ML20116J655
| ML20116J655 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 10/31/1992 |
| From: | Miller L, Payne A, Shaffer C SANDIA NATIONAL LABORATORIES, SCIENCE & ENGINEERING ASSOCIATES, INC. |
| To: | NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES) |
| References | |
| CON-FIN-A-1393 NUREG-CR-5305, NUREG-CR-5305-V03, NUREG-CR-5305-V3, SAND90-2765, NUDOCS 9211160279 | |
| Download: ML20116J655 (577) | |
Text
{{#Wiki_filter:j - - . , , . . . . , , , . . NUREG/CR-5305 SAND 90-2765 Vol. 3 Integrated Risk Assessment -?o.r
- he LaSale Unit 2 Nuclear Power Plant Phenomenology and Risk Uncertainty Evaluation Program (PRUEP)
A MELCOR Code Calculations i Prepared by C J. Shaffer L A. Miller A. C. Payne, Jr. Sandia National I aboratories Operated by Sandia Corporation Prepared for U.S. Nuclear Regulatory Commission 4 hDk D K O O 0374 p PDR
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NUREG/CR-5305 SAND 90-2765 Vol. 3 RX Integrated Risk Assessment for the LaSalle Unit 2 Nuclear Power Plant Phenomenology and Risk Uncertainty Evaluation Program (PRUEP) MELCOR Code Calculations
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l Manuscrip. Completed: October 1992 Date Published: October 1992 Prepared by C. J. Shafferi,1. A. Miller, A. C. Payne, .'r. Sandia Nr.tional 1.aboratories Albuquerque, NM 87185 Prepared for Division of Safety Issue Resolution Office of Nuclear Regulatoy Research ' U.S. Nuclear Regulatory Commission Washington, DC 20555 - NRC FIN A1393 1 Science and Engineering Associates SEA Plaza 6100 Upton Blvd., N.E. Albuquerque, NM 87110
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ABSTRACT A Level III Probabilistic Risk ' Assessment (PRA) has been performed for LaSalle Unit 2 under the Risk Methods Integration and . Evaluation Program-(RMIEP) and the Phenomene. logy and Risk Uncertainty Evaluation Program (PRUEP). This report documents the phenomenole.gical calculations and sources of. uncertainty in the calculations perforn ed with MELCOR in support of the Level II portion of the PRA. These calculations are an integral, part of the Level II analysis since they provide quantitative input to the Accident Progression Event Tree (APET) and Source Term Model (LASSOR). Howeve_r, the uncertainty associated with the code results must be considered in the use of the results. The MELCOR calculations performed include ' four integrated calculations: (1)_ a high-pressure _ shcrt-term station blackout. (2) a low-pressure short-term station blackout, (3) an intermediate-term station blackout, and (4) a long term station blackout. Several sensitivity studies investigating the effect- of variations in containment failure size and location, as well as hydrogen ignition concentration are also documented. ]
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n1 i 1 l TABLE OF CONTENTS 4 1 i Section Pace ! ABSTRACT................................. ......................... iii-l - LIST OF FIGURES...... ...................,......................... vili LIST OF TABLES...... ..,.......... ................................ xvii ACKNOWLEDGMENTS........... ... .................................... xxi FOREWORD........... .......... .. ... .............. . ............ xxiii 1 LIST OF ACRONYMS.........................,.......................... xxvii i 1.0 Introduction............. . .................................. 1-1 !4 1.1 References.......... ... .... ........ .................. 1-1 3. 4 l' l 2.0 MELCOR Code De sc ri p t ion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 1 f' 2.1 Introduction........ .................................... 2-1 i
- 2.2 Thermal-Hydraulic Models......................... ....... 2-1 i 2.3 Core Degradation and Debris Interaction Models........... 2-3 i
J 2.4 Co mbus t i o n Mode l s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 4 i > ! 2.5 Engineered Safety Feature Modela.... . ...... ..... ..... 2-4 i l 2.6 Decay Heat Power Models............................ ..... 2-4 I' l 2.7 Radionuclide Release and Transport Models . ............. 2-4 2.8 References; ..... ... ........ ........... .. ............ 2-7 n
- 3.0 Input Model Description.......... ............................ 3-1 3.1 Thermal and Hydrodynamics Models......................... 3-1 I
} !. 3.1.1 Primary System...................................... 3-1 !- 3.1.2. Containment Building................................ 3-11 { 3.1.3 Reactor _ Building...................... ........... . 3-15' 1 4 i ! 3.2 Reactor Core Mode 1................................'........-3-19 l
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! -3.3 Core / Concrete Interaction Model..........................;3 ' 3.4 Combustion Model; .... ....... .......... . . . . . . . . . . . . . 3 ! 1 i
.3.5 Decay Heat Power Model............;.~. . ............ .... 3 s
! 3.6 Radionuclide Transport.Models... ........................ 3-35 i-I l f
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TABLE OF CONTENTS (Continued) Section Pa r.e 3,7 References. .. . . . , , , ,. ............. ...... 3-37 4.0 Sequence Description... ......... .. ..... ........... . .. .4-1 4.1 High-Pressure Short-Term Station Blackout (T63)....... .. 4-1 4.1.1 Base Case.... .. .. .. .. .. .... .. ..... ... . 4-1 4.1.2 Sensitivity Cases. .. . . ... ... ............... .4-2 4.2 Low-Pressure Short Term Station Blackout (T100). ...... . 4-4 4.2.1 Base Case....................... ............... ... 4-4 4.2.2 Sensitivity Case.... .. ..... ...................... 4-5 4.3 Intermediate-Term Station Blackout (T62)............. .
,4 5 4.4 Long-Term Station Blackout (T24).... . ... . .. .......,4-6 4.5 References., , , , , .......... ............ ... ....,4-7 5.0 Results... . . . ... . . ,,, ........ . ......... 5-1 5.1 High Pressure Short-Term Station Blackout...... . .. ,5-1 5.1.1 Base Case. .... ... . ......... .. ...... . .5-1 5.1.1.1 Overview. . ... . ... ... . . . ...... .. 5-1 5.1.1.2 Primary System Hydrodynamics........ ... ... .54 5.1.1.3 Core Meltdown.. .. ....... .... .. .. .. .5-7 5.1.1.4 Containment Hydrodynamics..................... 5-11 5.1.1.5 Reactor Building Hydrodynamics. ... . . .. 5-17 5.1.1.6 Radionuclide Transport...... . ............... 5-18 5.1.2 Containment Failure Sensitivity Calculations...... 5-25 5.1.2.1 Break Flow Resistance Sensitivity..... ... ... 5-25 5.1.2.2 Break Location Sensitivity........... ... . .5-25 5.1.2.3 Combustion Ignition Limit Parameters.......... 5-26 5.2 Low-Pressure Short-Term Station Blackout........ ...... . 5-127 5.2.1 Base Case. , , ....... . ................. ,. . .. 5-127 5.2.1.1 Overview. ... . .......... .................. 5-127 5.2.1.2 Primary System hydrodynamics.. ...... ... ... 5-128 5.2.1.3 Core Meltdown. . ....... .................. 5-128 5.2.1.4 Containment Hydrodynamics.... ...... ......... 5-129 5.2.1.5 Reactor Building Hydrodyaamics., ..... ... ... 5-132 5.2.1.6 Radionuclide Transport..... . ....... . . .. . 5-132 5.2.2 Pedestal Wall Failure Sensitivity Calculation.... ... 5-133
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TABLE OF CONTENTS _(Concluded)- - Section Engg 5.3 Intermediate-Term Station Blackout Calculation........ ... 5-180 5.3.1 Overview .... ...................... .............-.5-180 5.3.2 Primary System Hydrodynamics...................... 5-181 5.3.3 Core Meltdown.. ................. ................ 5-181 5.3.4 Containment Hydrodynamics,... ............ ....... 5-182 5.3.5 Reactor Building Hydrodynamics.................... 5-183 5.3.6 Radionuclide Transport......................... . .5-184 5.4 Long-Term Station Blackout Calculation............... ... 5-228 5.4.1 Overview .................................... .... 5-228 5.4.2 Primary System Hydrodynamics............ ......... 5-229 5.4.3 Core Meltdown...........................- ........ 5-229 5.4.4 Containment Hydrodynamics........... . . . .... 5-230 5.4.5 Reactor Building Hydrodynamics.................... 5-232 5.4.6 Radionuclide Transport............ ............... 5-233 6.0 Discussion of Uncertainty...,,......... ............ .. .. 6-1 6.1 Introduction.... ........................................ 6 1 6.2 The Codes Used'in the LaSalle PRA........................ 6-2 6.3 Sources of Uncertainty in Thermal-Hydraulic Calculations. 6 6.3.1 Initial and Boundary Conditions..................... 6-4 6.3.2 Plant Mode 1........................... ............. 6-5 6.3.3 Physics Model., ....... ............................ 6-6 6.3.4 Parame te r Unc e r ta in ty . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 6 6.3.5 Coding Errors............................... .., ... 6-7 6.4 Discussion of Specific Selected Uncertainties in the- _LaSalle MELCOR Calculations........... .................67- ,
-6.5 Conclusions............................................ . 6-19 6.6 References.... ..... ............................... .... 6-20 7.0 Conclusions.. .................................. ............. 7-1 7.1 References................ . .... .......-................ 7 Appendix A: MELCOR input listing...................................A-1
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- _ 1 J
- LIST OF FICURES a
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!: Figure Title Page 3.1-1 LaSalle HELCOR Primary System........................... 3 3 i 3.1-2
- LaSalle MELCOR Containment System......... .............. 3-12 i
L 3.1-3 LaSalle MELCOR Reactor Building......... ............... 3-17 1 5 3.2-1 LaSalle MELCOR Core Map.........,....................... 3-22 i e n , 3.5-1 Comparison of Decay Heat Power Models,........,......... 3 34 !. 5.1-1 Core Channel Pressures,,..................... .......... 5 l ! 5,1 2 Reactor Vessel Water Levels................. ... ....... 5-56 1 l- 5.1-3 -Reactor Vessel-integrated Steam Flow Rates.............. 5-57 f i l 5,1-4 Reactor Vecsel Integrated Hydrogen Flow Rates........... 5 ;
- 5.1-5 Reactor Vessel Vapor Temperatures....................... 5-59 5.1-6 Reactor Vessel' Structure Temperatures..............,.... 5-60 I
- - 5,1-7 Cladding Failure for Cap Fission Product Releases....... 5-61 i
- 5.1-8 Inner Ring. Fuel Cladding Temperatures.,................. 5-62 1
m l 5.1-9 Core Cell 109 Tempe ratures . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 6 3 4 5 j 5.1-10 Core-Cell _109. Masses.........,.......................... 5 64: p - 5.1-11 Lower Core-Temperatures of Inner Ring................... 5 . . . a
- 5.1-12. . Core Cell 106 Masses.,.... ............................. 5-66 I
5.1-13 ' Core Plate Temperatures.............. ..................;5 j. 5.1 Inner Lower Head Temperature............................. 5-68 i . . j - 5.1-15 Lower Plenum Cell: 101. Masses................... ........ 5 - 5.1 Lower Head Penetration-Temperatures..................... 5-70 i: 5.1-17 Lower Plenum Fuel Masses........ ....................... 5-71 i-4-
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LIST OF FIGURES (Continued) Eigurg Title ......................................... .........Eaga 5.1-18 Total Core Masses.................... .................. 5-72 5.1-19 Core Exit Vapor Temperatures........ . ................. 5-73 5.1-20 Containment Pressures........ ..... ........... ........ 5-74 5.1-21 Drywell Temperatures.... ............................... 5 75 5.1-22 Wetwell Temperatures. ................................. 5-76 5.1 23 Pedestal Temperatures................................... 5 77 5.1-24 Containment Condensation................................ 5-78 5.1 25 Drywell and Pedestal Vater Masses....................... 5 79 5.1 26 Suppression Pool Mass and Temperature....... ........... 5-80 5.1-27 Suppre s sion Pool Wa ter Leve1s . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 81 5.1-28 D ywell Pedestal Debris Masses and Temperatures......... 5-82 5.1 29 We twell Pede s tal Debris Mas s e s . . . . . . . . . . . . . . . . . . . . . . . . . . ,5 - 8 3 5.1-30 Vetwell Pedsstal Debris Temperatures.................... 5-64 5.1-31 Wetwell Pedestal Melt Elevations........................ 5-85 5.1 32 Maximum Concrete Penetration Distances.................. 5-86 5.1-33 Axisymmetric Cavity Shape.. .... ....................... 5-87 5.1-34 Core / Concrete Interaction Cas Production. . . . . . . . . . . . . . . . 5-88 5.1-35 Hydrogen Production..................................... 5-89 5.1-36 Conc re te De gas sing Mas ses . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 90 5.1 37 Containment Hydrogen Distribution....................... 5-91 5.1-38 Containment Vent Exit Cas Flow Rates.................... 5-92 5.1-39 Containment Vacuum Breaker Flow Rate ... ............... 5-93 5.1-40 Wetwell Pedestal to Drywell Pedestal Flow Rates......... 5-94.
