component (0.0 kg mass associated with that component). While the core is covered, fuel and cladding temperatures remain near operational temperatures (approximately 570 K). Core temperatures increase rapidly to failure temperatures (2479.0 K) during peak core oxidation and the formation of hot debris begins. Core temperatures decrease for a short period during vessel depressurization as a result of steam flow through the core. As the water/steam inventory remaining in the RPV are depleted, core temperatures begin to rise more slowly than before. Fuel and cladding temperatures do not rise as quickly during this time because of reduced decay heat

- significant fuel rod failure and debris formation has occurred relocating heat bearing materials downward in elevation - and limited oxidation. Complete failure of fuel and cladding 219

components is signified by 0.0 K temperatures prior to lower head failure. After lower head failure, core temperatures decrease as heat bearing materials are ejected from the RPV into the drywell cavity.

Figure B-63 GG5a Peak Core Temperatures Core degradation begins shortly after the core is uncovered. Core damage fraction is shown as a representative of core degradation in Figure B-64. Rapid oxidation and heat-up of the core starts as early as 1.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. By 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, the core is severely degraded, and the majority of the fuel rods have failed, with some persistent fuel rods remaining in the peripheral core regions. In all cases of this scenario (GG5a), the persistent peripheral structures collapse, and the core damage fraction reaches 1.0 prior to lower head failure.

220

Figure B-64 GG5a Core Damage Fraction Following gross lower head failure, massive debris ejection from the RPV occurs as shown in Figure B-65. About 300 MT of core debris are ejected upon lower head failure, with total ejected masses reaching nearly 400 MT over the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> in all cases.

221

Figure B-65 GG5a Ejected Mass B.2.2.3. Fission Product Release Fractions Total fission product release fractions to containment for the noble gases, halogen, alkali metals, Te group, Ba/Sr group, Ru group, Mo group, lanthanides, and Ce group are shown in Figure B-66 to Figure B-74. Each figure shows the deposited fraction (solid green) and airborne fraction (dash-dotted blue) of the radionuclide inventory in containment as well as the fraction of the radionuclide inventory that has escaped containment (dashed teal). The total fraction of radionuclides released to containment (the sum of the airborne, deposited, and escaped fractions) is also shown in solid black. In this analysis deposited radionuclides include all radionuclides that are not airborne (i.e., scrubbed in the suppression pool, settled, etc.). Radionuclides that have escaped containment include those radionuclides that have reached the environment or been released into the enclosure building. Vertical shaded regions are used to indicate vessel depressurization (by SRV seizure), and containment failure (coincident with vessel failure in this 222

scenario). It is important to note that the majority of the releases to containment are deposited in containment with the exception of the noble gases.

Figure B-66 shows noble gas releases to containment. As vapors, noble gases transport efficiently to containment upon release from the reactor fuel through the SRVs. Vessel depressurization has little effect on the noble gas releases, because most of the inventory has already reached containment. Noble gases also transport to the environment efficiently after containment failure. Only a small quantity of noble gas radionuclides are retained in containment (approximately 1%) as airborne radionuclides. Noble gases are non-reactive and have large vapor pressures leading to negligible deposition.

Figure B-66 GG5a noble gases containment release fraction 223

The fraction of halogens released to containment is shown in Figure B-67. The majority of halogens that are released to containment prior to vessel depressurization are retained in the suppression pool after flowing through the SRVs. Very little iodine remains airborne in the containment, having either deposited or escaped. Halogen release to the environment occurs after containment failure, with less than 10% of the halogens inventory released to the environment at the end of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in all four cases.

Figure B-67 GG5a Halogens containment release fraction 224

The fraction of alkali metal radionuclides that reach containment are shown in Figure B-68. As with halogens, the majority of alkali metals are deposited in containment, with a small fraction of material remaining airborne. Around 1% of the alkali metal inventory is released to the environment after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in all cases.

Figure B-68 GG5a Alkali metals containment release fraction 225

Figure B-69 shows Te group releases to containment. Similar to the halogen and alkali metal releases to containment, the majority of Te group releases are deposited in containment with a small fraction of airborne material. Between 5% and 9% of the Te group inventory is released to the environment after 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

Figure B-69 GG5a Te group containment release fraction 226

The release fraction of Ba/Sr group radionuclides to containment is shown in Figure B-70. As with the previous radionuclide groups, most of the containment inventory of Ba/Sr group radionuclides are deposited, with a small fraction remaining airborne. No case exhibited Ba/Sr group releases to the environment in excess of 1% of the total inventory, with most cases exhibiting less than 0.5% of the total inventory released to containment.

Figure B-70 GG5a Ba/Sr group containment release fraction 227

The fraction of Ru group radionuclides released to containment are shown in Figure B-71. As with the previous radionuclide groups, the majority of the containment inventory is deposited, with a small fraction remaining airborne (less that 1x10-6). Ru group releases to the environment are also small, between 1x10-5 and 2x10-5.

Figure B-71 GG5a Ru group containment release fraction 228

Figure B-72 shows Mo group releases to containment. Between 10% and 20% of Mo group radionuclides are observed to reach and deposit in containment with a small fraction of material remaining airborne (less than 1x10-6 at the end of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />). The largest Mo group release fraction to escape containment is approximately 0.001.

Figure B-72 GG5a Mo group containment release fraction 229

The release fraction of lanthanides to containment is shown in Figure B-73. Lanthanide releases to containment are negligible prior to vessel failure, which is coincident with containment failure in this analysis). Initially, a fraction of lanthanides released to the containment are airborne, however, this portion of the lanthanide inventory quickly becomes negligible. The rest of the lanthanide inventory is generally deposited in containment while a small fraction escapes containment; the largest release fraction to escape containment is about 1x10-5.

Figure B-73 GG5a Lanthanides containment release fraction 230

Ce group releases to containment are shown in Figure B-74. Much like the lanthanide releases, Ce group releases prior to vessel failure are negligible. Also like the lanthanide releases, a small fraction of lanthanides released to the containment are initially airborne before quickly decreasing to negligible levels. The majority of Ce group radionuclides to reach containment are deposited, and a small fraction is released to the environment; the largest release fraction to escape containment is almost 1x10-3, the smallest fraction to escape containment is orders of magnitude lower at approximately 3x10-5.

Figure B-74 GG5a Ce group containment release fraction 231

APPENDIX C. ACCIDENT SEQUENCE EVENT TIMING TABLES C.1. Surry C.1.1. Surry, Case 1a Table C-1 Accident Sequence Event Timings for Surry, Case 1a Short-term Station Blackout. ECCS, AFW, CS, and FC all unavailable, Early Containment Failure Event 60 gwd/mtu 80 gwd/mtu 60 gwd/mtu 80 gwd/mtu (Time in hours unless noted otherwise) HALEU HALEU Loss of all off-site and on-site power 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

Reactor trip (SCRAM) 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

RCP trip on loss of power 0.1 [s] 0.1 [s] 0.1 [s] 0.1 [s]

Main feedwater trips on loss of power, AFW signal received 0.1 [s] 0.1 [s] 0.1 [s] 0.1 [s]

RCP seal cooling lost due to component cooling water 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

system failure RCP seals begin leak (approx 23 gpm/loop) 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

AFW fails to start 60.1 [s] 60.1 [s] 60.1 [s] 60.1 [s]

Steam generator reaches safety setpoint, SRVs begin to 130.0 [s] 130.0 [s] 130.0 [s] 130.0 [s]

cycle Pressurizer SRVs begin to cycle 1.5 1.4 1.5 1.4 PRT rupture disks fail 1.7 1.7 1.7 1.7 RPV level at TAF (begin gap release phase) 2.2 2.1 2.2 2.2 Start of fuel cladding failures 2.7 2.7 2.7 2.7 Total Xenon release exceeds 5% (begin early release 3 2.9 3 2.9 phase)

First relocation of fuel (UO2) and clad material (Zr) to lower 3.1 3.7 4 3.8 head Core support plate fails 3.3 3.7 4 3.8 Debris quench begins, steam cooling of particulate debris 3.3 3.7 4 3.8 above support plate Accumulator injection starts 3.4 3.3 3.4 3.3 Hot leg nozzle fails due to creep rupture 3.4 3.3 3.4 3.3 First hydrogen burn in containment 3.4 3.3 3.4 3.3 Accumulators empty 3.4 3.3 3.5 3.4 Lower core plate fails 4.3 4 4.4 4.2 Diffuser plate fails 4.6 4.8 5.3 5.3 Late in-vessel Cs release 95% complete 6.1 5.5 7.1 5.8 Vessel failure (begin late release phase) 6.7 6.2 7.1 7 Core debris relocation to cavity begins 6.7 6.2 7.1 7 Containment failure 6.7 6.2 7.1 7 Ex-Vessel Cs release 95% complete 7.2 13.8 11.2 9.2 Calculation terminated 3d 3d 3d 3d 232

C.1.2. Surry, Case 1b Table C-2 Accident Sequence Event Timings for Surry, Case 1b Short-term Station Blackout. AFW available. No ECCS, No CS, No FC, Containment Failure by overpressurization 60 80 Event 60 80 gwd/mtu gwd/mtu (Time in hours unless noted otherwise) gwd/mtu gwd/mtu HALEU HALEU Small-break LOCA in cold leg 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

Reactor trip (SCRAM) 107.0 [s] 107.0 [s] 107.0 [s] 107.1 [s]

Main feedwater trips, AFW signal received 110.0 [s] 110.0 [s] 110.0 [s] 110.0 [s]

ECCS signal received, ECCS fails to start 120.0 [s] 120.0 [s] 120.0 [s] 120.0 [s]

AFW starts after 60-second delay 167.0 [s] 167.0 [s] 167.0 [s] 167.1 [s]

RCP trip on high void 0.82 0.82 0.83 0.82 RPV level at TAF (begin gap release phase) 2 2 1.9 2 Start of fuel cladding failures 2.9 2.9 2.9 2.9 Total Xenon release exceeds 5% (begin early release phase) 3.3 3.3 3.3 3.3 Accumulator injection starts 3.4 3.4 3.4 3.3 Core support plate fails 3.5 5.1 3.6 4.8 First relocation of fuel (UO2) and clad material (Zr) to lower head 3.5 5.1 3.6 4.8 Debris quench begins, steam cooling of particulate debris above 3.5 5.1 3.6 4.8 support plate Lower core plate fails 4.6 5.5 4.5 5.9 Diffuser plate fails 4.9 5.8 4.8 6.5 7.4 8.4 7.1 5.6 Accumulators empty Late in-vessel Cs release 95% complete 8 5.9 8.6 7.5 Lower plenum dry 10.9 11.6 11 7.6 First hydrogen burn in containment 12 13.3 11.6 5.5 Vessel failure (begin late release phase) 11.7 13.1 11.2 8.7 Containment failure 58.4 61.8 66.2 N/A Core debris relocation to cavity begins 11.7 13.1 11.2 8.7 Ex-Vessel Cs release 95% complete 13.9 14.3 16.1 11.2 Calculation terminated 3d 3d 3d 3d 233

C.1.3. Surry, Case 1c Table C-3 Accident Sequence Event Timings for Surry, Case 1c Large-break LOCA, AFW available, ECCS in Injection Model Only, No CS, No FC, Early Containment Failure 60 80 Event 60 80 gwd/mtu gwd/mtu (Time in hours unless noted otherwise) gwd/mtu gwd/mtu HALEU HALEU Large-break LOCA occurs in cold leg 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

Reactor trip (SCRAM) 6.6 [s] 6.6 [s] 6.6 [s] 6.6 [s]

Containment spray signal, sprays fail to start 16.0 [s] 16.0 [s] 16.0 [s] 16.0 [s]

Main feedwater trips, AFW signal received 7.0 [s] 7.0 [s] 7.0 [s] 7.0 [s]

RPV level at TAF (begin gap release phase) 29.0 [s] 29.0 [s] 28.9 [s] 28.9 [s]

RCP trip on high void 24.0 [s] 24.0 [s] 24.0 [s] 24.0 [s]

Accumulator injection starts 50.4 [s] 50.4 [s] 50.4 [s] 50.4 [s]

AFW starts after 60-second delay 66.6 [s] 66.6 [s] 66.6 [s] 66.6 [s]

Accumulators empty 145.2 [s] 145.1 [s] 145.2 [s] 145.1 [s]

Start of fuel cladding failures 1.1 1.1 1.1 1.1 Total Xenon release exceeds 5% (begin early release phase) 1.3 1.3 1.3 1.3 Core support plate fails 1.9 2.1 2.3 1.9 First relocation of fuel (UO2) and clad material (Zr) to lower head 1.9 2.1 2.3 1.9 Lower core plate fails 2.5 2.2 2.3 2.1 2.9 2.4 2.6 2.4 Diffuser plate fails Lower plenum dry 3.6 2.9 3.2 3.1 Late in-vessel Cs release 95% complete 4.4 2.4 2.5 3.5 Vessel failure (begin late release phase) 5.2 4.1 4 4.1 Containment Failure 5.2 4.1 4 4.1 First hydrogen burn in containment N/A N/A 4 4.1 Core debris relocation to cavity begins 5.2 4.1 4.1 4.1 Ex-Vessel Cs release 95% complete 22.8 17.4 27.3 19.5 Calculation terminated 3d 3d 3d 3d 234

C.1.4. Surry, Case 1d Table C-4 Accident Sequence Event Timings for Surry, Case 1d STSBO, No ECCS, No AFW, No CS, No FC, No RCP Seal Failure, Early Containment Failure 60 80 Event 60 80 gwd/mtu gwd/mtu (Time in hours unless noted otherwise) gwd/mtu gwd/mtu HALEU HALEU Loss of all off-site and on-site power 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

Reactor trip (SCRAM) 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

RCP trip on loss of power 0.1 [s] 0.1 [s] 0.1 [s] 0.1 [s]

Main feedwater trips on loss of power, AFW signal received 0.1 [s] 0.1 [s] 0.1 [s] 0.1 [s]

AFW fails to start 60.1 [s] 60.1 [s] 60.1 [s] 60.1 [s]

ECCS signal received, ECCS fails to start 0.8 0.78 0.78 0.8 RPV level at TAF (begin gap release phase) 2.2 2.1 2.2 2.2 Start of fuel cladding failures 2.7 2.7 2.7 2.7 Total Xenon release exceeds 5% (begin early release phase) 3 2.9 3 3 Hot leg nozzle fails due to creep rupture 3.4 3.3 3.4 3.3 Accumulator injection starts 3.4 3.3 3.4 3.3 First hydrogen burn in containment 3.4 3.3 3.4 3.3 Accumulators empty 3.4 3.3 3.4 3.3 Core support plate fails 3.8 3.7 3.8 3.9 Debris quench begins, steam cooling of particulate debris above 3.8 3.7 3.8 3.9 support plate First relocation of fuel (UO2) and clad material (Zr) to lower head 3.8 3.7 3.8 3.9 Lower core plate fails 4.2 4.1 4.4 4.5 Diffuser plate fails 4.9 5.1 5.2 5.1 Lower plenum dry 5.5 5.6 5.7 5.6 Late in-vessel Cs release 95% complete 5.8 5.6 6.3 6.1 Vessel failure (begin late release phase) 6.5 6.7 6.9 6.6 Core debris relocation to cavity begins 6.5 6.7 6.9 6.6 Containment failure 6.5 6.7 6.9 6.6 Ex-Vessel Cs release 95% complete 11.7 11.2 8.3 12.3 Calculation terminated 3d 3d 3d 3d 235

C.1.5. Surry, Case 1f Table C-5 Accident Sequence Event Timings for Surry, Case 1f Small-break LLOCA, AFW available, No ECCS, No CS, No FC, Early Containment Failure 60 80 Event 60 80 gwd/mtu gwd/mtu (Time in hours unless noted otherwise) gwd/mtu gwd/mtu HALEU HALEU Small-break LOCA in cold leg 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

Reactor trip (SCRAM) 106.9 [s] 106.9 [s] 106.9 [s] 106.9 [s]

Main feedwater trips, AFW signal received 110.0 [s] 110.0 [s] 110.0 [s] 110.0 [s]

ECCS signal received, ECCS fails to start 120.0 [s] 120.0 [s] 120.0 [s] 120.0 [s]

AFW starts after 60-second delay 166.9 [s] 166.9 [s] 167.0 [s] 166.9 [s]

RCP trip on high void 0.83 0.82 0.82 0.83 RPV level at TAF (begin gap release phase) 2 1.9 2 1.9 Start of fuel cladding failures 2.9 2.9 2.9 2.9 Total Xenon release exceeds 5% (begin early release phase) 3.3 3.3 3.3 3.3 Accumulator injection starts 3.4 3.4 3.4 3.4 Core support plate fails 4.5 3.8 3.5 3.5 First relocation of fuel (UO2) and clad material (Zr) to lower head 4.5 3.8 3.5 3.5 Debris quench begins, steam cooling of particulate debris above support plate 4.5 3.8 3.5 3.5 Lower core plate fails 4.6 4.5 4.7 4.4 Diffuser plate fails 5 4.8 4.9 4.7 Accumulators empty 7.4 10.7 7 9.7 Late in-vessel Cs release 95% complete 8.4 8.3 8.7 8.2 Lower plenum dry 11.5 5.7 11 9.7 First hydrogen burn in containment 11.7 13.3 11.2 9.7 Vessel failure (begin late release phase) 12.6 10.6 11.8 9.7 Containment failure 12.6 10.6 11.8 9.7 Core debris relocation to cavity begins 12.6 10.6 11.8 9.7 Ex-Vessel Cs release 95% complete 23.6 34.8 18.1 19.2 Calculation terminated 3d 3d 3d 3d 236

C.2. Sequoyah C.2.1. Sequoyah, Case 4a Table C-6 Accident Sequence Event Timings for Sequoyah, Case 4a Reactor Coolant Pump (RCP) Seal Loss-of-Coolant Accident (LOCA), No Emergency Core Cooling System (ECCS), Turbine-Driven Auxiliary Feedwater (TDAFW) Operates in controlled mode, Containment Sprays (CS) Operate in Injection Mode Only, Cavity Flooded at Vessel Failure 60 80 Event 60 80 gwd/mtu gwd/mtu (Time in hours unless noted otherwise) gwd/mtu gwd/mtu HALEU HALEU Single loop RCP fails, leaks at 180 gpm 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

Reactor scram 0.39 0.39 0.39 0.39 First ARV opening 0.4 0.4 0.4 0.4 TDAFW injection starts 0.41 0.41 0.41 0.41 TDAFW floods, injection stops 2.3 2.3 2.3 2.3 Containment sprays start in injection mode 7.5 7.4 7.4 7.4 Containment circulation fans start 7.5 7.4 7.4 7.4 RWST < 10%, containment sprays stop 8 7.9 7.9 7.9 Single SG dryout 9.1 8.8 9.2 9 Triple SG dryout 9.1 8.8 9.2 9 RPV level at TAF 9.6 9.4 9.8 9.6 First PORV opening 9.6 9.3 9.7 9.5 PRT rupture disc fails 9.7 9.4 9.9 9.6 Start of cladding oxidation 10.8 10.4 11 10.6 First gap release 10.9 10.5 11.1 10.7 Hot leg creep failure 11.5 11.2 11.7 11.4 Accumulator injection starts 11.5 11.2 11.7 11.4 Accumulators depleted 11.6 11.2 11.8 11.5 First UO2 debris formation 11.5 11.9 12.5 11.4 First deflagration in containment 11.5 11.1 11.7 11.4 First UO2 debris on support plate 11.5 11.1 11.7 11.4 First UO2 debris in lower plenum 12.5 12.1 12.7 12.4 Support plate failure 12.5 12.1 12.7 12.4 Containment at 12 psig design pressure 13.2 12.9 13.6 13.1 Diffuser plate failure 13.3 13 13.9 13.2 Lower support plate failure 14.1 14.2 15.3 14.5 RPV dryout 15 15 16 15.4 Lower head failure 16.4 16.6 17.5 16.6 Containment reaches quasi-steady-state conditions 27.3 26.4 28.9 26.3 Containment fails at 60 psig N/A N/A N/A N/A Simulation end 3d 3d 3d 3d 237

C.2.2. Sequoyah, Case 4b Table C-7 Accident Sequence Event Timings for Sequoyah, Case 4b RCP Seal LOCA, No ECCS, TDAFW Operates in controlled mode, Containment Sprays Operate in Injection Mode Only, Cavity Dry at Vessel Failure 60 60 80 Event 80 gwd/mt gwd/mtu gwd/mtu (Time in hours unless noted otherwise) gwd/mtu u HALEU HALEU Single loop RCP fails, leaks at 180 gpm 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

Reactor scram 0.39 0.39 0.39 0.39 First ARV opening 0.4 0.4 0.4 0.4 TDAFW injection starts 0.41 0.41 0.41 0.41 TDAFW floods, injection stops 2.3 2.3 2.3 2.3 Containment sprays start in injection mode 7.6 7.5 7.6 7.6 Containment circulation fans start 7.6 7.5 7.6 7.6 RWST < 10%, containment sprays stop 8.1 8 8.1 8.1 Single SG dryout 9.1 8.9 9.2 9 Triple SG dryout 9.1 8.9 9.2 9 RPV level at TAF 9.6 9.4 9.8 9.6 First PORV opening 9.6 9.3 9.7 9.5 PRT rupture disc fails 9.7 9.4 9.9 9.6 Start of cladding oxidation 10.8 10.5 11 10.7 First gap release 10.9 10.6 11.1 10.8 Hot leg creep failure 11.6 11.2 11.7 11.4 Accumulator injection starts 11.6 11.2 11.7 11.4 Accumulators depleted 11.6 11.3 11.8 11.5 First UO2 debris formation 11.6 11.2 11.7 11.4 First UO2 debris on support plate 11.5 11.2 11.7 11.4 First deflagration in containment 12.3 11.2 11.7 11.4 First UO2 debris in lower plenum 11.6 11.2 12.7 12.6 UO2 debris quenched in lower plenum 11.6 11.2 12.7 12.6 Support plate failure 13.5 13.1 13.6 13.2 Diffuser plate failure 13.6 13.2 13.8 13.5 Containment at 12 psig design pressure 13.1 11.9 13 12.6 Lower support plate failure 14.9 14.1 14.8 14.6 RPV dryout 15.6 15 15.6 15.4 Lower head failure 16.6 16.7 16.7 16.7 Containment reaches quasi-steady-state conditions 30.1 33.3 35 32.2 Containment fails at 60 psig 40 41.6 42.3 41.8 Simulation end 3d 3d 3d 3d 238