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i l l LIST OF FIGURES (Continued) l 11 cure Title ..................................................,Eagg 5.1-41 Containment Intedrated Hydrogen Flow Rates.............. 5-95 j 5.1-42 We twe ll Le akage Flow Ra te s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 9 6
- 5.1-43 Reactor / Turbine Building Atmosphere Temperatures........ 5-97 5.1-44 Upper Unit 2 Mole Fractions............................. 5-98 l'
- - 5.1 45 Reactor / Turbine Building _ Hydrogen Mole Tractions........ 5-99' i
5.1 46 Reactor Building Flow Rates............................. 5-100 j 5.1-47 Reactor Building Integrated Flow Rates............. .... 5-101 1 5.1-48 In and Ex Vessel Releases of Cerium from Fuel........... 5-102 s 5.1'-49 Non-Radioactive Ex-Vatcel Reiceree...................... 5 103 5.1-50 Primary System Retention Fractions...................... 5-104 . 5.1-51 Xenon Mass Distribution................................. 5-105 1 5.1-52 Cesium Mass Distribution................................ 5-106 i 5.1-53 Xenon Source Term to Environment......... ,,,........... 5-107 i i 5.1-54 Cerium, Barium, Tellurium, and Cesium Iodide Source l Terms.......... ................., ................... 5-108 i 5.1-55 Iodine and Ruthenium Source Terms..... ................. 5-109 4 5.1-56 Molybdenum, Cerium, and Lanth'num Source Terms.......... 5-110-i 5.1-57 Uranium and Tin Source Terms............................ 5-111 i 5.1-58 Cadmium Source Term:.................................... 5-112 5.-1-59 Reactor /Turbint Building Decontamination Factors. ...... 5-113 5.1 Reactor' Building Aerosol Size Distribution.............. 5-114 i, ) 5.1-61 Radioactive Decay Power Distribution.................... 5-115 5.1-62 Containment Depressurizations........................... 5-116 5.1-63 ~ Xenon Source Terms for Break Area Study................. 5 117 o
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.l i I LIST OF FIGURES _(Continued) 1 . Figure Titic ....................................................Eagg i k 5.1-64 Barium Source Terms fcr Break Area S tudy. . . . . . . . . . . . . . . . 5 118 i
- 5.1-65 Xenon Source Terms for - Large Break Location Study . . . . . . 5 119 4
l 5.1-66 Cesium Source Terms for Large Break Location Study...... 5-120 5.1-67 Barium Source Terms for Large Break Location Study. . . . . . 5-121 a 5.1-68 Cesium Source Terms for Small Break Location Study...... 5-122 I 5.1 69 Barium Source Terms for Small Break Location Study...... 5-123 l- 5.1-70 Barium Source Terms for Small Leak Location Study. . . . . . . 5-124 1 l 5.1-71 Barium Source Terms for Burn Effect Study...............,5-125 i . - [ 5.1-72 Xenon Source Terms for Hydrogen Ignition Limit Study ... 5-126-i j 5.2-1 Core Channel Pressures.................................. 5-140-i 5.2-2 Reactor Vessel Water Levels. . . ......................... 5-141 1,
- 5.2-3 Reactor Vessel Integrated Steam Flow Rates.......... . . . 5-142 <
j. f 5.2-4 Reactor Vessel Integrated Hydrogen Flow Rates........... 5-143 i 5.2 5 Total Core Masses..... . . .................... . . . . . . . . . . .-5-144- . 5.2-6 Containment Pressures............. ............ . . . . . . . . 5-145 1 ._ 5.2-7 Drywell Temperatures. . . . . . ............. ........., . . . 5-146 i 5-2-8
. Wetwell Temperatures..... . ............................ 5-147.
i 5.2-9 Pedestal Temperatures............. .. ............ ..... 5-148-4 1-5.2 Containment Condensation................ ............... 5-149 +
- 5 .' 2 - 11 Drywell and Pedestal Water Masses....................... 5-150 5.2-12 Suppression Pool Mass _and' Temperature................... 5-151~
15.2-13 Suppression Pool Water Levels............ 5-152 ]. , ~5.2-14 -Drywell Pedestal-Debris Masses and Temperatures... .. . . . 5 153 4-2 a
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LIST OF F1 CURES (Continued) Firure Iltig ... ...............................................Engg 5.2-15 Wetwell Pedestal Debris Masses................... .... . 5 154 5.2-16 Wetwell Pedestal Debris Temperatures.................... 5-155 5.2-17 Maximua Concrete Penetration Distances........... ...... 5 156 5.2-16 Core /;oncrete Interaction Gas Production................ 5 157 5.2-19 Hydrogen Production.......................,.... ........ 5-158 5.2-20 Concrete Degassing Masses... ........................... 5-159 5.2-21 Containment Hydrogen Distribution....................... 5 160 5.2-22 Cintainment Integrated Hydrogen Flow Rates...........,.. 5-161 5.2 23 We twe ll Rup tu re F1 ow . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 16 2 5.2-24 Reactor / Turbine Building Atmosphere Temperatures........ 5-163 5.2-25 Upper Unit 2 Mole Fractions............................. 5-164 5.2-26 Reactor / Turbine Building Hydrogen Mole Fractions........ 5-165 5.2-27 In and Ex-Vessel Releases of Cerium from-Fuel........... 5-166 5.2-28 Non-Radioactive Ex Vessel Releases...................... 5-167 ' 5.2-29 Primary System Retention Fractions...................... 5-166 5.2-30 Xenon Mass Distribution................................. 5-169 i-5.2-31 Cesium Mass Distribution........ ....................... 5-170 l ~ 5.2-32 Xenon Source Term to Environment........................ 5-171 l 5.2-33 Cesium, Barium,-and Cesium Iodide Source Terms.......... 5-172' I. I-L 5.2-34 Iodine Source Term...................................... 5-173-i 5.2-35 Tellurium and Uranium Source Terms.......... ........... 5-174 I 5.2 Ruthenium Source Terms. ... .. ......................... 5-175 5.2-37 Molybdenum, Cerium, and Lanthanum Source Terms.....,.... 5-176 i
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s 1 LIST OF FIGURES (Continued) t 2 E Figure Title .................... .. ...........................Eagn i 5.2-38 Cadmium and Tin Source Terms...................... ..... 5 177 i i e 5.2-39 Reactor / Turbine Building Decontamination Factors........ 5-178 b j 5.2 Wetwell Pressures. ... .. ............................ 5-179 1 1 i A 5.3-1 Core Channel Pressures.................... ............. 5-189 L L 5.3-2 Reactor Vessel Water Levels.. .......................... 5-190 j 5.3-3 Reactor Vessel .*ntegrated Hydrogen Flow Rates........... 5-191 4 1 5.3-4 Total Core Masses.... .................................. 5-192 i 5.3-5 Containment Pressures........... ....................... 5 193 l~ l 5.3-6 Drywell Temperatures.................. ................. 5-194 1 ) 5.3-7 Wetwell Temperatures.................................... 5-195 4 l 5.3-8 Pedestal Temperatures.... .............................. 5 196 i 5.3-9 Containment Condensation.......................... ..... 5-197 l 5.3-10 Drywell and Pedestal Water Masses....................... 5-198~ i 5.3-11 Suppression Pool Mass and Temperature................... 5-199 s 1 5.3-12 Suppression Pool Water Levels................ ........ ..5-200 j 5.3-13 Drywell Pedestal Debris Masses and Temperatures......... 5-201 2 5.3-14 Wetwell' Pedestal ~ Debris Masses............. ............ 5-202 2 5.3-15 Vetwell Pedestal Debris Temperatures.....................5-203
- 5.3-16 Maximum Concrete Penetration Dis aces... .............. 5-204
-5.3-17 Core / Concrete Interaction Gas Production... ............ 5-205 5.3-18 Hydrogen Production.. . .... ........... ........... ... 5-206_
5.-3-19 Concrete Degassing Masses... ........................... 5-207
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5.3-20 Containment-Hydrogen Distribution................... ... 5-208 o e
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LIST 0F-FIGURES (Continued) Figure Title ...................................................Eaga 5.3-21 Containment Integrated Hydrogen Flow Rates.............. 5-209 5.3-22 Wetwell Rupture Flow............................ ....... 5-210 5.3-23 Reactor / Turbine Building Atmosphere Temperatures........ 5-211. 5.3 24 Upper Unit 2 Mole Fractions............................. 5-212 5.3-25 Reactor / Turbine Building Hydrogen Mole Fractions........ 5-213 5.3 26 In and Ex-Vessel Relenses cf Cerium from Fuel........... 5-214-5.3-27 Non Radioactive Ex-Vessel Releases. .... ............... 5-215 5.3-28 Primary System Retention Fr' actions...................... 5-216-5.3-29 Xenon Mass Distribution................................. 5-217 5.3 30 Cesium Mass Distribution................................ 51218 5.3-31 Xenon Source Term to Environment........ ............... 5-219 5.3-32 Cesium, Tellurium, and Molybdenum Source Terms.......... 5-220 5 . 3 - 7.3 Barium Source Term........ ............................. 5-221 5.3-34 Iodine Source Term...................................... 5-222 5.3-35 Ruthenium Source Term,....... ....... .................. 5 223 5.3-36 Lanthanum, Uranium, and Cesium Iodide Source Terms...... 5-224 l 5.3 37 Cerium and Cadmium Source Terms......................... 5-225 5.3-38 Tin Source Term. .... ...... ........................... 5-226.
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5.3-39 Reactor / Turbine Building Deconta.aination Factors........ 5-227 1 r l 5.4-1 Core 1 Channel Pressures......................s ........... 5-238 5.4-2' Reactor Vessel' Vater Leve1s............................. 5-239; 5.4-3 Reactor Vessel Integrated Hydrogen Flow Rates........... 5-240 l- ! '5.4-4 Total Core Masses....................................... 5-241 L
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LIST OF FIGURES (Continued) Ficure Title .................... .................. ............Eggn. 5.4 5 Containment Pressures,....... ....... .................. 5-242 5.4-6 Drywell Temperatures............. ....................... 5-243 5.4-7 Wetwell Temperatures.................... ............... 5-244 5,4-8 Pedestal Temperatures...... ............................ 5-245 5.4-9 Containment Condensation................................ 5 246 5.4-10 Drywell and Pedestal Water Masses.......................,5-247 5.4 11 Suppression-Pool Mass and Temperature................... 5-248-5.4 12 Suppression Pool Water Levels................ 4 ......... 5-249 5.4-13 Drywell Pedestal Debris Masses and Temperatures......... 5-250 5.4-14 Wetwell Pedestal Debris Masses.............. ........... 5-251 5.4-15 Mu twell Pedestal Debris Tempuratures . . . . . . . . . .......... 5-252 5.4-16 Maximum Concrete Penetration Distancesi............. .. .5-253. 5.4-17 Core / Concrete Interaction Cas Production................ 5-254 5,4-18 Hydrogen Production.......... .... ................. ... 5-255 5.4-19 Concrete Degassing Masses.......... . .. ............... 5-256-5.4 Containment Hydrogen Distribution......... ............. 5-257 5.4-21 Containment Integrated Hydrogen Flow Rates.............. 5-258 5.4-22 Wetwell Rupture Flow...... ......................... ... 5-259-5.4-23 Reactor / Turbine Building Atmosphere Temperatures.........5-260.- 5.4-24 Upper Unit 2 Mole Fractions... ............. .... ...... 5-261-5.4 Reactor / Turbine Building Hydrogen Mole' Fractions........ 5-262
- 5.4-26 In and Ex-Vessel Releases of Cerium from Fuel........... 5-263' 5.4-27 Non Radioactive Ex-Vessel Releases...................... 5-264 .
5.4-28' Primary System Retention Fractions...................... 5-265 -
-xV-y__ , _ __ . . , , - - - -
LIST OF FIGURES (Concluded) Ficure Titic .., ...............................................Eagt 5.4-29 Xenon Mass Distribution................................... 5-266 . 5.4-30 Cesium Mass Distribution.. .................,........... 5 267 5.4-31 Xenon Source Term to Environment........................ 5-268 5.4-32 Cesium and Barium Source Terms.................. ....... 5-269 5.4-33 Iodine Source Term...................................... 5-270 5.4-34 Tellurium, Lanthanum, and Uranium Source Terms.......... 5 271 5.4-35 Molybdenum and Cesium Iodide Source Terms................ 5e272 5.4-36 Cerium Source Term....... .....................-........ 5-273 5.4-37 Cadmium and Tin Source Terms............................ 5-274 5.4-38 Ruthenium Source Term................................... 5-275 5.4-39 Reactor / Turbine Building Decontamination Factors....... .5-276
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b LIST OF TABLES 1 Table Title East 2.7-1 Default Radionuclide C1 asses............................ 26 3.1-1 LaSalle Model Control Volume Input Data............ .. .32 3.1 2 LaSalle Model Flow Path Input Data...... ...... , ,.... .3-5' ] 3.1-3 LaSalle Model Heat Structure Input Data.. .............. 3-9 3.1-4 LaSalle Concre te Dega s s ing Da ta . . . . . . . . . . . . . . . . . . . . . . . . . 3 - 16 3.2-1 Radial Volume and Power Distribution............... .... 3-20 3.2-2 Axial Power Distribution.. .......... .... ........,,. 3;20 3.2-3 Core Model Cell Component Masses.................. ..... 3 21 3.2-4 Core Model Cell Surface Areas.......................... .3 24 3.2-5 Core Model Flow Areas.............. ......... .......... 3 3.2 6 Lower Head Structure Input.............................. 3-29 3.2-7 Lower Head Penetration Input. ............... .. ....... 3-29 3.3-1 LaSalla Concrete Composition...................... . ... 3 32 3.3 2 LaSalle Concrete Properties... .. .. .... ............. 3-32 3.4-1 Combustion Parameters............. ..... .. ......... 3-32 3.6-1 CORSOR Release Coefficients.................... ...... . 3 36 4.1-1 Sensitivity Cases for the High-Pressure Short-Term-Station Blackout Sequence.. . ....... .... ....... . . . 4 -'3
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5.1 1 High-Pressure Short-Term Station Blackout Events........ 5-28 5.1 High-Pressure Short-Term Station Blackout Core Meltdown-Temperatures (K). . . ........ . ........... .5-29
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LIST OF TABLES (Continued)- Table Title Eggg 5.1-31 Reactor / Turbine _ Building Peak Pressures and Temperatures for the-High-Pressure Short-Term Station Blackout ......... ............................ 5-32 5.1-4 Summary of Radionuclide Releases and Transport.......... 5-33 5.1-5 Distribution for the Xenon Class................. ...... 5-34 5.1 6 Distribution for the Cesium Class....... ............... 5-34 5.1-7 Distribution for the Barium.... ....................... 15 35 5.1-8' Distribution for'the Iodine Class....................... 5-35 3.1-9 Distribution for the Tellurium Class............. ...... 5-36 5.1-10 Distribution for the Ruthenium Class., ................. 5-36 5.1-11 Distribution for the Molybdenum Class................... 5-37 5.1-12 Distribution-for.the Cerium class. .............. ...... 5-37 5.1-13 Distribution for the Lanthanum Class......... .......,. 5-38 5.1-14 Distribution for the Uranium Class..........c. ........ 5-38 5.1-15 Distribution for the Cadmium Class...................... 5-39
- 5.1-16 Distribution for.the Tin C1 ass.......................... 5 39 5.1-17 Distribution for the Cesium Iodide Cless...... ......... 5-40 5.1-18. -Early and Late Transport of Fission Products from:the RCS....................... ......................... .. 5-41 5.1-19 Containment Distribution of Fission Products ........... 5-42 5.1-20 Airborno,= Pool,land Deposition; Distribution Fractions of Fission ProductsLfor!the ReactorLCooling System at End of Calculation........................, ........... 5-43 5.1-21 Airborne, Pool, and Deposition Distribution _ Fractions of Fission Products forzthe Containment at Endtof
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Calculation. ............... .......................... 5-43
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LIST OF TABi.ES (Continued) Table Title Engg 5.1-22 Airborne, Pool, and Deposition Distribution Fractions of Fission Products for the Reactor / Turbine Building. ._ at End of Calculation.................................. 5-44 1 5.1-23 Airborne, Pool, and Deposition Distribution Fractionc of Fission Products for the Entire Plant at End of Calculation........................................ ... 5-44 5.1 24 Fractional System Distribution for Deposited Fission Products at the Enu of Calculation. . . . . . . . . . . . . 5-45 5.1-25 Fission Product Vapor to Aerosol Mass Ratios in the Atmospheres at End of Calculation........... ... 5-45 5.1-26 High-Pressure STSB Sensitivity Depressurizations. ..... 5-46 5.1-27 High-Pressure STSB Break Area Sensitivity Results for a Wetwell Break and 5% Ignition Limit........... .. 5-47 5.1-28 'iign Pressure STSB Break Area Sensitivity Results: for a Drywell Break and 5% Ignition Limit....... .... 5 5.1-29 High-Pressure STSB Break Area Sensitivity Results: for a Drywell Head Break and 5% Ignition Limit......... 5-49 5.1-30 High-Pressure STSB Break Location Sensitivity Results: for a Break Area of .0924m2 and 5% Ignition Limits . . . . . 5-50 5.1-31 High-Pressure-STSB Break Location l Sensitivity:Results: for a Break Area' of .00924m2 and 5% Ignition Limits . . . . 5-51 5.1-32 High-Pressure STSB Break Location-Sensitivity Results; for a Break Area of .000799m2 and 5% Ignition Limits . . . 5-52 5.1-33 High-Pressure STSB Ignition Limit Sensitivity Results:
~
for a Break ' Area of .0924m2 and both Vetvelliand Drywell= Head Breaks............ ....................... 5-53 a 5.1 3h High-Pressure STSB Ignition Limit Sensitivity Results: for a-Wetwell Wall Break With an Area of? .000799m2. . . . . 5-54 , 5.2-1 Low Pressure Short-Term Station Blackout Events.........,5-1351
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... .- . -_ - .. . ~ _ _ - . -
LIST OF TABLES .(Concluded) Table Title Pace 5.2-2 Reactor / Turbine Building Peak Pressures and Temperatures for Low-Pressurc Short Term Station Blackout..................................... ... ..... 5-136 5.2 s Summary of Radionuclide Releases and Transport.for Low-Pressure Short Term Station Blackout......... ..... 5 136 5.2 4 Fission Product Distribution (kg), Classes 1-4.......... 5-137
~
5.2-5 Fission Product Distribution (kg) , Classes .5-8. . . . . . . . . 5-137: 5.2-6 Fission Product Distribution (kg), Classes 9-12......... 5-138 5.2-7 Fission Product Distribution.(kg), Class 16............. 5-138-5.2-8 Containme t Distribution of Fission Products............ 5-139- : 5.2-9 Environmental _ kelease Comparison. . . . . . . . . . . . . . . . . . . . . . . . 5-139 5.3-1 Intermediate-Term Station Blackout Events................ 5-185 5.3-2 Summary of Radionuclide Releases and Transport.......... 5-186 , 5.3-3 Fission Product Distribution _(kg),_ Classes- 1-4..........,5-186 5,3-4 Fission Product Distribution'(kg), ClassesLS-8.......... 5-187 5.3-5 Fission Product Distribution (kg), Classes 9-12......... 5-187 5.3-6 Fission Product Distribution (kg). Class 16............. 5-188 5;3-7 Containment Distribution of Fission Products... ........ 5-188-5.4-1 -Long-Term Station Blackout Events.... .................. 5-234 5.4 Summary of Radionuclide Releases and Transport.......... 5-235 5.4-3 Fission Product Distribution (kg), Classes 1-4..... ... 5 235. 5.4-4 Fission' Product Distribution (kg), ClassesL5 8. ........ 5-236 5'4-5
. _ Fission Product Distribution (kg), Classes 9-12., ...... 5-236
_5.4-6 Fission Product Distribution (kg), Class 16............. 5-237-(- 5.417 Containment Distribution of Fission Products............ 5-237 L t -xx-
ACKNOULEDGMENTS-i The authors' acknowledge the contribution's made by _ Susan Dingman . and Christi Leigh in ene developraent of the input for these calculations. Randall Cole, Edward Boucheron, Randall-Summers, and Mike Carmel provided MELCOR code support and review of the calculations. 1
-l 1
1 l I
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FOREWORD LaSalle Unit 2 Level III Probabilistic Risk Assessment In recent: years, applications of Probabilistic Risk Assessment (PRA). to nuclear power _ plants have experienced increasing acceptance and use, particularly in addressing regulatory issues. Although progress on the-PRA front has been impressive, the usage of PRA methods and insights to-address increasingly _ broader regulatory issues hes resulted in the need - i for continued improvement in and expansion of PRA methods to support the needs of the Nuclear Regulatory Commission (NRC). j i Before any new PRA methods-can be considered suitable for toutine_use in- 1 I the regulatory arena, they need to be integrated into the overall framework of a PRA, appropriate interfaces defined, and the - utility _of-the methods evaluated. The LaSalle' Unit 2 Level III PRA, described in this c.nd associated reports, integrates new methods and new applications of previous methods into a PRA framework that provides for - this l integration and evaluation. It helps lay the bases for both the routine ! use of the methods and the preparation of procedures that_will provide guiLnce for future FRAs used in addressing regulatory issues, These new methods, once integrated into the framework of a-PRA and evaluated, lead l to .* more complete PRA analysis , a better _ understanding of E the uncertainties in PRA results, and broader insights;into the importance of ple.nt design and operational characteristics to public risk.' In order to satisfy the needs described above, the LaSalle Unit 2, Level. III PRA addresses the following broad objectives:
- 1. To develop and apply' methods to integrate internal, external, and-dependent failure risk methods to -- achieve - greater efficiency, consistency, and completeness in the conduct of risk assessments; 2, To evaluate PRA technology developments and- for: .ilate improved PRA procedures;
- 3. To identify, evaluate, and effectively display the ' uncertainties.:
in PRA risk predictions that stem from- limitations- in plant modeling, PRA methods, data, or physical ' processes ~ that occur during the evolution of a severe accident; 4 To conduct a PRA on a BWR 5, Mark r II nuclear power ': plant, ascertain the - plant's dominant accident sequences, evaluate the core 'and containment response to accidents , calculate _ the consequences of the accidents, and assess overall risk; _and finally
- 5. To formulate the results in such a manner as to allow the PRA to be easily updated and to allow testing of future improvements in methodology, data, and the treatment of phenomena.
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The LaSalle Unit 2 PRA was performed - for the NRC by Sandia_ National Laboratories (SNL) with substantial help from Commonwealth Edison (CECO)- Because of the size and scope of'the PRA, various
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and its contracton related prograns were set up to conduct dif ferent - aspects of the analysis. Additionally, existing programs had _ tasks added to perform some analyses for the LaSalle PRA. The responsibility for overall - direction of the PRA was assigned to the Risk Methods- Integration and Evaluation Program (RMIFP). RMIEP was specifically responsible for-all aspects of the Level I analysis (i.e. , - the core damage analysis). The Phenomenology and Risk Uncertainty Evaluation Program _(PRUEP) - was responsible for the Level II/III analysis (i.e., accident pro 5ression, source term, consequence analyses, and risk integration). Other programs provided support in various areas or performed some of the subanalyses. l These programs include the Seismic - Safety Margins Research Program (SSMRP) at Lawrence Livermore National-Laboratory (LLNL), which performed the seismic analysis; the Integrated Dependent-Failure Analysis Program, which developed methods and analyzed data for dependent failure modeling; the MELCOR Program, which modified the MELCOR code _ in response to the PRA's modeling-needs; the Fire Research Program, which performed _the fire analysis; the PRA Methods Development Program, which developed some of the new methods used in the PRA; and the Data Programs, which provided new and updated data for BWR plants similar to LaSalle. CECO provided plant design and operational information and reviewed many of the analysis results. The LaSalle PRA was begun before thr NUREG 1150 analysis and the LaSalle program has supplied the NUREG-1150 program with simplified location analysis methods for integrated analysis of external events, insights on possible subtle interactions that come from the very detailed system models used in the LaSalle PRA, core vulnerable sequence resolution methods, methods for handling atti propagating statistical uncertainties in an integrated way 'through the entire analysis, and BWR thermal-hydraulic models which were adapted for the Peach Bottom and Grand Gulf analyses. The Level 1 results of the LaSalle Unit 2 PRA are presented -in:
" Analysis of the LaSalle Unit 2 Nuclear Power Plant: Risk : Methods Integration and Evaluation Program (RMIEP),"- NUREC/CR-4832, SAND 92-0537, ten volumes. The reports are organized as.follows:
NUREG/CR 483'2 - Volume 1: Summary Report. NUREG/CR-4832 - Volume 2: Integrated Quantification and Uncertainty Analysis. NUREG/CR-4832 Volume 3: Internal Events Accident Sequence Quantification. NUREG/CR-4832 - Volume 4: Initiating-Events and Accident Sequence Delineation.