C.2.3. Sequoyah, Case 4c Table C-8 Accident Sequence Event Timings for Sequoyah, Case 4c RCP Seal LOCA, ECCS Operates in Injection Mode, Motor-Driven Auxiliary Feedwater (MDAFW)

Operates, Containment Sprays Operate in Injection Mode , Cavity Dry at Vessel Failure 60 80 Event 60 80 gwd/mtu gwd/mtu (Time in hours unless noted otherwise) gwd/mtu gwd/mtu HALEU HALEU Single loop RCP fails, leaks at 180 gpm 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

Reactor trip 0.39 0.39 0.39 0.39 First ARV cycle 0.4 0.4 0.4 0.4 TDAFW injection starts 0.41 0.41 0.41 0.41 ECCS starts in injection mode 0.49 0.5 0.49 0.5 TDAFW floods, injection stops 1.4 1.4 1.4 1.4 Containment sprays start 3.6 3.6 3.6 3.5 Containment circulation fans start 3.6 3.6 3.6 3.5 RWST < 10%, containment sprays stop 4 4 4 4 Single SG dryout 11.8 11.6 12 11.8 Triple SG dryout 11.8 11.2 12 11.5 First primary valve cycle 12.3 11.9 12.6 12.2 RPV level at TAF 12.5 12.1 12.8 12.4 PRT rupture 12.5 12.1 12.8 12.4 Start of cladding oxidation 13.7 13.3 14.1 13.6 First gap release 13.8 13.5 14.2 13.7 Hot leg creep failure 14.5 14.1 14.9 14.4 Accumulator injection starts 14.5 14.1 14.9 14.4 Accumulator injection ends 14.5 14.1 15 14.4 First UO2 debris formation 15.4 14.1 16 14.4 First deflagration in containment 15.7 14.1 16.2 15.5 First UO2 debris on support plate 15.5 14.1 16 15.4 First UO2 debris in lower plenum 15.8 15.1 16.3 15.6 UO2 debris quenched in lower plenum 15.8 15.1 16.3 15.6 Support plate failure 16.6 15.6 17.1 16.6 Diffuser plate failure 16.7 15.9 17.2 16.7 Lower support plate failure 17.8 16.7 18.5 17.8 RPV dryout 18.3 17.5 19.4 18.5 Lower head failure 18.8 18.7 20.4 19.7 Containment at 12 psig design pressure 16.2 18.9 20.9 16.3 Containment reaches quasi-steady-state conditions 0.0 [s] 29.8 0.0 [s] 0.0 [s]

Containment fails at 60 psig 25 31.2 28.8 32.5 Simulation end 3d 3d 3d 3d 239

C.2.4. Sequoyah, Case 4d Table C-9 Accident Sequence Event Timings for Sequoyah, Case 4d Station Blackout (SBO), No ECCS, TDAFW Operates at Max Flow until Loss of Steam, No Containment Sprays, Cavity Dry at Vessel Failure 60 80 Event 60 80 gwd/mtu gwd/mtu (Time in hours unless noted otherwise) gwd/mtu gwd/mtu HALEU HALEU Loss of all offsite and onsite power 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

Reactor trip 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

RCP seals begin leak (approx 23 gpm/loop) 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

First secondary SV cycle 6.3 [s] 6.3 [s] 6.3 [s] 6.3 [s]

TDAFW injection starts 60.2 [s] 60.1 [s] 60.2 [s] 60.2 [s]

TDAFW floods, injection stops 2.2 2.2 2.2 2.2 Triple SG dryout 9.6 9.3 9.8 9.5 Single SG dryout 9.7 9.4 9.8 9.5 First primary SV cycle 9.8 9.5 10 9.7 PRT rupture 9.9 9.6 10.1 9.8 RPV level at TAF 10.2 9.9 10.4 10.1 Start of cladding oxidation 11.5 11 11.7 11.3 First gap release 11.7 11.2 12 11.5 Hot leg creep failure 12.4 11.9 12.7 12.2 Accumulator injection starts 12.4 11.9 12.7 12.2 Accumulator injection ends 12.4 11.9 12.7 12.3 First deflagration in containment 13.5 13 13.7 13.2 First UO2 debris formation 13.5 13 13.8 13.3 First UO2 debris on support plate 13.5 13 14.1 13.3 First UO2 debris in lower plenum 13.6 13.2 14.2 13.5 UO2 debris quenched in lower plenum 13.6 13.2 14.2 13.5 Support plate failure 14.4 13.8 14.7 14.2 Diffuser plate failure 14.6 14 14.9 14.3 Lower support plate failure 15.8 15.3 16.3 15.5 RPV dryout 16.5 15.7 17 16 Lower head failure 17.6 16.5 18 16.5 Containment at 12 psig design pressure 19.2 18.6 20.6 17.8 Containment reaches quasi-steady-state conditions 33.6 32.8 35.5 28.5 Containment fails at 60 psig N/A N/A N/A N/A Simulation end 3d 3d 3d 3d 240

C.2.5. Sequoyah, Case 4e Table C-10 Accident Sequence Event Timings for Sequoyah, Case 4e SBO, No ECCS, No AFW, No Containment Sprays, Cavity Dry at Vessel Failure, Early Containment Failure 60 80 Event 60 80 gwd/mtu gwd/mtu (Time in hours unless noted otherwise) gwd/mtu gwd/mtu HALEU HALEU Loss of all offsite and onsite power 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

Reactor trip 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

RCP seals begin leak (approx 23 gpm/loop) 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

First secondary SV cycle 6.2 [s] 6.4 [s] 6.3 [s] 6.3 [s]

First primary SV cycle 1.4 1.4 1.4 1.4 Triple SG dryout 1.6 1.6 1.6 1.6 Single SG dryout 1.6 1.6 1.6 1.6 PRT rupture 1.8 1.8 1.8 1.8 RPV level at TAF 2.1 2.1 2.1 2.1 Start of cladding oxidation 2.8 2.7 2.8 2.7 First gap release 2.8 2.8 2.9 2.8 First UO2 debris formation 3.3 3.3 3.4 3.3 First UO2 debris on support plate 3.3 3.2 3.4 3.3 Hot leg creep failure 3.5 3.4 3.5 3.4 Containment at 12 psig design pressure 3.5 3.4 3.5 3.4 Accumulator injection ends 3.5 3.4 3.5 3.4 Accumulator injection ends 3.5 3.4 3.5 3.4 First deflagration in containment 4.1 4 4.1 4 Containment fails at 60 psig 4.1 4 4.1 4 First UO2 debris in lower plenum 4.4 4.2 4.6 N/A UO2 debris quenched in lower plenum 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

Support plate failure 4.9 4.7 4.9 4.8 Diffuser plate failure 4.9 4.7 4.9 4.8 Lower support plate failure 5.6 5.5 5.7 5.5 RPV dryout 6.1 5.9 6.4 6 Lower head failure 6.8 6.9 7.4 6.7 Simulation end 3d 3d 3d 3d 241

C.2.6. Sequoyah, Case 4f Table C-11 Accident Sequence Event Timings for Sequoyah, Case 4f Large-break LOCA, No ECCS, TDAFW Operates in controlled mode, No Containment Sprays, Cavity Dry at Vessel Failure 60 80 Event 60 80 gwd/mtu gwd/mtu (Time in hours unless noted otherwise) gwd/mtu gwd/mtu HALEU HALEU Double guillotine break of single loop cold leg 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

Reactor trip 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

Containment fans begin circulation 0.5 [s] 0.7 [s] 0.4 [s] 0.4 [s]

RPV level at TAF 1.1 [s] 1.0 [s] 1.1 [s] 1.0 [s]

Containment at 12 psig design pressure 1.3 [s] 1.3 [s] 1.3 [s] 1.3 [s]

Accumulator injection starts 8.3 [s] 8.3 [s] 8.3 [s] 8.3 [s]

TDAFW injection starts 60.1 [s] 60.1 [s] 60.1 [s] 60.1 [s]

Accumulator injection ends 80.0 [s] 79.8 [s] 79.9 [s] 79.9 [s]

Start of cladding oxidation 0.17 0.16 0.16 0.16 First gap release 0.18 0.17 0.18 0.17 First UO2 debris formation 0.3 0.27 0.3 0.28 First UO2 debris on support plate 0.3 0.34 0.47 0.44 First UO2 debris in lower plenum 0.36 0.53 0.49 0.52 UO2 debris quenched in lower plenum 0.36 0.53 0.49 0.52 Support plate failure 0.83 1.1 0.75 1 Diffuser plate failure 0.9 1.1 0.86 1 Lower support plate failure 1.7 1.8 1.4 1.6 TDAFW floods, injection stops 2.1 2.1 2.1 2.1 RPV dryout 2.2 2.4 2.1 2.3 Lower head failure 3.3 3.1 3.3 3.6 First deflagration in containment 3.3 3.1 3.3 3.6 Containment reaches quasi-steady-state conditions 23.7 20.2 19.2 23.2 Containment fails at 60 psig N/A N/A N/A N/A Simulation end 3d 3d 3d 3d 242

C.2.7. Sequoyah, Case 4g Table C-12 Accident Sequence Event Timings for Sequoyah, Case 4g Small-break LOCA in Cold Leg, No ECCS, No AFW, No Containment Sprays, Cavity Dry at Vessel Failure Event 60 gwd/mtu 80 gwd/mtu (Time in hours unless noted 60 gwd/mtu 80 gwd/mtu HALEU HALEU otherwise) 1 in2 breach opens on single cold leg 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

Reactor trip 73.3 [s] 73.3 [s] 73.3 [s] 73.3 [s]

First secondary valve cycle 77.8 [s] 77.9 [s] 77.9 [s] 77.8 [s]

Single SG dryout 1.4 1.4 1.4 1.4 Triple SG dryout 1.4 1.4 1.4 1.4 RPV level at TAF 1.6 1.6 1.6 1.6 PRT rupture 1.9 1.9 1.9 1.9 First primary valve cycle 1.8 1.7 1.8 1.7 Start of cladding oxidation 2.5 2.5 2.5 2.5 First gap release 2.6 2.6 2.6 2.6 First UO2 debris formation 3 2.9 3 3 First UO2 debris on support plate 3 2.9 3 3 First UO2 debris in lower plenum 3.3 3.2 3.2 3.3 Support plate failure 3.3 3.3 3.3 3.3 Diffuser plate failure 3.3 3.3 3.3 3.3 Hot leg creep failure 3.5 3.5 3.5 3.5 First deflagration in containment 3.5 3.5 3.5 3.5 Containment at 12 psig design pressure 3.5 3.5 3.5 3.5 Accumulator injection starts 3.5 3.5 3.5 3.5 Accumulator injection ends 3.5 3.5 3.5 3.5 Lower support plate failure 4 4 4 4.1 RPV dryout 4.2 4.1 4.2 4.2 Lower head failure 4.8 4.7 4.8 4.7 Containment reaches quasi-steady-state conditions 15.2 12.2 12.4 12.6 Containment fails at 60 psig N/A N/A N/A N/A Simulation end 3d 3d 3d 3d 243

C.3. Peach Bottom C.3.1. Peach Bottom, Case 1a Table C-13 Accident Sequence Event Timings for Peach Bottom, Case 1a Short-term Station Blackout. One (1) SRV seizes in the open position before reactor vessel breach.