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1 NUREC/CR-4L32 - Volume 5: Parameter Estimation Analysis and Human Reliability _' Screening Analysis. NUREC/CR 432 - Volume 6: System Descriptions and Fault Tree Definition. NUREG/CR-4832 - Volume 7: External Event Scoping Quantification. NUREG/CR-4832 Volume 8: Seismic Analysis. NUREG/CR 4832 - Volume 9: Internal Fire Analysis. NUREG/CR-4832 - Volume 10: Internal Flood Analysis, The Level II/III results of the LaSalle Unit 2 PRA are presented in:
" Integrated Risk Assessment For the LaSalle Unit 2 Nuclear Power Plant:
Phenomenology and Risk Uncertainty Evaluation Progrcm (PRUEP)," NUREG/CR-5305, SAND 90-2765, 3 volumes. The reports are organized as follows: NUREG/CR 5305 - Volume 1: Main Report-NUREG/CR 5305 - Volume 2: Appendices A-C. NUREC/CR-5305 - Volume 3: MELCOR Code Calculations Important associated reports i been issued by the RMIEP Methods Development Program in: NUREC/C( J4, Recovery Actions in PRA for the Ris'- Methods Integration and Emuation Program (RMIEP); NUREG/CR-4835, Comparison and Application of Quantitative Human Reliability Analysis Methods for the Risk Methods Integration and Evaluation Program (RMIEP); NUREG/CR-4836, Approaches to Uncertainty Analysis in Probabilistic- Risk Assessment; NUREG/CR-4838, Microcomputer Applications and Modifications to thi Modular Fault Trees; and NUREG/CR-4840, Procedures for the External Event Core Damage Frequency Analysis for NUREG-1150. Some of the computer codes, expert judgement elicitations, and other. supporting information used in this analysis are documented in associated reports, including: NUREC/CR 4586, User's Guide for a Personal-Computer-Based Nuclear Power Plant Fire Data Base; NUREC/CR-4598,-A User's Guide for the Top Event Matrix Analysis Code (TEMAC); NUREG/CR-5032, Modeling Time to Recovery and Initiat; ag Event Frequency for _ Loss of Off-Site Power Incidents at Nuclear Power Plants; NUREG/CR-5088,. Fire Risk Scoping? Study: Investigation of Nuclear Power Plant _ Fire Risk _,- Including Previously Unaddressed Issues; NUREG/CR-5174, A Reference Manual for the Event Progression Analysis Code (EVNTRE); NUREC/CR-5253, PARTITION: A Program for Defining the Source Term / Consequence Analysis Interface in the NUREG-1150 Probabilistic Risk Assessments , - User's ' Guide; NUREG/CR-5262, PRAMIS: Probabilistic Risk Assessment Model- Integration System, User's Guide; NUREG/CR-5331, MELCOR Analysis for Accident Progression Issues; NUREC/CR-5346, Assessment of the XSOR Codes; and NUREC/CR-5380. A
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NUREG/CR-4832 Volume 5: Parameter Estimation Analysis and Human Reliability Screening Analysis. NUREG/CR 4832 - Volume 6: System Descriptions and Fault Tree Definition. NUREC/CR-4832 - Volume 7: External Event Scoping Quantification. NUREG/CR 4832 - Volume 8: Seismic Analysis. NUREG/CR-4832 - Volume 9: Internal Fire Analysis. NUREG/CR-4832 - Volume 10: Internal Flood Analysis. The Level II/III results of the LaSalle Uait 2 PRA are presented in:
" Integrated Risk Assessment For the LaSalle Unit 2 Nuclear Power Plant:
Phenomenology and Risk Uncertainty Evaluation Program (PRUEP)," NUREC/CR-5305, SAND 90 2765, 3 volumes. The reports are organized as follows: NUREG/CR-5305 - Volume 1: Main Report NUREG/CR-5305 - Volume 2: Appendicos A-G , NUREG/CP.-5305 - Volume 3: MELCOR Code Calculations Important associated reports have been issued by the RMIEP Methods Development Program in: NUREC/CR 4834,- Recovery Actions in PRA for ' the ' Risk Methods Integration and Evaluation Program (RMIEP); _ NUREC/CR-4835, Comparison and Application of Quantitative Human Reliability Analysis Methods for the Risk Methods Integration and Evaluation Program (RMIEP); NUREC/CR-4836, Approaches to Uncertainty Analysis in Probabilistic Risk Assessment; NUREG/CR-4838, Microcomputer Applications and Modifications to the Modular Fault Trees; and NUREG/CR 4840, Procedures for the External Event Core Damage Frequency Analysis for NUREG-1150. Some of the computer codes, expert judgement elicitations, and other support *ng information used in this analysis are documented in associated reports, including: NUREG/CR-4586, User's Guide for a Personal Computer-Based Nuclear Power Plant Fire Data Base; NUREC/CR-4598, A User's Guide for the Top Event Matrix Analysis Code (TEMAC), NUREG/CR-5032,_Modeling Time to Recovery and Initiating Evc.nt Frequency for Loss of Off-Site Power Incidents at Nuclear Power Plants; NUREC/CR-5088 Fire Risk Scoping Study: Investigation of Nuclear Power Plant Fire f Risk, Including - ) Previously-Unaddressed Issues; NUREG/CR-5174, A Reference Manual for the Event Progression Analysis Code (EVNTRE); NUREG/CR-5253 PARTITION: A Program for Defining the Source Term / Consequence Analysia ' Interface in the NUREG-1150 Probabilistic Risk Assessments, User's Guide; NUREG/CR-- 5262,- PRAMIS: Probabilistic Risk Assessment Model Integration System, User's ' Guide; NUREG/CR-5331, 'McLCOR An' ~ "is ' for Accident Progression Issues; NUREG/CR-5346, Assessment of the Az..... Codes; and NUREG/CR-5380, A l 1
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o - ,
User's Manut'. for the Postprocessing Program PSTEVNT. In addition the ree 'or is directed to the NUREG-ll50 technical support reports in NUREG/CR 4550 and 4551. Arthur C. Payne , Jr. Principal Investigator Phenomenology and Risk Uncertainty Evaluation Program and Risk Methods Integration and Evaluation Program Division 6412, Reactoc Systems Safety Analysis Sandia National Laboruturies Albuquerque, New Mexico 87185 4
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LIST OF ACRONYMS AC Alternating Current ADS Automatic Depressurization System ANS American Nuclear Society-APET Accident Progression Event Tree ATWS Anticipated Transient Without Scram CECO Commonwealth Edison Company CST Condensate Storage Tank DC Direct Current DCH Direct Containment Heating ECC Emergency Core Cooling FCI. Fuel-Coolant Interactions FSAR Final Safety Analysis _ Report HPCS High Pressure Core Spray HPME High Pressure Melt _ Ejection LOCA- Loss Of Coolant Accidant LPCI Low Pressure ~ Coolant Injection LPCS Low Pressure Core Spray , LWR- Light Water Reactor _ _ USNRC U.S. Nuclear Regulatory Commission-PRA Probabilistic Risk Assessment -k PRUEP Phenomenology end Risk Uncertainty Evaluation Program RCIC Reactor Core Isolation Cooling RMIEP- Risk Methods Integration and-Evaluation Program SNL- Sandia National' Laboratories SRV Safety /Reli.ef Valve STCP Source Term Code Package' 6-
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. . - -_..- - ~ .. - -. . . . _ ._- ... . - - - -
1
1.0 INTRODUCTION
The phenomenological ca!culations done in -' support of - the Level _ II/ITT portions of a Probabilistic kisk A'sessment (FRA) for the- LaSalle County Unit. 2 Nuclear Power Plant are presented in this report. The primary object %es of this PRA were to integrate internal and external events and to cevelop methods t or - representing and propagating uncertainties , I In support of these objectives, the U.S. Nuclear throughoutCommission Regulatory the analysis,_(USNRC) and Sandia decided to: (1) use detailed integrated thermal-hydraulic calculations to evaluate a base _line , representation of the dominant accident progressions of the PRA analysis, (2) Anvectigate the uncertainties arising from model limitations, l phenomenological uncertainties, and uncertainties in the initial conditions using sensitivity calculations and expert judgement. L The computer code chosen to perform the phenomenological calculations was d MELCOR1 which has.been under development since '082 at Sandia for the USNRC as a second-generation plant risk assessment too. and the successor to the Source Term Code Package (STCP).2 HELCOR 1, a fully integrated,
; engineering-level code.that models _the progression of severe accidents . in ,
nuclear sy s te'ns and facilicies. Characteristics ,of severe accident progre=sion that can be treated include tae thermal-hydraulic response. in the teactor coolant system and containment; core heatup and degradation; co e-concrete attack; combustible gas generation, transport, and 4 combastion; plant-structure thermal response; radionuclide release snd transport; and the impact of engineered safety features on thermal- [ hydraulic and radionuclide behavior. The lat< released version of the code at the time, version 1.8.0, was used in the .hese calculations. The calculations were designed to yiel'd necessary information for developing uncertainty distributions for key parameters in the accident. progression and source term _ analyses._ The sequences analyzed include high and low-pressure, short-term station blackouts;- an intermediate-term station blackout; and a long-term station blackout. .Scveral studies _ were performed for the high pressure short-term station blackout sequence to investigate the sensitivity of _ the source term 1results to the size - and , location of the containment. failure, and to combustion ignition parameters. , Following this introduction, a brief des ription of -the MELCOR code is given in Chapter 2. Chapter 3 contains the description of the input models used in the MELCOR calculations performed for this PRA. The sequences modeled are described in Chapter 4, while the results of the, calculations are presented in Chapter 5. Chapter 6 contu ns a ' discussion of the uncertainties in these analyses and the conclusions are in_ Chapter 7. a detailed listing of the input model is contained in Appendix A. l l -. l 1-1
. . = _ _ _
) i 1.1 References j 1. R. M. Summers, R. K. Cole. Jr., E. A. Boucheron, M. K. Carmel. S. E.
- Dingman, and J. E. Kelly, "MELCOR 1.8.0: A Computer Code for Nuclear '
) Reactor Severe Accident Source Term and Risk Assessment Analyses," NUREC/CR 55.11, SAND 90 0364, Sandia National Laboratories, Albuquerque, { NH, January 1991.. 4 ! 2. J. A. Cieseke, et. al., " Source Term Code Package, A User's Guide (Mod l 1)," NUREC/CR 4587, BMI 2138 Batte11e's Col.tmbus Division, Columbus, - j OH, July 1985.
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h e I e 2.0 MELCOR CODE DESCRIPTION 2.1 Introduction ) MELCOR is a fully integrated, engineering. level computer code that models I the progression of severe accidents in Light kater Reactor (LVR) nuclear ! power plants. The code has been under continuing development at Sandia for l the USNRC since 1982. The folleving summary was taken from Reference 1 which is available to provide a more detailed description of the code. One intended role of the code is the generation of information such as the timing of severe accident sequences and radionuclide source term estimates for support to a PRA. Characteristics of severe accident progression that can be treated with MELCOR include: the thermal-hydraulic response in the reactor coolant system, and containment; core hnatup and degradation; core-- t concrete attack; combustible gas generation, transport, and combustion;' i plant structure thermal response; radionuclide release and tranaport; . and the impact of engineered safety features on therma 14ydraulic and radionuclide behavior. HELCOR has been designed to facilitate sensitivity and uncertainty analyses through the use of sensitivity coefficients. Many parameters in the model-correlations are coded as sensitivity coefficients that can be altered by user input. The MELCOR code is a large and complex code with a structured modular architecture which manages large volumes of data and incorporates a wide variety of models with extensive coupling among them. MELCOR has reached the point were it is now being successfully applied in severe accident analysis. Some limited technical assessment activities have beer performed, although, a thorough assessment is needed to ensure that . the phenomenological modeling in MELCOR is technically justifiable and correctly implemented in the code. , i 2.2 Thermal-llydraulic Models The thermal-hydraulic behavior-is modeled with a lumped sum approach using_ control volumes connected by flow paths. Each volume is,= defined spatially by its volt.se versus altitude, may contain a gravitationally separated pool of single or two-phase water, and an atmosphere consistin5 'of any i combination of water vapor, suspended water droplets or noncondensible gases. The pool .and the atmosphere are each -individually treated ; by equilibrium thermodynamics (option selected for these calculations) such that they have equal pressures but may have unequal temperatures. Mass and energy sources may be defined'for any volume. Non condensible gases are modeled as ideal gases with temperatureJdependent specific heat capacities. The specific internal energy and enthalpy of an ideal gas is a ; function only of the temperature, the natural. state temperature, the energy of formation, the enthalpy of formation, the universal gas constant, and the molecular weight. The pressure is a 2-1
- _._.- - -- -.-.--_~..-------
function of the mass density, the temperature, the universal gas constant, and the molecular weight. The flow paths connect control volumes and define paths for moving hydrodynamic materials. Flow within the paths is treated as adiabatic but not, in general, isotropic. M.aterials do not reside within the flow paths, ' and do not transfer heat within the path. The flow paths may represent either a pipe Mka connection in a tank +and tube model or a cell boundary in a f'. nite difference mode'. allowing considerable modeling flexibility. A flow prth is characterized by a flow area, an inertial length, forward i and reverse form loss coefficients, wall friction parameters, volume connection elevations and opening heights and a valvo or a pump model may , optionally be attached. The governing thermal hydraulic equations are the equations of conservation of mass, momentum, and energy. The kinetic energy term in the energy , equation and the momentum flux term in the momentum equation are omitted on the a s s, ump t ion that MELCOR calculations will model flows in the far subsonic rarge making these terms unimportant here. Models exist for two-phase interactions such as two phase flow momentum exchange, bubbles rise ' and phase separation within a pool, and pool surface heat and mass transfer. Flows of unequal or even counter current velocities are ' possible. Critical flow models are included to predict critical flows at such locations as pipe breaks and pressure relief valv6s. Heat transfer between control volume materials _and surrounding curfaces is modeled using heat structures. Reactor system components such as pressure vessels, internal support structures, pipes, and steam generator tubes and containment components such as walls, floors, beams, and equipment may be modeled as heat structures. Heat transfer within tb structure is modeled as one dimensional conduction. Heat trans fer is exchanged between the structure's surface and the control volume materials by convection and by thermal radiation to the volume atmosphere. Thermal radiation between the structures is not currently modeled directly but may be simulated-indirectly under certain simplified conditions. The heat structures are characterized by: the conditions imposed on each of the two structure boundaries, geometry type, orientation, elevation, multiplicity, material composition, optional - power sources, length, and surface areas. The boundary conditions specify the connecting control volume, type of- surface heat transfer, and whether or not the structure connects to the pool, atmosphere or both. _A wide variety of surface heat transfer conditions are possible including a suite of convection heat transfer correlations. Models are included for surface condensation and evaporation of wster vapor in a pure steam environment or in the presence of noncondensible gases. Liquid films may.be present on the surface. In severe accident calculations, concrete may be heated under certain conditions to temperatures which will cause the concrete to . decompose. Trapped water and chemically bound water and carbon dioxide may be driven off. A model is included in MELCOR to simulate this degassing (different than ablation) of concrete. 2-2 ..
_ ___ __ _ _ _ _ _ _ . _ _ _ _ _ . _ _ . . _ . - _ m.____- _ __ _ _ _ . . . _ 2.3 Core Derradation and Debris Interaction Modela
~
Core models are included to simulate the nuc1 ear core ,during heatup and chemical and mechanical degradation during a severe accident. The core models treat all important modes of heat transfer within the core, as well as fission power generation, metal oxidation, debris formation, and relocation of core and structural materials during molting, candling, and l l slumping. Lower head heatup, failure, cnd debris ejection are also l modeled. ; The core and lower plenum regions of the reactor vessel are divided into a user specified number of concentric radial rings - and axial segments ' defining core cells. Each cell initially contains one or more of the , I possible intact components, i.e., fuel', - cladding, canister, or structure . (control rods are modeled as structures). The degradation processes may-produce a particulate debris component. Conglomerate debris -- (i.e. , core material that has melted and resolidified) becomes a part of the component ' onto which it resolidified. The components are made of some combination of six materials: urantwn dioxide, zircaloy, zirconium dioxide, steel, steel ' oxide, and control rod poison. Eact core cell interf acas with a control volume, primarily _through heat transfer which removes the heat from the. core when adequate coolant is available and heats the inter core gases when coolant is not available. Heat is exchanged with adjoining-heat structures surrounding the core. Models are available to treat the debris coolant and debris concrete interactions af ter ti,e debris is ejected from the reactor vessel by the core models. Ultimately, models will treat the debris-coolant interactions for both low and high pressure molten fuel pours from the reactor pressure-vessel and ex vessei steam explosions. Currently, only the mixing phase for a low pressure molten pour preceding a steam-explosion is treated. An option to bypass the debris coolant models and. go directly to the debris. 1 concrete interaction models we.s used for the LaSalle calculations. The debris concrete interaction models consist of the CORCON MOD 2 coder together with all necessary interfaces to the other models. Heat transfer from the debris to the concrete causes the concrete to . abate' and when concrete ablation occurs, water vapor and carbon dioxide gases are released-and the residual oxides melt a.nd join the debris pool. The- gases either bubble up through the debris or flow around the debris through a film between the. debris and the concrete. As the concrete ablation progresses, the cavity changes shaps. The CORCON code models the effects of heat transfer, concrete ablation, cavity shape change, ' gas generation,- and debris / gas chemistry. The cavity shape. model is - two dimensional. _ The debris-cot. crete interaction models interact with the control volume models to_ obtain' boundary cond! tions and to exchange heat. _For instance, any water existing in the reactor cavity control voluce is assumed to form a pool over -the molten debris pool, then,= heat transfer .between the debris pool and the water pool will bail-away the water, i' l r 23
.y--- ..~ m,,., ,_.,y.y-r ,-,w., ,wc- -
.,my,,ee,.,-m g g.-.,.im.. ..._---fv - - , . - , , , - . , . , , - - , .