Containment failure initially occurs as a result of drywell shell melt-through or head flange leakage.

Event 60 gwd/mtu 80 gwd/mtu (Time in hours unless noted 60 gwd/mtu 80 gwd/mtu HALEU HALEU otherwise)

Station Blackout [Battery Failure] 0.0 [0.0 [s)) 0.0 [0.0 [s)) 0.0 [0.0 [s)) 0.0 [0.0 [s))

RPV Water Level At TAF 0.5 0.5 0.5 0.5 First Hydrogen Production 0.91 0.9 0.92 0.91 First Gap Release 0.94 0.93 0.94 0.93 First Channel Box Failure 1.2 1.2 1.2 1.2 First Fuel Rod Collapse [cell] 1.3 1.3 1.3 1.2 RPV Water Level At Bottom of Lower Core Plate 2.2 2.2 2.1 2.2 First Core Support Plate Failure 3.5 3.5 3.6 3.6 SRV Sticks Open 2.3 2.3 2.3 2.3 RPV Pressure Drops Below LPI Setpoint 4 3.9 4.1 2.5 Lower Plenum Dryout 4 4 4.1 4.2 Ring 1 CRGT Collapse 5.3 4.4 5.4 5.2 Ring 2 CRGT Collapse 5.7 4.6 6 5.3 Ring 3 CRGT Collapse 5.4 4.7 5.5 5.5 Ring 4 CRGT Collapse 5.9 4.7 6 5.5 Ring 5 CRGT Collapse 5.6 4.8 6.5 5.4 Lower Head Failure 7.3 7.3 7.5 7.6 Wetwell Failure (Above Water Level) Due To N/A N/A N/A N/A Overpressure Drywell Head Flange Leaks 8.1 8.1 8.3 8.2 Drywell Liner Melt-Through 7.5 7.6 7.7 7.8 Refueling Bay to Environment Blowout Panels Open 7.5 7.5 7.7 7.8 First Hydrogen Deflagration In Reactor Building 7.5 7.6 7.7 7.8 Equipment Access Door Opens To Environment Due 7.8 7.9 7.9 8 To Overpressure Reactor Building To Turbine Building Blowout Panel 7.8 7.9 7.9 8 Opens Total Iodine Release to Environment Exceeds 1% 7.6 7.9 7.8 8.3 Door To Environment In Southwest Stairwell Opens 7.8 7.9 7.9 8 Due To Overpressure Refueling Bay Roof Overpressure Failure N/A 7.9 N/A 8 Drywell Liner Penetration Shear Failure 10.7 10.8 10.8 10 Calculation Terminated 3d 3d 3d 3d 244

C.3.2. Peach Bottom, Case 1b Table C-14 Accident Sequence Event Timings for Peach Bottom, Case 1b Short-term Station Blackout. One (1) SRV seizes in the open position before reactor vessel breach.

Containment failure initially occurs as a result of drywell shell melt-through.

Event 60 gwd/mtu 80 gwd/mtu (Time in hours unless noted 60 gwd/mtu 80 gwd/mtu HALEU HALEU otherwise)

Station Blackout [Battery Failure] 0.0 [0.0 [s)) 0.0 [0.0 [s)) 0.0 [0.0 [s)) 0.0 [0.0 [s))

RPV Water Level At TAF 0.5 0.5 0.5 0.5 First Hydrogen Production 0.91 0.9 0.92 0.91 First Gap Release 0.94 0.93 0.94 0.93 First Channel Box Failure 1.2 1.2 1.2 1.2 First Fuel Rod Collapse [cell] 1.3 1.3 1.3 1.2 RPV Water Level At Bottom of Lower Core Plate 2.2 2.2 2.1 2.2 First Core Support Plate Failure 3.5 3.5 3.6 3.6 SRV Sticks Open 2.3 2.3 2.3 2.3 RPV Pressure Drops Below LPI Setpoint 4 3.9 4.1 2.5 Lower Plenum Dryout 4 4 4.1 4.2 Ring 1 CRGT Collapse 5.3 4.4 5.4 5.2 Ring 2 CRGT Collapse 5.7 4.6 6 5.3 Ring 3 CRGT Collapse 5.4 4.7 5.5 5.5 Ring 4 CRGT Collapse 5.9 4.7 6 5.5 Ring 5 CRGT Collapse 5.6 4.8 6.5 5.4 Lower Head Failure 7.3 7.3 7.5 7.6 Wetwell Failure (Above Water Level) Due To N/A N/A N/A N/A Overpressure Drywell Head Flange Leaks N/A N/A N/A N/A Drywell Liner Melt-Through 7.5 7.6 7.7 7.8 Refueling Bay to Environment Blowout Panels Open 7.5 7.5 7.7 7.8 First Hydrogen Deflagration In Reactor Building 7.5 7.6 7.7 7.8 Equipment Access Door Opens To Environment Due 7.8 7.9 7.9 8 To Overpressure Reactor Building To Turbine Building Blowout Panel 7.8 7.9 7.9 8 Opens Total Iodine Release to Environment Exceeds 1% 7.6 7.9 7.8 8.3 Door To Environment In Southwest Stairwell Opens 7.8 7.9 7.9 8 Due To Overpressure Refueling Bay Roof Overpressure Failure N/A 7.9 N/A 8 Drywell Liner Penetration Shear Failure 10.7 10.8 10.9 10.1 Calculation Terminated 3d 3d 3d 3d 245

C.3.3. Peach Bottom, Case 1c Table C-15 Accident Sequence Event Timings for Peach Bottom, Case 1c Short-term Station Blackout. One (1) SRV seizes in the open position before reactor vessel breach.

Containment failure initially occurs as a result of drywell head flange leakage. This case uses Basaltic concrete for the drywell floor.

Event 60 gwd/mtu 80 gwd/mtu (Time in hours unless noted 60 gwd/mtu 80 gwd/mtu HALEU HALEU otherwise)

Station Blackout [Battery Failure] 0.0 [0.0 [s)) 0.0 [0.0 [s)) 0.0 [0.0 [s)) 0.0 [0.0 [s))

RPV Water Level At TAF 0.5 0.5 0.5 0.5 First Hydrogen Production 0.91 0.9 0.92 0.91 First Gap Release 0.94 0.93 0.94 0.93 First Channel Box Failure 1.2 1.2 1.2 1.2 First Fuel Rod Collapse [cell] 1.3 1.3 1.3 1.2 RPV Water Level At Bottom of Lower Core Plate 2.2 2.2 2.1 2.2 First Core Support Plate Failure 3.5 3.5 3.6 3.6 SRV Sticks Open 2.3 2.3 2.3 2.3 RPV Pressure Drops Below LPI Setpoint 4 3.9 4.1 2.5 Lower Plenum Dryout 4 4 4.1 4.2 Ring 1 CRGT Collapse 5.3 4.4 5.4 5.2 Ring 2 CRGT Collapse 5.7 4.6 6 5.3 Ring 3 CRGT Collapse 5.4 4.7 5.5 5.5 Ring 4 CRGT Collapse 5.9 4.7 6 5.5 Ring 5 CRGT Collapse 5.6 4.8 6.5 5.4 Lower Head Failure 7.3 7.3 7.5 7.6 Wetwell Failure (Above Water Level) Due To N/A N/A N/A N/A Overpressure Drywell Head Flange Leaks 7.8 7.8 7.8 7.9 Drywell Liner Melt-Through N/A N/A N/A N/A Refueling Bay to Environment Blowout Panels Open 7.9 7.9 7.9 8 First Hydrogen Deflagration In Reactor Building 11.9 11.7 12.3 10.5 Equipment Access Door Opens To Environment Due 11.9 11.7 12.3 N/A To Overpressure Reactor Building To Turbine Building Blowout Panel 11.9 11.7 12.3 10.9 Opens Total Iodine Release to Environment Exceeds 1% 10.2 12.3 10.5 10.8 Door To Environment In Southwest Stairwell Opens N/A N/A N/A N/A Due To Overpressure Refueling Bay Roof Overpressure Failure 9 8.8 9.2 8.5 Drywell Liner Penetration Shear Failure 11.4 11.2 11.7 10.5 Calculation Terminated 3d 3d 3d 3d 246

C.3.4. Peach Bottom, Case 1d Table C-16 Accident Sequence Event Timings for Peach Bottom, Case 1d Short-term Station Blackout. All SRV operate as designed in safety relief mode, therefore, reactor vessel breach occurs at high pressure. Containment failure initially occurs as a result of drywell shell melt-through.

Event 60 gwd/mtu 80 gwd/mtu (Time in hours unless noted 60 gwd/mtu 80 gwd/mtu HALEU HALEU otherwise)

Station Blackout [Battery Failure] 0.0 [0.0 [s)) 0.0 [0.0 [s)) 0.0 [0.0 [s)) 0.0 [0.0 [s))

RPV Water Level At TAF 0.5 0.5 0.5 0.5 First Hydrogen Production 0.91 0.9 0.92 0.91 First Gap Release 0.94 0.93 0.94 0.93 First Channel Box Failure 1.2 1.2 1.2 1.2 First Fuel Rod Collapse [cell] 1.3 1.3 1.3 1.2 RPV Water Level At Bottom of Lower Core Plate 2.2 2.2 2.1 2.2 First Core Support Plate Failure 3.6 3.4 3.4 3.6 SRV Sticks Open N/A N/A N/A N/A RPV Pressure Drops Below LPI Setpoint 4.9 5.6 6 5.6 Lower Plenum Dryout 4.2 4 4 4.1 Ring 1 CRGT Collapse 5.4 4.9 6.2 4.7 Ring 2 CRGT Collapse 5.2 5.7 6.3 6.1 Ring 3 CRGT Collapse 5.4 5.6 6.3 5.8 Ring 4 CRGT Collapse 6.6 5.3 6.5 5.4 Ring 5 CRGT Collapse N/A 5.2 7 5.1 Lower Head Failure 4.9 5.6 6 5.6 Wetwell Failure (Above Water Level) Due To N/A N/A N/A N/A Overpressure Drywell Head Flange Leaks N/A N/A N/A N/A Drywell Liner Melt-Through 5.1 5.8 6.2 5.8 Refueling Bay to Environment Blowout Panels Open 5 5.7 6.1 5.8 First Hydrogen Deflagration In Reactor Building 5.1 5.8 6.2 5.8 Equipment Access Door Opens To Environment Due 5.4 6.1 6.5 6.1 To Overpressure Reactor Building To Turbine Building Blowout Panel 5.4 6.1 6.5 6.1 Opens Total Iodine Release to Environment Exceeds 1% 6.3 5.9 6.2 5.8 Door To Environment In Southwest Stairwell Opens 5.4 6.1 6.5 6.1 Due To Overpressure Refueling Bay Roof Overpressure Failure 5.4 6.1 6.5 6 Drywell Liner Penetration Shear Failure 9.5 9.7 10.1 9.7 Calculation Terminated 3d 3d 3d 3d 247

C.3.5. Peach Bottom, Case 2a Table C-17 Accident Sequence Event Timings for Peach Bottom, Case 2a Long-term Station Blackout. RCIC operates for 8 hrs then terminates due to battery exhaustion.

One (1) SRV seizes in the open position before reactor vessel breach. Containment failure initially occurs as a result of drywell shell melt-through.

Event 60 gwd/mtu 80 gwd/mtu (Time in hours unless noted 60 gwd/mtu 80 gwd/mtu HALEU HALEU otherwise)

Station Blackout [Battery Failure] 0.0 [8.0] 0.0 [8.0] 0.0 [8.0] 0.0 [8.0]

RPV Water Level At TAF 11.9 11.7 12.1 11.9 First Hydrogen Production 12.4 12.2 12.6 12.4 First Gap Release 12.4 12.2 12.6 12.4 First Channel Box Failure 12.7 12.4 13 12.7 First Fuel Rod Collapse [cell] 13.1 12.8 13.4 13 RPV Water Level At Bottom of Lower Core Plate 12.4 12.1 12.6 12.3 First Core Support Plate Failure 14.4 14 14.6 14.3 SRV Sticks Open 11.9 11.6 12.1 11.8 RPV Pressure Drops Below LPI Setpoint 15.8 15.4 12.3 12.1 Lower Plenum Dryout 15.8 15.4 17.2 16.3 Ring 1 CRGT Collapse 19.3 18.4 20.7 19.3 Ring 2 CRGT Collapse 19.5 18.5 20.8 19.7 Ring 3 CRGT Collapse 19.7 18.4 N/A 20.3 Ring 4 CRGT Collapse 20.8 N/A N/A N/A Ring 5 CRGT Collapse N/A N/A N/A N/A Lower Head Failure 20.5 19.7 21.8 20.8 Wetwell Failure (Above Water Level) Due To N/A N/A N/A N/A Overpressure Drywell Head Flange Leaks N/A N/A N/A N/A Drywell Liner Melt-Through 20.7 20 22 21 Refueling Bay to Environment Blowout Panels Open 20.7 19.9 22 21 First Hydrogen Deflagration In Reactor Building 20.7 19.9 22 21 Equipment Access Door Opens To Environment Due N/A 42.3 N/A N/A To Overpressure Reactor Building To Turbine Building Blowout Panel N/A N/A N/A N/A Opens Total Iodine Release to Environment Exceeds 1% 20.8 20 22.1 21 Door To Environment In Southwest Stairwell Opens 28.3 26.6 29.2 27.2 Due To Overpressure Refueling Bay Roof Overpressure Failure N/A N/A 22.2 N/A Drywell Liner Penetration Shear Failure 32.8 31.6 35.2 32.3 Calculation Terminated 3d 3d 3d 3d 248

C.3.6. Peach Bottom, Case 2b Table C-18 Accident Sequence Event Timings for Peach Bottom, Case 2b Long-term Station Blackout. RCIC operates for 8 hrs then terminates due to battery exhaustion.

One (1) SRV seizes in the open position before reactor vessel breach. Containment failure occurs as a result of drywell head flange leakage.

Event 60 gwd/mtu 80 gwd/mtu (Time in hours unless noted 60 gwd/mtu 80 gwd/mtu HALEU HALEU otherwise)

Station Blackout [Battery Failure] 0.0 [8.0] 0.0 [8.0] 0.0 [8.0] 0.0 [8.0]

RPV Water Level At TAF 11.9 11.7 12.1 11.9 First Hydrogen Production 12.4 12.2 12.6 12.4 First Gap Release 12.4 12.2 12.6 12.4 First Channel Box Failure 12.7 12.4 13 12.7 First Fuel Rod Collapse [cell] 13.1 12.8 13.4 13 RPV Water Level At Bottom of Lower Core Plate 12.4 12.1 12.6 12.3 First Core Support Plate Failure 14.4 14 14.6 14.3 SRV Sticks Open 11.9 11.6 12.1 11.8 RPV Pressure Drops Below LPI Setpoint 15.8 15.4 12.3 12.1 Lower Plenum Dryout 15.8 15.4 17.2 16.3 Ring 1 CRGT Collapse 19.3 18.4 20.7 19.3 Ring 2 CRGT Collapse 19.5 18.5 20.8 19.7 Ring 3 CRGT Collapse 19.7 18.4 N/A 20.3 Ring 4 CRGT Collapse 69.6 N/A N/A N/A Ring 5 CRGT Collapse N/A N/A N/A N/A Lower Head Failure 20.5 19.7 21.8 20.8 Wetwell Failure (Above Water Level) Due To N/A N/A N/A N/A Overpressure Drywell Head Flange Leaks 20.9 20 22 21 Drywell Liner Melt-Through N/A N/A N/A N/A Refueling Bay to Environment Blowout Panels Open 21 20.1 22.2 21.2 First Hydrogen Deflagration In Reactor Building 38.7 36.7 45.8 39.4 Equipment Access Door Opens To Environment Due To N/A N/A N/A N/A Overpressure Reactor Building To Turbine Building Blowout Panel N/A N/A N/A N/A Opens Total Iodine Release to Environment Exceeds 1% 22.9 21.3 24.2 22.8 Door To Environment In Southwest Stairwell Opens Due 46.1 41.8 52.5 44.4 To Overpressure Refueling Bay Roof Overpressure Failure N/A N/A 37.3 31.5 Drywell Liner Penetration Shear Failure 30.8 28.9 37.3 31.5 Calculation Terminated 3d 3d 3d 3d 249

C.3.7. Peach Bottom, Case 2c Table C-19 Accident Sequence Event Timings for Peach Bottom, Case 2c Long-term Station Blackout. RCIC operates for 8 hrs then terminates due to battery exhaustion.

One (1) SRV seizes in the open position before reactor vessel breach. Containment failure occurs as a result of static over-pressure (high-temperature failure mechanisms are neglected).

Event 60 gwd/mtu 80 gwd/mtu 60 gwd/mtu 80 gwd/mtu (Time in hours unless noted otherwise) HALEU HALEU Station Blackout [Battery Failure] 0.0 [8.0] 0.0 [8.0] 0.0 [8.0] 0.0 [8.0]

RPV Water Level At TAF 11.9 11.7 12.1 11.9 First Hydrogen Production 12.4 12.2 12.6 12.4 First Gap Release 12.4 12.2 12.6 12.4 First Channel Box Failure 12.7 12.4 13 12.7 First Fuel Rod Collapse [cell] 13.1 12.8 13.4 13 RPV Water Level At Bottom of Lower Core Plate 12.4 12.1 12.6 12.3 First Core Support Plate Failure 14.4 14 14.6 14.3 SRV Sticks Open 11.9 11.6 12.1 11.8 RPV Pressure Drops Below LPI Setpoint 15.8 15.4 12.3 12.1 Lower Plenum Dryout 15.8 15.4 17.2 16.3 Ring 1 CRGT Collapse 19.3 18.4 20.7 19.3 Ring 2 CRGT Collapse 19.5 18.5 20.8 19.7 Ring 3 CRGT Collapse 19.7 18.4 N/A 20.3 Ring 4 CRGT Collapse 40.6 N/A N/A N/A Ring 5 CRGT Collapse N/A N/A N/A N/A Lower Head Failure 20.5 19.7 21.8 20.8 Wetwell Failure (Above Water Level) Due To Overpressure 21.9 20.8 23.8 22.2 Drywell Head Flange Leaks N/A N/A N/A N/A Drywell Liner Melt-Through N/A N/A N/A N/A Refueling Bay to Environment Blowout Panels Open 21.9 20.8 23.7 22.2 First Hydrogen Deflagration In Reactor Building 21.9 20.8 23.7 22.2 Equipment Access Door Opens To Environment Due To 21.9 20.8 23.7 22.2 Overpressure Reactor Building To Turbine Building Blowout Panel Opens 21.9 20.8 23.7 22.2 Total Iodine Release to Environment Exceeds 1% 21.9 20.9 23.8 22.2 Door To Environment In Southwest Stairwell Opens Due N/A N/A N/A N/A To Overpressure Refueling Bay Roof Overpressure Failure 21.9 20.8 23.7 22.2 Drywell Liner Penetration Shear Failure N/A N/A N/A N/A Calculation Terminated 3d 3d 3d 3d 250

C.3.8. Peach Bottom, Case 3 Table C-20 Accident Sequence Event Timings for Peach Bottom, Case 3 Small Break LOCA in a Main Steam Line. All forms of coolant injection fail or are unavailable for coolant makeup. Containment failure occurs as a result of drywell head flange leakage.

Event 60 gwd/mtu 80 gwd/mtu (Time in hours unless noted 60 gwd/mtu 80 gwd/mtu HALEU HALEU otherwise)

Small Break LOCA 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

RPV Water Level At TAF 0.36 0.37 0.37 0.37 First Hydrogen Production 0.78 0.76 0.78 0.77 First Gap Release 0.81 0.79 0.81 0.8 First Channel Box Failure 1 1 1 1 First Fuel Rod Collapse [cell] 1 1 1 1 RPV Water Level At Bottom of Lower Core Plate 1 1 1 1 First Core Support Plate Failure 2.4 2.5 2.4 2.6 SRV Sticks Open N/A N/A N/A N/A RPV Pressure Drops Below LPI Setpoint 3.2 1.1 1.1 1.1 Lower Plenum Dryout 3.3 3.3 3.4 3.4 Ring 1 CRGT Collapse 5.1 5.2 5.5 5.4 Ring 2 CRGT Collapse 5.1 5.5 5.9 5.7 Ring 3 CRGT Collapse 5.6 5 6.4 5 Ring 4 CRGT Collapse 5.3 5.8 6.7 6 Ring 5 CRGT Collapse 5 5.2 6.3 5.7 Lower Head Failure 7.3 7.5 7.5 7.8 Wetwell Failure (Above Water Level) Due To N/A N/A N/A N/A Overpressure Drywell Head Flange Leaks 7.3 7.6 7.6 7.7 Drywell Liner Melt-Through N/A N/A N/A N/A Refueling Bay to Environment Blowout Panels 7.4 7.6 7.6 7.6 Open First Hydrogen Deflagration In Reactor Building 8 8 8.2 8.3 Equipment Access Door Opens To Environment 8 8 8.2 8.4 Due To Overpressure Reactor Building To Turbine Building Blowout N/A N/A N/A 8.4 Panel Opens Total Iodine Release to Environment Exceeds 1% 7.6 7.8 7.8 8 Door To Environment In Southwest Stairwell N/A N/A N/A 8.4 Opens Due To Overpressure Refueling Bay Roof Overpressure Failure 7.7 8 8 8.1 Drywell Liner Penetration Shear Failure 8 8.1 8.2 8.3 Calculation Terminated 3d 3d 3d 3d 251

C.3.9. Peach Bottom, Case 4 Table C-21 Accident Sequence Event Timings for Peach Bottom, Case 4 Large Break LOCA in Recirculation Suction. All forms of coolant injection fail or are unavailable for coolant makeup. Containment failure occurs as a result of drywell liner melt through.