2.4 Combustion Models The oxidation processes and debris chemistry during a severe accident would produce combustible hydrogen and carbon monoxide gases and models are included to simulate their combustion. These models consider the effects of burning on a global basis without modeling the actual reaction kinetics or tracking the actual flame front propagation. These models were extracted from the llECTR code.3 A burn is initiated if certain criteria are satisfied in a control volume, causing the reactants to be converted to water vapor and carbon dioxide. After a burn is initiated, its completeness depends upon the control volume conditions and it may propagate to adjoining volumes if a second set of criteria are satisfied. The threshold criteria for a deflagration burn is determined using a form of the LeChatelier's formula. MELCOR does not currently contain a detonation model, however, tests for the possibility of detonation are performed and a warning message is written should . the possibility exist. Steam inerting, oxygen depletion, and ign!ters may be simulated. 2.5 Entineered Safety Feature Models
!!any engineered safety features may be modeled using control volumes and flow paths such as emergency coolant injection systems which pump water from a soarce to the reactor vessel but additional models are needed to model containment sprays and fan coolers. The containment spray models treat the heat and mass transfer between spray droplets and steam. The droplets are assumed to fall through the containment at their terminal velocity and to be spherical and isothermal. A fan cooler may be specified for any control volume, the discharge optionally . specified to another volume, and its control tied to other facets of too calculation.
2.6 Decay Heat Power Models The decay heat power resulting from the radioactive decay of fission products is modeled. The decay heat power is evaluated for the fission products predicted to reside in the reactor core materials, cavity materials, and in suspended or deposited aerosols and vapors. MELCOR does not explicitly treat each decay. chain, but instead models the whole core decay heat power. The three options available in MELCOR are: (1) a summation of decay heat data from ORICEN code' calculations, (2) the American Nuclear Society's (ANS) 1979 National Standard 5 for decay heat power, and (3) user specified input. 2.7 Radionuclide Release and Transnort Models Radioactive fission products are released from the core - in the reactor. vessel during degradation and from the core debris -in the containment. 2-4
MW are provided to simulate these releases, their transport throughout th *: ; tint, and to predict the environmental source terms. Since a large _ unsuty ? it.sion product isotopes are poss!ble, the MELCOR inodels group
'h* Leotope= 'nto elemental classes with similar chemical behavior which in then teh - ad to by the name of a representative element. Each class . 3s v w % .is though it is a single element or compound. The models
' .t 4:9 m 15 default classes, listed in Table 2.7-1, but optional donal classes may be formed frota the default classes. These }
- 4. H :tive fission products may exist in either a vapor or an aerosol form
{ -in t .ay combine with other non radioactive materials such as aerosols
- tocmed during the melting of_ structural materials and concrete ab,lation.
The associated decay heat follows the fission products as they are transported throughout the plant. Models from other computer codes are included in MELCOR to simulate various aspects of the release and transport of fission products. The release of' fission products from the fuel or the core debris within the reactor vessel is predicted using either the CORSOR or CORSOR-M models8 and the ex vessel releases during the debris concrete interactions is predicted using the VANESA model.7 Fission products trapped in the fuel rod cladding gaps are released if any cladding temperature within a radial ring exceeds a user input tempeteture critorion. The gap inventory of fission products for that radial ring are then released at once. The aerosol dynamics models are acapted from the multicomponent MAEROS codes without condensation which is handled separately. The condensation and evaporation of fission product vapors from heat structures and aerosols is evaluated by the same equations 3 as in the TRAP MELT 2 code e Aerosols and f'ssion product vapors may deposit directly on surfaces such as-heat stru:tures . and water pools and aerosols may agglomerate and settle. The coagulation models include kernels for Brownian diffusion, gravitational settling, and impaction and the deposition models include kernels for gravity, diffusion, thermophoresis, and diffusiophoresis. Resuspension is currently not calculated. Model are available for the collection of fission products by pools, spra,vs . and filters. The pool model includes condensation at the pool entrance, Brownian diffusion, gravitational setting, inertial impaction, and evaporative forces for the rising bubbles -but only the aerosols are removed by pool scrubbing. The containment spray radionuclide models remove both vapors and aerosols _ from the atmosphere in a mechanistic treatment of the removal proces3es, closely, coupled to the behvior predicted by thermal hydraulic mouels. The aerosol removal is primarily by inertial impaction and interception, with diffusiophoresis and ' diffusion effects also included. The filter model can remove aerosols and fission product vapors with a specified maximum mass loading and the influence of the mass loading on the flow path resistance simulated. Fission product vapor chemistry effects can be simulated in MELCOR through the class reaction and class transfer models. Reversible and irreversible reactions can be used to model adsorption, chemisorption, and chemical reactions. 2-5
Table 2.7-1: Default Radionuclide Classes
- flans Number and Notre tintrher Elements 1 Noble Cases Ke Kr Rn lie Ne Ar 11 N 2 Alkali Metals fg Rb Li Na K Fr Cu 3 Alkaline Earths h.a Sr Be Mg Ca Ra Es Fm 4 Italogens 1 Br F Cl At 5 Chalcogens le Se S 0 Po 6 Platinoids En Pd Rh Ni Re Os Ir Pt Au 7 Transition Metals lip Tc Nb Fe Cr Mn V Co Ta V 8 Tetravalents C, Zr Th Np Ti lif Pa Pu C 9 Trivalents La Pm Sm Y Pr Nd Al Sc Ac Eu Cd Tb Dy llo Er Tm Yb Lu Arn Cm Bk Cf 10 Uranium M 11 More Volatile Main Group CsL lig Pb Zn As Sb Tl Bi 12 Less volatile Main Group SI} Ag In Ga Go 13 Boron B Si P 14 Water 112 0 15 Concrete ---
- Representative elements in each class are listed first and underlined.
l
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}'
_ . _ . _ _ _ _ _ _ _ _ _ _ _ _ . . _ . _ _ _ _ _ _ . _ ____ _._.__.______.m . s i ! 2.8 References
- 1. R. M. Summers, R. K. Cole, Jr., E. A. Boucheron, M. K. Carmel, S. E.
1 Dingman, and J. E. Kelly, "MELCOR 1.8.0: - A Computer Code for Nuclear j Reactor Severe Accident Source Term and Risk Assessment Analyses," 1 NUREC/CR 5531, SAND 90 0364, Sandia National Laboratories, Albuquerque, , NM, January 1991. i ! 2. R. K. Cole, Jr., D. P. Kelly, and M. A. Ellis, "CORCON MOD 2: A Computer i Program for Analysis of Molten Coro Concrete Interactions ," NUREC/CR-3920, SAND 84 1246, Sandia National Laboratories, Albuquerque, NM, 3 August, 1984 I J
- 3. S. E. Dingman, A. L. Camp, C. C. Wong, D. B. Kin 6, and R. D. Casser, i
< "HECTR Version 1.5 User's Manual," NUREG/CR 4507, SAND 86-0101, Sandia j
- National Laboratories. Albuquerque, NM, April, 1986. '
$ 4 D. E. Bennett, "SANDIA-ORIGEN User's - Manual, " NUREG/CR 0987, i SAND 79 0299, Sandia National Laboratories, Albuquerque, NM, October j 1979.
- 5. American Nuclear Society Standards Committee Working Group ANS 5.1, >
"Ame*1can National Standard for Decay Heat Power in Light Water Reactors," ANSI /ANS b.1-1979, American Nuclear Society, La Grange Park, '
IL, 1979. i 6. M. R. Kuhlman, et al., "CORSOR User's Manual," NUREG/CR 4173,_BMI 2122, Batte11e's Columbus Division, Columbus, OH, March, 1985,
- 7. D. A. Powers , J . E. Brockmann, and A. W. Shiver - "VANESA: A Mechanistic Model of Radionuclide Release and Aerosol Generation During Core Debris
,. Interactions with Concrete," NUREG/CR 4308, SAND 85-1370, Sandia National Laboratories,, Albuquerque, NM, September, 1985. i
- 8. F. Gelbard, "MAEROS User Manual," NUREG/CR-1391, SAND 80 0822, Sandia National Laboratories, Albuquerque, NM, December 1982.
- 9. H. Jordan, and M. R. Kuhlman, " TRAP MELT 2 Users Manual," NUREC/CR 4205, BM1-2124, Batte11e's Columbus Division, Columbus, OH, May 1985.
i s E G l~ 2-7 L
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3.0 INPUT MODEL DESCRIPTIONS The input models are described below according to their associated MELCOR code models. All of the calculations were started at the time of accident initiation (time zero) when the reactor tripped and the primary system immediately isolated. The models were initialized to the plant operating conditions when the accident was initiated. The sources of plant data used in developing the MELCOR input included the LaSalle Final Safety Analysis Reporti (PSAR), the LaSalle Systems Training Manual,2 a RETRAN code input file,' and other miscellaneous materials provided by plant personnel and their consultants. 3.1 Thermal and Hydrodynnmic Models The plant thermal and hydrodynamic models are discussed as they relate to the primary system, the containment, and the reactor / turbine building. Each of these three major _ systems represents a physical barrier to the release of the radioactive fission products to the environment. These systems are described in Chapter 1 of Volume 1 of-this report, 3.1.1 Primary System Control Volumes - The primary system was subdivided into seven control volumes as shown in Figure 3.1 1 and listed in Table 3.1-1, Figure 3.1 1 shows volume elevations versus scaled values for . the cross sectional areas, i.e., the width of the volume represents the cross sectional area at that particular elevation. The seven volumes represent the reactor vessel lower plenum, the core channels, the core bypass, _the upper plenum anu separators, the steam dryers and steam dome, the downcomer including the volume surroundin6 the separator tubes, and the recirculation loops. The vessel lower plenum free volume extends from the bottom head to the top of the core support plate (excluding the downcomer extending below the core plate) plus the jet pump diffuser volumes. The core volumes (channel and bypass) include the free volume within the core shroud between . the core support plate and the top of the fuel assembly canisters. The core channel volume is the sum of all the. free volume-within the fuel assembly canisters and the volume outside the canisters belongs to the ' bypass volume. The upper plenum and separator volume ' includes the free volume from the i top of the fuel assembly canisters up_ to the top of the separators enclosed by the upper plenum shroud, upper plenum shroud dome, and the
- RETRAN code input file and workbook provided by Commonwealth Edison Company.
31
Table 3.1 1: LaSalle Model Control Volume Input Data Control Brief Bottom Top Volume Description Elevation Elevation Volume (m) (m) (m3) Primary System 100 Lower Plenum 0. 8.25 107.2 103 Upper Plenum / Inner Separators 9.66 15.43 64.6-104 Steam Dryers and Dome 15.43 22.25 191.4 105 Downcomer/ Outer Separators 3.34 15.43 196.4 106 Recirculation Loops -3.37 4.38 24.8 111 Core Channels 5.27 9.66 54.8 j 121 Core Bypass 5.27 19.66 14.8 i l Containment 200- Wetwell -25.53 -7.15 8049, 201 Downcomers 21.11 6.08- 412. +
-203 Wetwell Pedestal -17.45- 10.29 ~214, 204 Drywell Pedestal 9.15 0.00- 256. ,
205 Drywell 6.24 23.84 '5933.
^
Reactor / Turbine Building 401 Unit 2 Lower Reactor Bld. 19.25 14.37 3036. 402 Unit 2 Upper Reactor bid. 25.53 26.17 31410, 403 Refueling Floor 14.13 41.56 58770. 404 Unit 2'. Steam Tunnel / Turbine Rm 28.68 29.30 38990. 405 Unit 1 - Reactor Bld. 19.25 26.17 34450. 406 Unit 1 - Steam Tunnel / Turbine Rm-28.68 29.30 38990. Environment 500 Environment -25.70 100. 1.0E9
-3 2
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o i i i e i i i i i 6 o o o o o o o o o o o 6 d i 6 d o 6 4 4 d c N N * * " a l I (R ) N011VATI3 l- Figure 3.1 1 LaSalle MELCOR Primary Syst.em 3-3
separator tubes. The steam dryers and stearn dome volume includes the free volume above the separators, and the portion of the stearn lines between the vessel and the isolation valves. The downcomer volume includes the separator return volume (outside of the separator tubes! and the downcomer volume. The recirculation loops' volurne includes the volurnes of the recirculation pumps, the pump suction lines, the pump discharge lines, and the jet pump inlet risers. The recirculation loops' volurne was separated from the downcomer volume because water would be trapped in the loops' volume, whereas, the downcomer water would be boiled away by core shroud heat transfer. 4 The primary sys tern control volumes were initialized near the plant operating parameters defined in the Final Safety Analysis Report.1 The pressure of the steam dorne was set at 7.213E6 Pa (1046 psia) with the pressure in the other volurnes adjusted for their hydrostatic heads. The core and bypass volumes, the upper plenum / separators volume and the steam dryers / dome volume were all initialized at_ their saturation temperatures. The core channels were initially voided at 384 and the upper plenuin at 37%. The recirculation loops,-the downcomer, and the lower plenum all had some initial subcooling. The initial water level was at 14.312 meters above the bottorn head and was located within the steam separators. Elow Paths - The primary system was modeled with nine flow paths inside the system, four flow paths. connecting the primary system to the containment, and one mass / energy source path representing the one emergency core cooling system modeled. These flow paths are shown in Figure.3.1 1 and listed in Table 3.1 2. The inside flow paths include the core channel and bypass inlets and exits, steam separators and separator returns, dryer drains, jet pump diffusers, and the recirculation loops. The bypass loss coefficients were calculated to establish a calculational channel to bypass flow ratto equivalent to the actual plant ratio of 10. The four flow paths between the primary system and the containment include the safety / relief valves and associated piping, the path simulating lower head vessel breach, the recirculation pump seal leakage, and the. control rod drive leakage paths. The LaSalle reactor has a total of _18 safety / relief valves ,which were modeled with a single flow path. Seven of the valves are also-used by the Automatic Depressurization System (ADS). The flow path valve is controlled such that the appropriate number of valves open or close according to their _ specified set points and the differential pressure between the vessel dome and the containrnent wetwell. The applicable operating procedures
- for the-ADS were rnodeled. The procedures modeled include:
- Plant Operating Procedures-supplied by Commonwealth Edison Cocpany i
3-4
. . . . . . - . . . . . . - - - - . - . _ . ~ . - - . - - . _ - . - - . - - . . . - . . - . - - . _ - - Table 3.1-2: LaSalle Model Flow Path Input Date Flow Volumes Elevations Flow Loss Coeff. LLt.h Descr8otion EL91D _I9 fIED - I9 MAA ESL. BfL. (m2) 1 1 Primary System 15 Separators 103 104 15.43 15.43 3.32 9.1 2.8 16 Dryer Drains 104 105 15.43 15.43 13.9 1. 1. 17 Separator Drains 103 105 13.10 13.10 3.20 3. 3. 18 Jet Pump Diffuser 105 100 8.25 3.34 .678 1.4 2.6 19 To Recirculation Loops 105 106 4.38 4.38 .458 4. 4, 51 Inlet to Core Channels 100 111= .5.27 5.27 7.86 21.8 29.6-61 Inlet to Core Bypass 100 - 121 5.27 5.27- 5.53 1340. 1640, 71 Outlet from core-Channe- 111 103 9.66 9.66 7.86 9.1 .9.4 1 81 Outles from Core Bypass 121 103 9.66 9.66 5.53 446. -546. l Primary System to Containment 21 Safety / Relief Valve Piping 104 200 16.46 24.85 1.104 7.6 7.6 31 Vessel Lower Head Breach 100 204 0. .23 .1 1. 1. ' 370 Recirculation Pump Seal Leaks 106 204 -3.30 3.3- 2.8E 6 1. 1. 371 Control Rod Drive Seal Leaks 100 204 .10 -1. 3.4E 6 1. 1. Containment 24 Downconer Inlets 205 201 -6.08 -6.08 27.42 5.2 5.2 25 Downcomer Exits 201 200 -21.11 -21.11 27.42 1. 1. 26 Vacuum Breakers 200 205 -7.15 -6.24 1. 1. 1. 27 Drywell Pedestal Hatches 204 205 4.5 -4.5 2.46 ~2. 2, 28 Drywell Floor Drains 205 204 -6.24 8,5 .0649 2.5 2.5 29 Nitrogen Inerting Bypass Line 205 200 14.9 -9.2 .0011 8. 8. l 40 Drywell Pedestal Drain Pipes 203 204 10.3 - -9.15 -.0162 1. 1. 41 Wetwell Pedestal Ports 203 200:-15.11 -15.11 4.573 1. 1. 42 Wetwell Pedestal Rupture 200 203 17.35- 17.35 .1 10. 10. ; Containment to Reactor-Building 101 Small Drywell Vent (2 in.) 205 402 0. O. .0020 15.- 15. 102 Large Drywell Vent (18 in.) 205 402~ 0. O. .1642 10. 10, 103 Small Wetwell Vent (2 in.) 200 402- -9. 9. .0020 15. 15. 104 Large Wetwell Vent (18 in.) 200 402 -9. - 9. : .1642 10, 10. 201 Wetvell Wall Leak Failure- 200 402- -12. -12. .00929 '100. 100, 202 Wetwell Wall Break Failure 200 402- 12. -12. 1, 1. 1. 203 Drywell Wall Leak _ Failure 205 402- 5. 5.. .00929 100. 100. ' 204 Drywell' Wall Break Failure 205 402 -5. -5. 1. 1. 1. 205- Drywell Head Leak failure 205 403 -22, 27. .00929 100. 100. 206 -Drywell Head Break Failure 205 403: 22; 27. 1. '1. 1. L l r 35
. _ _ _ _ . . . _ _ . _ . . - _ , . _ _ _ . _ _ _ .. _ . ,._.c
Table 3.1 2: LaSalle Model Flow Path Input Data (Concluded) Finw Volumes Elevations Flow Loss Coeff. Egib Deseriotion EI.9JD A EIPlB 19 M ESL. EC.L. (m2) Reactor / Turbine Building 401 Upper to Lower Unit 2 402 401 -14.17 14.37 ?.2.17 4 4. 402 Unit 2 to Refueling F1cor 402 403 26.17 27, 33.1 1. 1. 403 Unit 2 Stair to Refuel. Floor 402 403 26.17 27. 3.08 2.9 2.9 404 Unit 2 Steam Tunnel Raceway 401- 404 18.41 18.41 3.01 2.8 2.8 405 Unit 2 to Unit 1 Doors 402 405 4. 4.0 18.9 1. 1. 412 Unit 1 to Refueling Floor 405 403 26.17 27, 33.1 1, 1. 413 Unit 1 Stair to Refuel. Floor 405 403 26.17 27, 3.08 2.9 2.9 414 Unit 1 Steam Tunnel Raceway 405 406 18.41 18.41 3.01 2.8 2.8-Reactor / Turbine Building to Environment 406 Unit 2 Steam Tunnel Top Panel 404 500 28.8 28.8 3.53 1. 1. 407 Refueling Floor Metal Walls 403 500 33.9 33.9 40. 1. -1. 408 Unit 2 Infiltration 402 500 -9.88 -9.88 .092 1, 1, 409 Unit 1 Infiltration 405 500 -9.88 9.88 .092 1. 1. 410 Unit 2 Standby Gas Treatment 402- 500 22.7 122.7 .164 1. - 1. 411 Unit 1 Standby Gas Treatment 405 500 22.7 22.7 .164 1. 1. 415 Unit 1 Steam Tunnel Top Panel 406 500 28.8 28.8 3.53 1. 1. 416 Refueling Floor Infiltration 403 500 27. 27. .01 10. 10, 417 Unit 2 Turb. Rm. Infiltration 404 500- -14.3 -14.3 .01 10, 10. 418 Unit 1 Turb. Rm. Infiltration 406 500 -14.3 14.3 .G1 10, 10. P 3-6
- gradual depressurization (at 100 oF per hour) to limit stresses to the l vessel,
- opening one ADS valve if the reactor vessel water level drops below 275 inches (instrument zero is $27.5 inches above bottom head),
opening all seven ADS valves if the level drops below -275 inches and j the vessel pressure is less than 700 psig, )
- opening all seven ADS valves if the drywell temperature exceeds 340
*F, opening all seven ADS valves if the drywell pressure exceeds 29 psig,
- the depressurization associated with the suppression pool heat capacity operating-temperature limit,
- the failure of all ADS valves from loss of adequate air pressure should the drywell pressure exceed 85 psig.
- the failure of all- ADS valves from loss of control _due to - battery depletion,
- operator closure of the valves if the vessel drops below the 57 psig
',ch is needed to supply turbine steam for the low pressure injection system, and
- manual opening / closing override controls.
- 1 The vessel breach flow path which connects the vessel lower plenum and the- I drywell pedestal is initially closed. Af ter lower head - failure occurs, the flow area is set equal to the vessel breach area calculated by MELCOR, but limited to an arbitrary maximum area of . 1. m2 The flow paths modeling the leakage from the recirculation pump seals : and control rod drive seals were modeled to simulate a typical combined leakage flow rate of about 4 gpm during normal operation.
The High Pressure Core Spray System (HPCS), - the only emergency core-cooling system modeled, 'is _ modeled using mass / energy sources / sinks. -A source / sink input model merely adds .the emergency - coolant and its associated energy to the rea:cor vessel while simultaneously removing the same from the coolant supply._ HPCS could- also have been modeled using MELCOR's flow path and pump models, but.the source / sink model'.is adequate.
- less complicated, and potentially mere stable. The emergency procedures specify that HPCS is initially supplied - by 1 the.. Containment--- Spray Tank until the suppression pool water level exceeds'two inches above the normal level, then HPCS is supplied by'the suppression pool. The water level in-the reactor vessel z in these calculations is maintained cin the steam separators until af ter HPCS fails,- therefore, it is not necessary to model HPCS as an actual spray since:the core remains. covered.
~
(Note that this ' _ input model would be inadequate in a calculation involving the initiation l l , 3-7 1
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of IIPCS af ter the core is significantly exposed since a calculation of this type would involve the interaction of the core spray and the heated core.) The HPCS flow rate was controlled using the differential pressure between the reactor vessel and the suppression pool (2 feet from the bottom), and the llPCS pump head curve. HPCS will activate on either a low vessel water level (-50 inches of collapsed water) or 1.69 psig drywell pressure and will trip off when the level exceeds 55.5 inches. Once activated, op e r.n o r action will maintain a 12.5 inch minimum level until llPCS failuse. The llPCS was modaled to fail whenever the containment failed or was vented. .,pothetically due to a severe environment in the reactor building where the pumps are located. Additionally, the HPCS controls included the ability to manually control the HPCS model by attenuating the flow rate or altering pump trip levels to overcome numerical problems encountered when the water level of a two phase pool reaches the top of a control volume. When this happens in the current version of MELc0R, the two-phase pool is converted into a single phase . pool and an atmosphere. This causes the volume water level to rapidly oscillate, which can also propagate into other volumes as well, effectively destabilizing and slowing the calculation. . Heat Structur,n The reacto). vessel, the vessel internals not' included in the input to the core models, and the recirculation loops' piping are divided i :to 22 heat structures. These structures are listed in Table 3.1 3. The core fuel assemblies, the control rods, the top guide, the core plate, the control rod tubes and. housings, and the portion of the reactor vessel head below the core (with the same diameter as the core shroud) are modeled with input to the core models , so they . are not represented as heat structures. The reactor vessel (excluding the portion of the lower head directly under the core) was divided into six structures. These are the hemispherical upper head, a cylindrical section ' adjacent to the steam dryers / dome volume, a cylindrical section adjacent to the downcomer volume, and three lower head rings adj acent to the bottom three core axial cells. The vessel outer insulation was simulated by specifying - a constant heat transfer coefficient (10 W/m/K ) which predicted a heat loss of about 1 MW during normal operation. In retrospect, this insulation model is inadequate to accurately estimate the temperatures. of the vessel, and internals late in the accident when heat transfer. through . the vessel becomes important. This is discussed further in the results Section 5.1.1.2 and in Chapter 6 on uncertainties. The internal structures that are modeled as heat structures : .iclude the-steam dryers, the steam separators, the upper plenum shroud dome, the upper plenum shroud, the core shroud. and the lover plenum shroud. The core shroud, the lower plenum shroud, and. lower head were subdivided into rings corresponding to each of the 13. axial core cells, This.not only 38
Table 3.1 3: LaSalle Model Heat Structure Input Data . Heat Volume Lower Thick Structure Description ._L. .JL. linterial Elevation Ana ness (m) (m2) (m) Primary System 10402 Vessel Upper Head 104 205 Ste31 19.04 63.84* .18 10403 Vessel Upper cylinder 104 205 Steel 15.43 72.30b .18 10501 Vessel Main cylinder 105 205 Steel 3.34 242.lb .18 10401 Steam Dryers 104 104 Steel 15.60 697.0 .0066 10303 Steam Separators 103 105 Steel 10.46 $20.0 .016 10302 Upper Plenum Shroud Dome 103 105 Steel 10.46 21.60 .051 10301 Upper Plenum Shroud 103 105 Steel 9.66 14.04b .051 12113 Core Shroud Axial Cell 13 121 105 Steel 9.30 5.82b .051 12112 Core Shroud Axial Cell 12 121 105 Steel 8.67 10.29b ,051 12111 Core Shroud - Axial Cell 11 121 105 Steel 8.03 10;29b ,031 ' 12110 Core Shroud - Axial Cell 10 121 105 Steel 7.40 10.29b .051 12109 Core Shroud Axial Cell 9- 121 105 Steel 6.76 10;29b .051 12108 Core Shroud - Axial Cell 8 121 105 Steel 6.13 10. 29b .051 12107 Core Shroud - Axial Cell 7 121 105 Stael 5.49 10.29b .051 12106 Core Shroud - Axial Cell 6 121 105 Steel 5.27 3.68b .051 10005 Lower Plenum Shroud - Cell 5 100 105 Steel- -5.19 1.19b .051 10004 Lower Plenum Shroud - Cell 4 100 105 Steel 4.56 10.34b .051 10014 Lower Plenum Shroud - Cell 3 100 105 Steel- 3.34 20.10b .051 10003 Lower Head - Axial Cell 3 100 205 Steel 1.28 27.84b .18 10002 Lower Head Axial Jell 2 100 205 Steel .64 8.71b .18 10001 Lower Head Axial Cell 1 100 205 Steel 0. 8.71b .18 10601 Recirculation Loop Piping 106 205 Steel -3.37 52.57b .035 Containment 20001 Wetwell Va.1 200 402- St/ Conc.-25.53 1525. 1.22 20002 Wetwell Base Slab 200 Sym. St/ Cone 25.53 482.8 7,01 20003 Drywell Floor Support Columns Sym. 200 -St/ Cone 25.53 1122.5 .540 20004 Wetwell Pedestal Section 2 200' 204 St/ Conc ~9.14 - 58.61 1.48 20005 Wetwell Pedestal Section 3 200 203 St/ Cone -17.44 160.8 1.49 20006 Wetwell Pedestal Section 4 200 Sym. St/ cone -25.52 215.4- 1.48 20007 WW Misc. Steel (above water) 200 Sym. Steel- -8.0 1500. .019 20008 WW Misc. Steel (below water) 200- Sym. Steel 24.0 940. .019 20101 Downcomers 201- 200 Steel. -21.11 2761.b .0095 20301 Drywell Pedestal Floor 203 204 Conc. -10.29 29.92 1.14 l 20501 Drywell Floor 200 205 St/ Cone 7.15 419.0 .921 ; I 20502 Reactor Shield 205 205 St/ Conc. .26 313.4b .568 l 20503 Drywell-Walt - 205 402. St/ Cone- 6.24 2100. 1.83-i 20504- Drywell Pedestal Section 1 205 204 Cone. 6.23 - 147.6 1,47 ' 20505 Drywell Head 205 Sym. S/ Gap /C-14.8- 181.Ob 4.44 20506 DW Misc. Steel (vertical) 205 Sym. Steel- -1.0 7500. 0058' 20507 DW Misc. Steel (horir.ontal)_ 205- Sym. Steel -1.0 =7500. ,0058 i 39 i