60 80 Event 60 80 gwd/mtu gwd/mtu (Time in hours unless noted otherwise) gwd/mtu gwd/mtu HALEU HALEU Large Break LOCA 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

RPV Water Level At TAF 3.0 [s] 3.0 [s] 3.0 [s] 3.0 [s]

First Hydrogen Production 245.0 [s] 245.1 [s] 245.1 [s] 245.0 [s]

First Gap Release 277.0 [s] 275.1 [s] 273.4 [s] 274.3 [s]

First Channel Box Failure 0.25 0.25 0.25 0.25 First Fuel Rod Collapse [cell] 0.34 0.34 0.34 0.34 RPV Water Level At Bottom of Lower Core Plate 12.0 [s] 12.0 [s] 12.0 [s] 12.0 [s]

First Core Support Plate Failure 1.2 0.74 0.77 0.79 SRV Sticks Open N/A N/A N/A N/A RPV Pressure Drops Below LPI Setpoint 17.6 [s] 17.6 [s] 17.5 [s] 17.6 [s]

Lower Plenum Dryout 1.8 1.6 1.7 1.5 Ring 1 CRGT Collapse 2.6 2.9 3.1 2.8 Ring 2 CRGT Collapse 2.5 2.2 3 2.4 Ring 3 CRGT Collapse 2.5 2.2 2.2 1.9 Ring 4 CRGT Collapse 3.2 53 3.1 3.2 Ring 5 CRGT Collapse 3.1 N/A 4 3 Lower Head Failure 4.2 3.7 4.1 3.7 Wetwell Failure (Above Water Level) Due To Overpressure N/A N/A N/A N/A Drywell Head Flange Leaks N/A N/A N/A N/A Drywell Liner Melt-Through 4.4 4 4.3 3.9 Refueling Bay to Environment Blowout Panels Open 4.3 3.9 4.3 3.9 First Hydrogen Deflagration In Reactor Building 4.4 4 4.3 4 Equipment Access Door Opens To Environment Due To 5.8 5.4 4.5 4.1 Overpressure Reactor Building To Turbine Building Blowout Panel Opens 5.8 5.4 7.8 7.9 Total Iodine Release to Environment Exceeds 1% 4.4 4 4.3 4 Door To Environment In Southwest Stairwell Opens Due To 5.8 5.4 N/A N/A Overpressure Refueling Bay Roof Overpressure Failure 5.8 5.5 N/A N/A Drywell Liner Penetration Shear Failure 7.3 7.2 7.4 7 Calculation Terminated 3d 3d 3d 3d 252

C.4. Grand Gulf C.4.1. Grand Gulf, Case 5a Table C-22 Accident Sequence Event Timings for Grand Gulf, Case 5a Short-term Station Blackout. No coolant injection. One (1) SRV seizes in the open position before reactor vessel breach; low vessel pressure at vessel failure. Hydrogen combustion event at vessel failure causes containment failure.

Event 60 gwd/mtu 80 gwd/mtu (Time in hours unless noted 60 gwd/mtu 80 gwd/mtu HALEU HALEU otherwise)

Station Blackout 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

Battery Failure 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

RPV Water Level At TAF 0.69 0.67 0.69 0.68 First Hydrogen Production 1.4 1.4 1.4 1.4 First Gap Release 1.1 1.1 1.1 1.1 First Channel Box Failure 1.4 1.4 1.4 1.4 First Fuel Rod Collapse 1.6 1.5 1.6 1.5 First Core Support Plate Failure 3.6 3.3 3.6 3.2 SRV Sticks Open 3.6 3.3 3.7 3.3 Lower Plenum Dryout 4.6 4.5 4.8 4.2 Ring 1 CRGT Collapse 5.9 5.2 5.7 5.5 Ring 2 CRGT Collapse 6 5.1 6.1 6.2 Ring 3 CRGT Collapse 6.2 5.1 5.9 6.4 Ring 4 CRGT Collapse 6.3 5.1 6 5.8 Ring 5 CRGT Collapse 6.3 5.1 8.1 7.1 Lower Head Failure 8.7 8.4 8.5 7.9 First Hydrogen Burn In Reactor Pedestal 8.7 8.4 8.5 7.9 First Hydrogen Burn in Containment 8.7 8.4 8.5 7.9 Containment Overpressure Failure 8.7 8.4 8.5 7.9 Drywell Failure 8.7 8.4 8.5 7.9 Hydrogen Burns Cease in Drywell 8.8 8.4 8.6 7.9 Hydrogen Burns Cease in Containment 8.7 8.4 8.6 7.9 Total Iodine Release to Environment Exceeds 1% 9.6 9.6 9 7.9 Total In-Vessel H2 Generation 999.2 [kg] 1091.4 [kg] 1062.0 [kg] 1200.4 [kg]

Calculation Terminated 3d 3d 3d 3d 253

C.4.2. Grand Gulf, Case 5b Table C-23 Accident Sequence Event Timings for Grand Gulf, Case 5b Short-term Station Blackout. No coolant injection. High vessel pressure at vessel failure. Hydrogen combustion event at vessel failure causes containment failure.

Event 60 gwd/mtu 80 gwd/mtu (Time in hours unless noted 60 gwd/mtu 80 gwd/mtu HALEU HALEU otherwise)

Station Blackout 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

Battery Failure 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

RPV Water Level At TAF 0.69 0.67 0.69 0.68 First Hydrogen Production 1.4 1.4 1.4 1.4 First Gap Release 1.1 1.1 1.1 1.1 First Channel Box Failure 1.4 1.4 1.4 1.4 First Fuel Rod Collapse 1.6 1.5 1.6 1.5 First Core Support Plate Failure 3.6 3.3 3.6 3.2 SRV Sticks Open N/A N/A N/A N/A Lower Plenum Dryout 4.7 4.9 4.7 4.4 Ring 1 CRGT Collapse 8.7 6.7 7.5 6.8 Ring 2 CRGT Collapse 8.8 6.7 7.5 6.8 Ring 3 CRGT Collapse 8.9 6.8 9.4 6.7 Ring 4 CRGT Collapse 9 7.5 N/A 4.4 Ring 5 CRGT Collapse 8.9 7.4 N/A 8.2 Lower Head Failure 8.4 6 6.8 5.7 First Hydrogen Burn In Reactor Pedestal 8.3 6 N/A N/A First Hydrogen Burn in Containment 8.3 6 6.8 5.7 Containment Overpressure Failure 8.3 6 6.8 5.7 Drywell Failure N/A N/A N/A N/A Hydrogen Burns Cease in Drywell 8.3 6 N/A N/A Hydrogen Burns Cease in Containment 13.3 12.4 13.6 12.8 Total Iodine Release to Environment Exceeds 1% 8.4 6.6 7.4 12.3 Total In-Vessel H2 Generation 987.1 [kg] 1156.4 [kg] 973.5 [kg] 1122.8 [kg]

Calculation Terminated 3d 3d 3d 3d 254

C.4.3. Grand Gulf, Case 5c Table C-24 Accident Sequence Event Timings for Grand Gulf, Case 5c Short-term Station Blackout. No coolant injection. One (1) SRV seizes in the open position before reactor vessel breach; low vessel pressure at vessel failure. Late overpressure containment failure.

Event 60 gwd/mtu 80 gwd/mtu (Time in hours unless noted 60 gwd/mtu 80 gwd/mtu HALEU HALEU otherwise)

Station Blackout 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

Battery Failure 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

RPV Water Level At TAF 0.69 0.67 0.69 0.68 First Hydrogen Production 1.4 1.4 1.4 1.4 First Gap Release 1.1 1.1 1.1 1.1 First Channel Box Failure 1.4 1.4 1.4 1.4 First Fuel Rod Collapse 1.6 1.5 1.6 1.5 First Core Support Plate Failure 3.6 3.3 3.6 3.2 SRV Sticks Open 3.6 3.3 3.7 3.3 Lower Plenum Dryout 4.6 4.5 4.8 4.2 Ring 1 CRGT Collapse 5.9 5.2 5.7 5.5 Ring 2 CRGT Collapse 6 5.1 6.1 6.2 Ring 3 CRGT Collapse 6.2 5.1 5.9 6.4 Ring 4 CRGT Collapse 6.3 5.1 6 5.8 Ring 5 CRGT Collapse 6.3 5.1 8.1 7.1 Lower Head Failure 8.7 8.4 8.5 7.9 Containment Overpressure Failure 55.3 50 58.1 52.4 Total Iodine Release to Environment Exceeds 1% 55.3 50 58.1 52.4 Total In-Vessel H2 Generation 990.5 [kg] 1085.5 [kg] 1051.5 [kg] 1210.9 [kg]

Calculation Terminated 3d 3d 3d 3d 255

C.4.4. Grand Gulf, Case 6a Table C-25 Accident Sequence Event Timings for Grand Gulf, Case 6a Long-term Station Blackout. Battery power available for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. RCIC injection. One (1) SRV seizes in the open position before reactor vessel breach; low vessel pressure at vessel failure. Hydrogen combustion event at vessel failure can cause containment failure.