- ,,----n.~,. - , . , -, n -,,ne, ,~m. ,,,,e -,-i,,. ~~ , -
- ~~--
Table 3.1-3: LaSalle Model Heat Structure Input Data (Concluded) llent Volume lower Thick Structure pescription _lu _R_ Material Elevation Area -ness (m) (m2 ) (m) Reactor / Turbine Bu!1 ding 40101 Floor of Lower RB 2 Sym. 401 Conc. -19.24 911. .406 40102 Interior Walls of Lower RB 2 401 _Sym. Conc. 19.25 893. .523 40103 Floor between Up RB 2/Lo RB-2 401 402 Conc. 14.95 911. .670 40104 Wall between Io RB-2/ST-2 401 404 Conc. -19.25 68. .914 40105 Vall between Lo RB 2/RB 1 401 405 Conc. -19.25 105. .702 40106 Lower RB 2 Hisc. Steel 401 Sym. Steel 19.25 558. .0031 40201 Interior Floors in Up RB 2 402 Sym. Conc. 4.40 8730. .320 40202 Interior Walls of Upper RB 2 402 Sym. Conc. 0.69 2630. .501 l 40203 Upper RB.2 Misc. Steel 402 Sym. Steel 0.72 8820. .0071 l 40204 Floor between Up-RB 2/Reful-Bay 402 403 Conc. 22.36 1450. 1.00 ) 40205 Refueling Pool Walls in RB 2 402 403 Conc. 16.70 864, 1.83 40206 Floors between Up RB 2/ST-2 402 404 Conc. -2.39 150. 1.22 40207 Walls between-Up RB 2/ST 2 402 404 Conc. 10.29 171. 1.05 40208 Walls between Up-RB-2/RB-1 402 405 Conc. 1.07 1270. .670 40209 Exterior Walls of Upper RB 2 402 500 Conc. 2.24 4440. .476 40301 Floor of Refueling Bay 403 Sym Conc. 25.56 490. .670 40302 Refueling Bay Misc. Steel 403 Sym Secel 26.20 464. .0071 40303 Refueling Bay Ceiling 403 500 Steel 41.35 3660. .0032 40304 Refueling Bay Exterior Walls 403 500 Steel 26.20 4020. .0032 40401 Interior Walls of S1/ Turbine 2 404 Sym. Conc. -12.14 135. .670 40402 ST/ Turbine 2 Exterior Wallt 404 500 Conc. -21.00 1570. 1.83 40403 ST/ Turbine 2 Floors 404 500 Conc. -16.20 413. 1.83 40501 Floors of Reactor Bld. 1 Sym, 405 Conc. 0.60' 11500. .325 40502 Interior Walls of RB 1 405 Sym. Conc. 3.53 6050. .560 40503 Reactor Bid. 1 Misc. Steel 405 Sym. Steel -0.47 9360. .0069 40504 Floor between RB.1/ Refuel Bay 405 403 -Conc. 22.36- 1450. 1.00 40505 Refueling Pool Walls in RB 1 405- 403 Conc. 16.70 864, 1.83 40506 Floors between RB-1/ST 1 405 406 Conc. -2,39 150, 1.22 40507 Walls between RB-1/ST 1 405 406 Conc. 12.83 239. 1.01 40508 Exterior Walls of RB 1 405- 500 Conc. 2.24 4440. .476 t 40601 Interior Walls of ST/ Turbine-1 406 Sym Cone. -12.14 135. .670 l 40602 ST/ Turbine 1 Exterior Walls 406 500 Conc. -21.00 1570. 1.83 40603 ST/ Turbine-1 Floors 406 500 Conc. -16.20 413, 1.83 i l l l
- spherical geometry (inner area)
I b - cylinder geometry (inner area)
- cylinder geometty (outer area) 3-10
l allows the calculation to determine an axial temperature gradient for i these components but also avoids potential numerical problems associated I with the interface between the core and heat structure models. One heat structure was used to simulate the recirculation loops' pumps and i 1 piping. i 3.1.2 Containment Building Control Volumes The containment was subdivided into five control volumes as shown in Figure 3.1-2 and listed in Table 3.1-1. These volumes are the . drywell, the wetwell, the vent downcemers, and the drywell and wetwell ) pedestal regions. The downcomers were included as a separate volume to l model the hydrodynamics of clearing the downcomers. The upper and lower l pedestal regions are modeled~ separately to support a debris holdup model. . The molten core debris is initially contained in the upper pedtstal l region; but, upon failure of the floor separating the two regions, the i debris is transferred to the lower pedestal region. The drywell volume includes the free volume above the drywell floor within the drywell containment wall excluding the drywell pedestal and- the reactor vessel, but including the space between the reactor vessel and the shield wall. The wetwell volume includes the free volume (including the pool) located bet.wcen the base slab and the drywell floor and between the pedestal wall and the containment wall. _ The downcomer volume includes the volume contained within all of the downcomers (98 total) from entrance to , exit. The drywell pedestal volume includes the volume within the pedestal above the cavity floor to the reactor vessel. The vessel skirt is solid, separating the pedestal volume from the gap between the vessel and shield wall. The wetwell pedestal volume is the space within the pedestal below the floor which separates the upper and lower regions. The containment control volumes were initialized at the plant operating parameters. The entire _ containment was' initialized at 1.065ES Pa . ( . 75 psig). The drywell atmosphere _ temperature was initialized at 330.4 K (135 oF), and the wetwell atmosphere and suppression pool at 310. 9 - K (100 *F) . The suppression pool was initialized at the normal water level. Flow Paths - The containment is' modeled ' vith nine flow paths interconnecting the containment volumes and ten flow paths connecting the containment to the reactor building. These flow paths are shown in Figure 3.1-2 and listed in Table 3.1-2. L The interconnenting flow paths include the vent downcomer entrances and l -exits, the drywell pedestal hatches and ports -(five accesser assumed fully ' open) between the pedestal and the drywell, the drywell floor drains (two 8 inch lines), an open line (1/2 inch) in the nitrogen inerting system between the drywell and the wetwell, and the wetwell pedestal vents (four 3-11
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ports each with a 3 foot 11.5 inch diameter) between the vetvell pedestal and the wetwell which are all modeled as fully open throughout the entire calculation. The downcomer exits are initially blocked with suppression pool water until sufficient drywell pressura forces the water from them and allows gases to pass inte the pool. The containment vacuum breakers between the drywell and vetwell are modeled to allow flow only in one direction (wetwell to drywell) . The breakers are controlled by the differential pressure acrcss them. The drywell pedestal drain lines (two 4 inch pipes), which transport water from the drywell pedestal sump tank downward through the cavity floor and then out through the wetwell to the reactor building, are modeled to fail as a result of interaction with the core debris. These drain pipes are modeled to fail 20 minutes after the first core debris is ejected from the reactor vessel onto the drywell pedestal floor. The 20 minute failure time is based on a solicitation or expert opinivn (described in Volume 2 of this report). The experts believed that the pip,es would not remain intact more than about 20 minutes af ter molten debris was ejected onto the ' cavity floor. The failure of the pipes are assumed to result in two four inch holes which are modeled as a single flow path. A 11os peth te included to model the possible interaction between the molten core debric m the wetwell pedestal and the supprusion pool should the pedestal wall fair. T'e valve associated with the flow path can be opened or closed on any code restart and a restart is written when the core debris is predicted to penetrate the concrete pedestal wall. When the valve opens, water may flow from the suppression pool into the wetwell pedestal region where it will interact with the debris, Four of the te.. flow paths connecting the containment ar,d the reattor building are containment vent lines consisting of 2 inch and 18 inch pipes from the containment drywell and from the containment wetwell terminating in the upper reactor building of Unit 2._ Vent ducting within the reactor building, designed to conduct the vent flows from the building, is assumed to easily fail by over pressure and so is not modeled. The valves associated with these flow paths can be opened or closed on any code restart. The other six flow paths connecting the containment and the reactor building modeled containment failure by overpressurization. These flow paths include a path modeling small leaks and a path modeling large breaks for each of three locations; the wetwell wall, the drywell vall, and the drywell head. Normally, a ec.culation will simulate just one_of the flow paths - and the selection is made by opening or closing the . valves associated with each path. The leakage flow paths were . modeled vith a high flow resistance and small' area to_ simulate small. jagged cracks in the concrete, whereas, the break paths were 'modeled _ with a large area and small flow resistance to simulate a large open hole in the wall'. Containment failure was determined when the pressure in the appropriate 3-13
compartment (wetwell or drywell) exceeded the specified failure pressure of 1.345E6 Pa (195 psig). The break flow area can be adjusted by changing the valve opening fraction. The leakage and break flow paths have riaximum flow areas of .00929 m a (,1 fe z) and 1 m2 (10.76 f t )2 . Senairivity studies involving the failure flow area and location are easily accomplished from a single restart at the time of predicted containment failute The elevation of the failure flow paths is e ,out midway between the top of the suppressien pool and the drywell floot n r the patna from the wetwell; j about four feet above the drywell floor I the paths from the drywell; , and about six feet below the top of the drywell for the drywell head j paths. The wetwell and drywell flow paths exited into the upper Unit 2 l reactor building and the drywell head paths exited into th refueling bay. ' Heat Structures - All of the containment internal and external valls and structures (except the wetwell pedestal floor which is included in the d;bris concrete interactions model) are modeled as 17 heat structures. These structures are listed in Table 3.1 3. The wetwell base slab, the wetwell vall, the drywell wall, and the drywell head which form the outer boundaries of the containment are constructed of steel-lined thick concrete. The reactor pedestal wall-, which extends from the wetwell base slab to the reactor vessel, was sutdivided inte four sections so that the structure boundaries matched the adjoining volumes. Pedestal Section 1 is the top portion of the pedestal wall above the midpoint of the- drywell floor and its boundaries are the drywell and the drywell pedestal. Section 2 consists of the pedestal between the midpoint of the drywell cavity floor and the midpoint of the drywell floor and its boundaries are the vetwell and the drywell pedestal. Section 3 consists of the pedestal between the wetwell . pedestal floor and the midpoint of the cavity floor and its boundaries are the wetwell and wetwell pedestal. Section 4 consists of the pedestal below the wetwell pedestal floor and its one boundary is toe vetwell; a symmetrical boundary condition is imposed on the second boundary because the solid concrete plug in the pedestal bottom is. included in the debrit interrection model, The the pedestal wall is steel-lined on the outsim in the wetwell and the inside of the - vetwell pedestal region. Due a the high temperatures associated with _the molten debris, it is probable that the steel liner inside the wetvell pedestal region w!11. not remain intact. The input models assume that this liner _ slumps or melts away from the concrete and falls into the molten debris. The mass of this steel is added to the debris-concrete interactions model. Other internal concrete structures include the drywell' floor, the drywell floor support columns, the drywell pedestal floor, and the reactor shield well. The steel structures include the . downcomers between the downcomer 3 14
1 j and wetwell volumes and miscellaneous steel. The miscellaneous steel is j subdivided into vetwell steel above the pool surface and steel below the i pool surface, and drywell steel with vertical and horizontal surfaces, i Internal concrete structures are modeled to release steam and carbon
- dioxide gases when heated to the appropriate temperatures. The gases released from the steel-lined internal structures are assumed to enter the
~
containment atmosphere either by Leping through the concrete wall to an t- unlined side or by flowing under the liner to an opening. The structures j modeled for degassing include all four reactor pedestel sections, the ! drywell pedestal floor, the drywell floor,-and the reactor shield. Since gases released from the wetvell and drywell walls are outside of the steel liner pressura boundary, degassing for these walls is not modeled. The concrete degassing input parameters, which are listed-in Table 3.1-4, are based on experimental data reported in Reference 3. The free water vapor in the concrete is modeled for release over three temperature ranges to accommodate its distinctly non linear release characteristics. These ranges are from 300 to 322 K, from 322 to 240 K, and from 340 to 350 K for 6.67, 13.33, and 80s, respectively, of the available free water, The bound H 2 O is released from 350 to 750 K. Above 750 K, all water _ vapor is ' assumed released. The carbon dioxide gas is. assumed to release linearly over a temperature range from 670 to 870 K. Since the drywell is normally ; operated at temperatures which exceeded 300 K,- the gas densities are modified for drywell structures to simulate partial initial dehydration of the concrete. The gases from the drywell floor and the reactor shield are added to the drywell atmosphere. The gasen from the upper two sections of the r/ actor pedestal wall are added to the drywell pedestal atmosphere, from the third section to the wetwell pedestal atmosphere, and from the bottom section to the wetwell atmosphere. The gases from the drywell pedestal floor are added to the wetwell pedestal atmosphere. 3.1.3 Reactor Building Control Volumag The reactor and turbine buildings are divided into six volumes as shown in Figure 3.1 and listed in table 3.1-1. The Unit 2 ; I reactor building is subdivided into a-lower and an upper volume, the Unit I reactor building is modeled with a single volume, and the refueling j floor with a single volume. The steam tunnel and turbine' room for each i unit are both modeled with single volumes. This six volume model was l developed from the results of a study made with a more detailed 31 volume model.' , In cddition, the outer environment is modeled as a single large control volume (1.E9 m3). This volume is large enough to ensure . that its temperature, pressure, and mole fractions do not change significantly during the course of a calculation. 't 3-15
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- Gas Mass lleat of lyra Lower Upoer Fraction Density Reaction (K) (K) (kg/in3) (J/kg)
!!a0 Free 6.67% 300 322 .0017 3.98 2.367E6 13.33% 322 340 .0034 7.95 2.367E6 80.00% 340 350 .0204 47.72 2.367E6 H2 0 Bound 350 750 .0189 44.18 6.067E6 CO 2 670 870 .2017 472.0 4.043E6 Concrete Density -2340 kg/m3 ,
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-- - . . . _ - - ~ - - - _ - . - - - . - - - -. - =- . ---
The lower volume of the Unit 2 reactor building includes the rooms of the second level (floor at the 694 ft level). The results of the detailed study indicated that the rooms of the bottom floor of both reactor buildings could be excluded from the six volume model, due to their
- relative unimportance to the overall calculation. The upper Unit 2 reactor building includes the rooms above the floor at the 710 ft level and belos the refueling floor. The refueling bay volume contains the refuelir,g bay and refueling pool. The reactor building e ontrol volumes wera initir?ized at atmospheric pressure and with a temperature of 305 K.