Event 60 gwd/mtu 80 gwd/mtu (Time in hours unless noted 60 gwd/mtu 80 gwd/mtu HALEU HALEU otherwise)

Station Blackout 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

Battery Failure 8 8 8 8 RCIC Suction Transferred to Suppression Pool 1.3 1.3 1.3 1.3 RPV Water Level At TAF 9.5 10 9.8 9.4 First Hydrogen Production 10.9 11.3 11.3 10.8 First Gap Release 10.4 10.8 10.7 10.3 First Channel Box Failure 10.9 11.4 11.3 10.8 First Fuel Rod Collapse 11.3 11.7 11.7 11.3 First Core Support Plate Failure 13.2 13.6 13.7 13.3 SRV Sticks Open 13.7 13.7 13.3 13.3 Lower Plenum Dryout 14.8 15.6 14.9 14.5 Ring 1 CRGT Collapse 16.8 18 17.6 17.6 Ring 2 CRGT Collapse 18.3 18.6 19.3 18.7 Ring 3 CRGT Collapse N/A 19.9 N/A 19.3 Ring 4 CRGT Collapse N/A N/A N/A 20.4 Ring 5 CRGT Collapse N/A N/A N/A 20.8 Lower Head Failure 18.9 19.9 19.4 19.9 First Hydrogen Burn In Reactor Pedestal 18.9 19.9 19.3 19.9 First Hydrogen Burn in Containment 18.9 19.9 19.3 19.9 Containment Overpressure Failure 18.9 30.7 19.4 31.3 Hydrogen Burns Cease in Drywell 18.9 19.9 19.4 19.9 Hydrogen Burns Cease in Containment 18.9 21.4 19.4 20.1 Total Iodine Release to Environment Exceeds 1% 18.9 30.7 24.1 31.3 Total In-Vessel H2 Generation 1223.0 [kg] 1430.7 [kg] 1387.2 [kg] 1213.3 [kg]

Calculation Terminated 3d 3d 3d 3d 256

C.4.5. Grand Gulf, Case 6b Table C-26 Accident Sequence Event Timings for Grand Gulf, Case 6b Long-term Station Blackout. Battery power available for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. RCIC injection. One (1) SRV seizes in the open position before reactor vessel breach; low vessel pressure at vessel failure. Late overpressure containment failure.

Event 60 gwd/mtu 80 gwd/mtu (Time in hours unless noted 60 gwd/mtu 80 gwd/mtu HALEU HALEU otherwise)

Station Blackout 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

Battery Failure 8 8 8 8 RCIC Suction Transferred to Suppression Pool 1.3 1.3 1.3 1.3 RPV Water Level At TAF 9.5 10 9.8 9.4 First Hydrogen Production 10.9 11.3 11.3 10.8 First Gap Release 10.4 10.8 10.7 10.3 First Channel Box Failure 11 11.4 11.3 10.8 First Fuel Rod Collapse 11.4 11.7 11.7 11.3 First Core Support Plate Failure 13.6 13.6 13.7 13.3 SRV Sticks Open 13.7 13.7 13.3 13.3 Lower Plenum Dryout 14.7 15.6 14.9 14.5 Ring 1 CRGT Collapse 17.5 18 17.6 17.6 Ring 2 CRGT Collapse 18.1 18.6 19.3 18.7 Ring 3 CRGT Collapse 19.4 19.9 N/A 19.3 Ring 4 CRGT Collapse N/A N/A N/A 20.4 Ring 5 CRGT Collapse N/A N/A N/A 20.7 Lower Head Failure 19.4 19.9 19.4 19.9 Containment Overpressure Failure 52.7 46.9 58.6 30 Total Iodine Release to Environment Exceeds 1% 52.8 46.9 58.6 30 Total In-Vessel H2 Generation 1478.7 [kg] 1424.3 [kg] 1423.7 [kg] 1207.8 [kg]

Calculation Terminated 3d 3d 3d 3d 257

C.4.6. Grand Gulf, Case 7 Table C-27 Accident Sequence Event Timings for Grand Gulf, Case 7 Accident transient without SCRAM. Spurious MSIV closure. RCIC injection. ADS actuates; low vessel pressure at vessel failure. Containment fails prior to core damage.

Event 80 60 gwd/mtu 80 gwd/mtu (Time in hours unless noted 60 gwd/mtu gwd/mtu HALEU HALEU otherwise)

Spurious MSIV Closure 0.0 [s] 0.0 [s] 0.0 [s] 0.0 [s]

RCIC Suction Transferred to Suppression Pool 113.6 [s] 113.4 [s] 117.5 [s] 114.1 [s]

Recirculation Pump Trip 5.4 [s] 5.6 [s] 5.5 [s] 5.6 [s]

RHR Initiated in Suppression Pool Cooling Mode 0.91 0.91 0.9 0.9 RCIC Isolation Due to High Turbine Exhaust Pressure 0.38 0.38 0.38 0.38 RCIC Isolation Due to Low Steam Line Pressure 1.1 1.1 1.1 1.1 ADS Actuation 119.9 [s] 119.9 [s] 120.1 [s] 120.0 [s]

RPV Water Level At TAF 0.24 0.24 0.24 0.24 First Hydrogen Production 1.1 1.1 1.1 1.1 First Gap Release 1 1 1 1 First Channel Box Failure 1.1 1.1 1.1 1.1 First Fuel Rod Collapse 1.2 1.3 1.3 1.2 First Core Support Plate Failure 3.3 3.3 3.3 2.9 Lower Plenum Dryout 4.5 4.5 4.5 4.4 Ring 1 CRGT Collapse 6.8 7.1 6.6 5.7 Ring 2 CRGT Collapse 6.8 6.7 7.1 5.6 Ring 3 CRGT Collapse 6.3 5 7.3 5.6 Ring 4 CRGT Collapse 4.4 6.6 4.3 5.1 Ring 5 CRGT Collapse 6.4 6 7.3 5.7 Lower Head Failure 8.6 8 8.4 7.8 Containment Overpressure Failure 0.91 0.91 0.9 0.9 Total Iodine Release to Environment Exceeds 1% 1.6 1.9 1.6 2.3 Total In-Vessel H2 Generation 641.8 [kg] 650.2 [kg] 652.7 [kg] 683.7 [kg]

Calculation Terminated 3d 3d 3d 3d 258

C.4.7. Grand Gulf, Case 8 Table C-28 Accident Sequence Event Timings for Grand Gulf, Case 8 Recirculation suction line LOCA. RCIC injection. Low vessel pressure at vessel failure.

Late overpressure containment failure.

Event 60 gwd/mtu 80 gwd/mtu (Time in hours unless noted 60 gwd/mtu 80 gwd/mtu HALEU HALEU otherwise)

RPV Water Level At TAF 8.5 [s] 8.5 [s] 8.6 [s] 8.5 [s]

First Hydrogen Production 0.16 0.16 0.15 0.16 First Gap Release 315.7 [s] 306.9 [s] 303.7 [s] 314.4 [s]

First Channel Box Failure 0.25 0.25 0.24 0.25 First Fuel Rod Collapse 0.37 0.36 0.36 0.37 First Core Support Plate Failure 0.6 0.77 1 0.74 Lower Plenum Dryout 1.6 1.5 1.9 1.7 Ring 1 CRGT Collapse 2.5 2.7 2.8 2.7 Ring 2 CRGT Collapse 3.3 2.9 3.2 3.2 Ring 3 CRGT Collapse 3.7 3.1 N/A 3.9 Ring 4 CRGT Collapse 3 2.8 N/A 3 Ring 5 CRGT Collapse 3.5 3.7 N/A 3.6 Lower Head Failure 3.7 4 3.9 3.9 First Hydrogen Burn in Containment 3.7 4 3.8 3.8 Containment Overpressure Failure 46 43.7 54.1 46.3 Hydrogen Burns Cease in Containment 6.6 6.9 7.5 7 Total In-Vessel H2 Generation 499.8 [kg] 404.6 [kg] 669.1 [kg] 423.1 [kg]

Calculation Terminated 3d 3d 3d 3d 259

APPENDIX D. SUPPORTING SOURCE TERM TABLES D.1. Low-Enriched Uranium Table D-1 Comparison of BWR LEU phase durations and release fractions with uncertainties for all LEU core variations (60 GWd/MTU LEU, 80 GWd/MTU LEU).

Gap Release Early In-vessel Late In-vessel Ex-vessel Phase Duration 0.69 (6%) 6.7 (5%) 46.8 (4%) 3.2 (31%)

Noble Gases 0.021 (32%) 0.95 (1%) 0.005 (24%) 0.008 (38%)

Halogens 0.006 (87%) 0.71 (3%) 0.17 (13%) 0.014 (24%)

Alkali Metals 0.005 (91%) 0.33 (12%) 0.021 (16%) 0.007 (39%)

Te Group 0.004 (107%) 0.57 (7%) 0.19 (22%) 0.003 (14%)

Ba/Sr Group 0.0008 (107%) 0.005 (12%) 0.003 (8%) 0.041 (24%)

Ru Group <1.0E-6 0.006 (6%) 5.5E-05 (24%) <1.0E-6 Mo Group 2.2E-05 (157%) 0.13 (10%) 0.002 (32%) 1.1E-05 (191%)

Lanthanides <1.0E-6 <1.0E-6 <1.0E-6 4.1E-05 (27%)

Ce Group <1.0E-6 <1.0E-6 0.0 (0.0%) 0.003 (43%)

Table D-2 Comparison of PWR LEU phase durations and release fractions with uncertainties for all LEU core variations (60 GWd/MTU LEU, 80 GWd/MTU LEU).

Gap Release Early In-vessel Late In-vessel Ex-vessel Phase Duration 1.3 (7%) 3.9 (9%) 25.4 (46%) 1.9 (27%)

Noble Gases 0.025 (11%) 0.94 (1%) 0.011 (41%) 0.015 (29%)

Halogens 0.006 (18%) 0.58 (17%) 0.041 (53%) 0.019 (15%)

Alkali Metals 0.002 (59%) 0.50 (28%) 0.015 (39%) 0.013 (28%)

Te Group 0.006 (15%) 0.54 (19%) 0.025 (46%) 0.005 (12%)

Ba/Sr Group 0.001 (17%) 0.002 (13%) 0.0002 (104%) 0.012 (37%)

Ru Group <1.0E-6 0.008 (23%) 5.9E-05 (41%) <1.0E-6 Mo Group 1.6E-05 (104%) 0.15 (28%) 0.002 (28%) 0.003 (78%)

Lanthanides <1.0E-6 <1.0E-6 <1.0E-6 1.7E-05 (57%)

Ce Group <1.0E-6 <1.0E-6 <1.0E-6 0.0007 (92%)

260

Table D-3 Comparison of BWR 60 GWd/MTU LEU phase durations and release fractions with uncertainties.

Gap Release Early In-vessel Late In-vessel Ex-vessel Phase Duration 0.67 (11%) 6.7 (10%) 44.8 (10%) 2.9 (39%)

Noble Gases 0.016 (45%) 0.94 (1%) 0.007 (27%) 0.014 (47%)

Halogens 0.008 (92%) 0.71 (4%) 0.15 (23%) 0.020 (28%)

Alkali Metals 0.008 (88%) 0.31 (19%) 0.017 (22%) 0.013 (48%)

Te Group 0.006 (98%) 0.54 (7%) 0.21 (30%) 0.004 (15%)

Ba/Sr Group 0.001 (100%) 0.005 (10%) 0.003 (12%) 0.038 (26%)

Ru Group <1.0E-6 0.006 (12%) 4.6E-05 (44%) <1.0E-6 Mo Group 7.9E-05 (66%) 0.12 (22%) 0.002 (23%) 9.5E-06 (262%)

Lanthanides <1.0E-6 <1.0E-6 <1.0E-6 3.5E-05 (26%)

Ce Group <1.0E-6 <1.0E-6 0.0 (0.0%) 0.003 (44%)

Table D-4 Comparison of PWR 60 GWd/MTU LEU phase durations and release fractions with uncertainties.

Gap Release Early In-vessel Late In-vessel Ex-vessel Phase Duration 1.3 (13%) 4.0 (10%) 27.0 (48%) 2.2 (37%)

Noble Gases 0.024 (19%) 0.93 (3%) 0.012 (54%) 0.014 (36%)

Halogens 0.007 (31%) 0.57 (21%) 0.050 (124%) 0.018 (20%)

Alkali Metals 0.003 (77%) 0.50 (31%) 0.018 (47%) 0.012 (34%)

Te Group 0.006 (28%) 0.55 (24%) 0.029 (89%) 0.005 (22%)

Ba/Sr Group 0.001 (32%) 0.002 (20%) 0.0005 (65%) 0.010 (73%)

Ru Group <1.0E-6 0.007 (28%) 5.6E-05 (36%) <1.0E-6 Mo Group 1.6E-05 (144%) 0.15 (34%) 0.002 (39%) 0.003 (78%)

Lanthanides <1.0E-6 <1.0E-6 <1.0E-6 1.7E-05 (173%)

Ce Group <1.0E-6 <1.0E-6 <1.0E-6 0.0007 (258%)

261

Table D-5 Comparison of BWR 80 GWd/MTU LEU phase durations and release fractions with uncertainties.

Gap Release Early In-vessel Late In-vessel Ex-vessel Phase Duration 0.64 (10%) 6.3 (10%) 46.7 (6%) 3.6 (56%)

Noble Gases 0.016 (42%) 0.96 (2%) 0.004 (71%) 0.006 (25%)

Halogens 0.006 (70%) 0.71 (4%) 0.17 (23%) 0.012 (15%)

Alkali Metals 0.006 (73%) 0.31 (13%) 0.026 (20%) 0.005 (22%)

Te Group 0.005 (83%) 0.59 (10%) 0.17 (28%) 0.003 (24%)

Ba/Sr Group 0.0009 (84%) 0.005 (21%) 0.002 (13%) 0.047 (50%)

Ru Group <1.0E-6 0.006 (9%) 6.8E-05 (27%) <1.0E-6 Mo Group 0.0001 (72%) 0.12 (14%) 0.004 (49%) 5.4E-05 (681%)

Lanthanides <1.0E-6 <1.0E-6 <1.0E-6 5.4E-05 (44%)

Ce Group <1.0E-6 <1.0E-6 0.0 (0.0%) 0.005 (58%)

Table D-6 Comparison of PWR 80 GWd/MTU LEU phase durations and release fractions with uncertainties.

Gap Release Early In-vessel Late In-vessel Ex-vessel Phase Duration 1.3 (13%) 3.8 (15%) 21.1 (65%) 1.8 (47%)

Noble Gases 0.024 (16%) 0.92 (3%) 0.011 (50%) 0.016 (39%)

Halogens 0.006 (31%) 0.56 (21%) 0.039 (60%) 0.019 (17%)

Alkali Metals 0.002 (73%) 0.50 (30%) 0.014 (49%) 0.014 (37%)

Te Group 0.005 (30%) 0.54 (22%) 0.024 (52%) 0.005 (14%)

Ba/Sr Group 0.0009 (36%) 0.002 (21%) 0.0008 (45%) 0.013 (40%)

Ru Group <1.0E-6 0.008 (26%) 6.9E-05 (72%) <1.0E-6 Mo Group 2.1E-05 (161%) 0.15 (31%) 0.002 (45%) 0.002 (101%)

Lanthanides <1.0E-6 <1.0E-6 <1.0E-6 2.2E-05 (65%)

Ce Group <1.0E-6 <1.0E-6 1.7E-05 (98%) 0.001 (82%)

262

D.2. High-Assay Low-Enriched Uranium Table D-7 Comparison of BWR HALEU phase durations and release fractions with uncertainties for all HALEU core variations (60 GWd/MTU HALEU, 80 GWd/MTU HALEU).

Gap Release Early In-vessel Late In-vessel Ex-vessel Phase Duration 0.70 (5%) 6.6 (3%) 42.5 (5%) 3.0 (34%)

Noble Gases 0.022 (33%) 0.94 (1%) 0.005 (75%) 0.012 (14%)

Halogens 0.006 (91%) 0.72 (4%) 0.13 (21%) 0.019 (12%)

Alkali Metals 0.005 (95%) 0.30 (14%) 0.021 (15%) 0.011 (12%)

Te Group 0.004 (109%) 0.53 (9%) 0.19 (19%) 0.003 (14%)

Ba/Sr Group 0.0008 (109%) 0.005 (7%) 0.002 (20%) 0.034 (31%)

Ru Group <1.0E-6 0.006 (4%) 0.0001 (32%) <1.0E-6 Mo Group 3.6E-05 (127%) 0.12 (9%) 0.002 (35%) 4.2E-05 (121%)

Lanthanides <1.0E-6 <1.0E-6 <1.0E-6 2.9E-05 (47%)

Ce Group <1.0E-6 <1.0E-6 0.0 (0.0%) 0.002 (87%)

Table D-8 Comparison of PWR HALEU phase durations and release fractions with uncertainties for all HALEU core variations (60 GWd/MTU HALEU, 80 GWd/MTU HALEU).

Gap Release Early In-vessel Late In-vessel Ex-vessel Phase Duration 1.3 (7%) 4.1 (8%) 21.7 (56%) 2.0 (19%)

Noble Gases 0.027 (11%) 0.92 (1%) 0.010 (24%) 0.018 (12%)

Halogens 0.007 (22%) 0.58 (18%) 0.024 (66%) 0.021 (12%)

Alkali Metals 0.004 (45%) 0.50 (29%) 0.013 (19%) 0.016 (12%)

Te Group 0.006 (20%) 0.55 (20%) 0.014 (47%) 0.005 (11%)

Ba/Sr Group 0.001 (20%) 0.002 (15%) 7.5E-05 (112%) 0.010 (19%)

Ru Group <1.0E-6 0.008 (24%) 4.6E-05 (44%) <1.0E-6 Mo Group 3.4E-05 (123%) 0.15 (29%) 0.002 (35%) 0.002 (129%)

Lanthanides <1.0E-6 <1.0E-6 <1.0E-6 1.3E-05 (29%)

Ce Group <1.0E-6 <1.0E-6 <1.0E-6 0.0005 (54%)

263

Table D-9 Comparison of BWR 60 GWd/MTU HALEU phase durations and release fractions with uncertainties.

Gap Release Early In-vessel Late In-vessel Ex-vessel Phase Duration 0.67 (11%) 6.5 (7%) 41.0 (9%) 3.0 (63%)

Noble Gases 0.017 (53%) 0.94 (1%) 0.007 (81%) 0.013 (16%)

Halogens 0.012 (58%) 0.76 (5%) 0.12 (33%) 0.020 (18%)

Alkali Metals 0.011 (63%) 0.31 (11%) 0.019 (22%) 0.011 (15%)

Te Group 0.009 (80%) 0.56 (12%) 0.19 (19%) 0.004 (20%)

Ba/Sr Group 0.002 (86%) 0.006 (5%) 0.002 (35%) 0.033 (37%)

Ru Group <1.0E-6 0.005 (9%) 0.0001 (47%) <1.0E-6 Mo Group 2.9E-05 (169%) 0.11 (12%) 0.002 (29%) 2.4E-05 (184%)

Lanthanides <1.0E-6 <1.0E-6 <1.0E-6 2.7E-05 (42%)

Ce Group <1.0E-6 <1.0E-6 0.0 (0.0%) 0.002 (83%)

Table D-10 Comparison of PWR 60 GWd/MTU HALEU phase durations and release fractions with uncertainties.

Gap Release Early In-vessel Late In-vessel Ex-vessel Phase Duration 1.3 (13%) 4.2 (10%) 14.1 (82%) 1.9 (35%)

Noble Gases 0.027 (16%) 0.91 (2%) 0.009 (43%) 0.018 (23%)

Halogens 0.008 (33%) 0.57 (23%) 0.026 (100%) 0.020 (20%)

Alkali Metals 0.003 (61%) 0.50 (32%) 0.014 (29%) 0.016 (22%)

Te Group 0.007 (33%) 0.55 (24%) 0.016 (63%) 0.005 (27%)

Ba/Sr Group 0.001 (34%) 0.002 (21%) 0.0004 (49%) 0.009 (43%)

Ru Group <1.0E-6 0.008 (28%) 3.7E-05 (52%) <1.0E-6 Mo Group 3.9E-05 (160%) 0.15 (32%) 0.002 (38%) 0.001 (154%)

Lanthanides <1.0E-6 <1.0E-6 <1.0E-6 1.2E-05 (46%)

Ce Group <1.0E-6 <1.0E-6 <1.0E-6 0.0005 (78%)

264

Table D-11 Comparison of BWR 80 GWd/MTU HALEU phase durations and release fractions with uncertainties.

Gap Release Early In-vessel Late In-vessel Ex-vessel Phase Duration 0.65 (11%) 6.3 (8%) 42.4 (6%) 3.0 (38%)

Noble Gases 0.016 (44%) 0.94 (1%) 0.004 (94%) 0.010 (47%)

Halogens 0.007 (65%) 0.71 (3%) 0.14 (18%) 0.017 (21%)

Alkali Metals 0.006 (69%) 0.26 (17%) 0.024 (29%) 0.009 (49%)

Te Group 0.005 (78%) 0.49 (10%) 0.16 (39%) 0.003 (19%)

Ba/Sr Group 0.0010 (81%) 0.005 (13%) 0.002 (23%) 0.032 (50%)

Ru Group <1.0E-6 0.006 (9%) 0.0001 (39%) <1.0E-6 Mo Group 2.7E-05 (151%) 0.12 (14%) 0.003 (73%) 6.4E-05 (87%)

Lanthanides <1.0E-6 <1.0E-6 <1.0E-6 3.6E-05 (79%)

Ce Group <1.0E-6 <1.0E-6 0.0 (0.0%) 0.003 (100%)

Table D-12 Comparison of PWR 80 GWd/MTU HALEU phase durations and release fractions with uncertainties.

Gap Release Early In-vessel Late In-vessel Ex-vessel Phase Duration 1.3 (13%) 3.8 (13%) 28.2 (48%) 2.1 (22%)

Noble Gases 0.027 (16%) 0.92 (2%) 0.011 (31%) 0.018 (14%)

Halogens 0.007 (36%) 0.58 (20%) 0.025 (73%) 0.021 (13%)

Alkali Metals 0.003 (61%) 0.51 (29%) 0.010 (45%) 0.016 (15%)

Te Group 0.005 (36%) 0.56 (22%) 0.014 (63%) 0.005 (7%)

Ba/Sr Group 0.0009 (37%) 0.002 (23%) 7.2E-05 (132%) 0.010 (28%)

Ru Group <1.0E-6 0.008 (30%) 4.3E-05 (56%) <1.0E-6 Mo Group 4.6E-05 (172%) 0.15 (32%) 0.002 (54%) 0.002 (104%)

Lanthanides <1.0E-6 <1.0E-6 <1.0E-6 1.3E-05 (75%)

Ce Group <1.0E-6 <1.0E-6 <1.0E-6 0.0006 (81%)

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