i Flow Paths The reactor and turbine buildings are modeled wie eight flow-paths interconnecting its volumes and ten flow paths connecting the ; buildings to the environment. These flow paths are shown in Figure 3.1 3 and listed in 'lable 3.1-2. The interconnecting flow paths include the combined doors, stairways, etc. between the lower and upper volumes of the Unit'2 reactor building, between the Unit 1 anu Unit 2 reactor buildings and the refueling bay, between the two reactor building units, and between the reactor-buildings and the turbine rooms _ through the steam tunnels. The flow-pathu between the reactor buildings and the refueling bay are modeled with blowout panels (opening on either .75 or 2 psi) which remain open after the simulated blowout of the hatch cover or door. The doors between the two reactor buildings are modeled to open with a differential pressure of .75 1 psi and to reinain open. The flow paths between the reactor buildings and
- the steam tunnels and turbine roorns model flap covers which open with a pressure differential of 100 Pa (.0145 psi) and close again upon the loss of differential pressure.
The flow - paths between the reactor / turbine buildings and the environment include a path modeling a large blowout. failure of the refueling bay sheet metal walls at 2 psi differential pressure, flow paths modeling a panel blowout at the top of each _ steam tunnel at 1_ psi, fan _ driven flow paths ' modeling the standby gas treatment system in each reactor building, and several infiltration flow paths. 'The fan driven paths are inactive during the calculations of station blackout sequences. The infiltration flow paths model various small building leaks; one path for each reactor building, one for each turbine room, and one for the refueling bay. Heat Structures - The reactor and turbine _ buildings' structures are modeled using 33 heat structures. These heat structures ' exclude the containment walls that are included in the containment models, and certain. portions of the buildings deemed unimportant. These heat structures are listed in Table 3.1-3. Walls, floors, or equipment with like or similar param'eters were-conglomerated into a single structure. Some of these structures - include the interior concrete walls or floors of the control volume, some are-3-18
- - - - - - - - . - . ..- ._ . . . . ~ - . ~ _ - _ - . _ - - . _ - - - . - -. -
joining walls or floors between two control volumes, and some are exterior walls joining with the environment. The refueling bay ceiling and walls are of sheet metal construction. Estimated masses of equipment contained within the rooms of the reactor buildings and the refueling bay are modeled as steel heat structures. 3.2 Reactor Core Medel The reactor core model consists of input describing the fuel rod l assemblies, the control rods, the top guide, and tha core plate in the core l region; and the control rod housings and guide tubes, and the portion of the reactor sassel lower head below the cora in the lower plenum region. All of these materials may oxidize, melt, and relocate during the core meltdown except the vessel lower head which is modeled to determine its failure and the subsequent release of molten debris from the lower plenum to the containment. The core is nodalized with 4 radial ring cells and 13 axial cells as shown in Figure 3.2 1. The core is subdivided into radial rings sized to match the radial power profile and the resulting volume fractions are listed in-Table 3.21 along with the radial power fractions and power factors. The radial power profile in the center of the core (about 60% of core volume) is relatively flat allowing this portion to be modeled as one ring (actually a cylinder). The core plate dividing the core and lower plenum regions ' is the narrow axial cell 5 shown it, the figure. The active fuel region was divided into 6 equally sized axial cells; axial cells 7 through 12. The axial power fractions for these cellt are listed in Tatie 3.2 2. Axial cells 6 and 13 are non fueled cells containing cladding, canister, and control rod materials but no uranium. The lower plenum axial cells 1 through 4 model the control rod housings and guide tubes; two for the housings and two for the guide-tubes. When the molten core debris accumulates on the bottorn head, the two housing cells and the first guide tube cell will provios an axial temperatute distribution in the debris. Since experience has shown _ that the axial temperature profile for the control rod guide tubes will be relatively flat, the majority of the guide tubes are modeled with one axial cell l (axial cell 3). The smaller upper guide tube axial cell (cell 4) provides i a numerical buffer between the hotter core plate and the- cooler- guide l tubes, which helps to prevent possible numerical heat transfer problems associated with directly connecting a large cell to a small cell. I Core Mar.ses - The core component masses are listed in Table 3.2-3 by axial l The individual radial cell-masses-are-obtained by cell and material type. multiplying these masses by the radial volume fractions. The-lower head mass is not included.in this table _. The total UO2 mass.is 158500 kg,_and the total Zr mass including cladding, canisters, and. fuel rod spacers, end I l
.3-19
O (
?
i t i 4 1 e i Table 3.2 1: Radial Voltune and Power Distribution i i ! Voltune Power Power
- Eing Fraction Fraction Factor-4 i .._
i 1 0.602 0.725 1.204 ; i 2 0.162 0.160 0.985 1 3 0.115 0.076 0.656 -i j 4 0.120 0.039 0.322 3 f 3 i 1
- 1. .
i 1 3 i f i s j Table 3.2 2: Axial Power Distribution l 4 - i j- Power ! Level Eta.ption s 13 .000 4 12 .140 i !' ' 11 .171 ' 10 .178
- 9 .183 i 8 ,188 - .
i 7 .140 ! 6 .000-I f. 4 i j
-1 1
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i j Table 3.2-3: Core Mod 1 Cell Component Masses
- Axial Component Mass, kg cell Engl Glad Etructure Poison Canister i 4 .(UO2 ) (Zr) (St. Steel) (B 5C) (Zr).
i 13 0 3350 .1160 0 2440' 12 26420 5930 .1630 146 4640 i 11 26420 5930 1630 146 4640 i 10 26420 5930 1630 146 -4640-l 9 2b420 5930 1630 146 4640 i 8 26420 5930 1630 146 4640 26420 5930 1630 146 4640 6 0 760 9580 0 560 5 0 0 10700 0 0 lt 4 0 0 9140 0 0 . 3 0 0 19600 0' 0 2 0 0 5660 0 0 1 0 0 9680 0 0 ] Total- 158500 39690 85300 875 '30840 i, Total of All Rings l Total Zr 70530 kg 4 i i t s i j. .) l
~
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l t 00.0 i. i i. , , , , , , 380 44 s 113 ' til i 413 sui _ _ _ - _ _ _ _ _ _ _ _ _ _ l_113 _ _l _ .i _ . m 360.0- ', l
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, i 0.0 1.0 20 3.0 - 4.0 5.0 : 6.0 . 7.0 8.0 9.0 10.0-Radius (ft) 1 L
i
- l. Figure 3.2-1-LaSalle MELCOR Core Map 3-22 L _ , _ _
plugs and springs is 70530 kg. The totn1 stainless steel mass including . the top guide , the control rods, the fuel assembly tie plates, the nose pieces and handles, the fuel support pieces, the core support plate, and i i the control rod-guide tubes, push rods, housings, and stub tubes is 85300 ) kg. The mass of the top guide is 8850 kg and is located in axial cell 13. l Axial cell 5 contains the core support plate with a mass of 9300 kg and i about 1400 kg of other steel, i.e., portions of the fuel support pieces, the control rod guide tubes and push rods. (The nominal thickness of the core plate is 2 inches, but the lumping of the stiffening plates and rods, j and the 1400 kg of other steel into the plate, while holding the plate flow I area constant, effectively increased its thickness te 2.9 inches.) The ; control rods contain 875 kg of B.C poison. i The core contains 764 fuel assemblies and each assembly (as.modeled) has 207.5 kg of UOz, 92.3 kg of Zr, and 10.9 kg of steel (310.7 kg total). The ' core contains 185 control rods and each rod (as modeled) has 79.6 kg of stainless steel and 4.73 kg of B4C poison (84.4 kg total). I Surface Areas - The core surface' areas are listed in Table 3.2 4 by axial cell and component. The individual radial cell surface areas-are obtained , by multiplying these areas by the radial volume fractions. The surface area of the core plate (axial cell 5) includes both top and bottom surface
- areas, the surface area ,f the stiffening bars, and the surface area inside the plate penetrations.
Core Channel and Bypass Flow Areaa.- The core channel and bypass flow areas are listed in Table 3.2-5 by axial cells. The individual radial cell areas are obtained by multiplying thei.e areas by the radial volume fractions, i Eauivalent Diameters - The hydraulic diameters of .0152, .0117, and .0151 meters, for the fuel rod channels, and inner and the outer canister regions, are used for the cladding, and inside and outside canister i equivalent diameters, respectively. The inner canister hydraulic diameter
- is based on the total channel flow area for a single assembly and the-outer canister diameter is based on the bypass flow area for a set of four fuel assemblies and one control rod.
The ' s t ruc ture equivalent diameter is axial cell dependent. The top non-fuel cell (cell 13) which included the top . guide was estimated at 020 m based on the associated masses and areas . The outer canister hydraulic diameter of .0151 m is also used as the equivalent ' diameter for the structure in the active fuel region. The lower non-fueled cell equivalent diameter was estimated at .076 m. The core plate equivalent diameter was estimated at .056 m using the flow. - diame te r through the fuel support picces. The control rod guide tube outer diameter of .275 m is used as the
- e utvalent diameter for axial cells 3 and 4-and the stub tubes and housings c er diameters of .191 m and .152 m are used for cells 1 and 2, rs pectively.
The appropriate equivalent diameter for debris particles is a user
~
specified input and only a single value can be used for each cell. All the 4 3-23
Table 3.2 4: Core Model Cel1~ Surface Areas
- Axial Component Surface Areas, ma Cell Engl- flad . Sintgiqte - Cani s t e r--
13 0 538 194 147= 12 984 -1205 115- 281 11 984 1205 115 281 10 984 1205 115 281 9 984 -1205 '115 281 8 984 1205 .115- 281 7 984 1205 115 281 6 0 56 210 33-5- 0 0 63 0 4 0 0 102 -0 3 0 0 525 0 2 0 0 56 0 1 0 .0 62 0 Total 5904 7824 1900 1868
- Total of-All Rings Table 3.2-5: Core Model Flow Areas
- Axial Flow Areas, m3 Cell . Channel Bvoass Total 13 7.86 2.37- 10.23 12 - 7.86 2.37 10. 2 3 -_.
'11 7.86 2.37 10.23 10 7.86- 2.37 L10.23 9 7.86- 2.37 10.23 8 7.86 2.37 10.23 7 .7.86 2.37 10.23 6 8.81- 2.65 11.46 5 1.91 1.91' 4 6.72J '6.72 3 6.72 6.72 2 14.38 14.38 1 14'.38 14.38
- Total of All Rings 3-24
debris diameters within the core region are set to the nominal value .01 m which is approximately the fuel - pellet diameter. . The debris diameter 'in the lower plenum affects the heat transfer, in particular, the lower plenum boiloff after core plate failure. Thus,- the particle diameter is chosen to yield a realistic bolloff rate rather than using a physically real particle size. (More discussion of this heat transfer can be found in Chapters 5 and 6 which report the results and discuss the associated uncertainties.) The diameter selected for the lower plenum debris size (obtained after some initial calculational attempts) is .05 m which reflects the lack of a cryout heat transfer model. Component Porosities - The porosities (the interstitial volume divided by the sum of the interstitial and component volumes) affects the relocation behavior of debris by deterntning the packing of- debris within the cell. 5 The porosity is specified for both the intact components and the debris for i each axial cell. When the sum of the interstitial volumes and the component volumes exceeds the cell volume then additional core debris from above is not allowed to enter the cell. A high porosity will let less debris enter a cell before the cell is full. The porosities are user specified inputs. The following considerations support the judgements on input porosities used in these calculations. Debris from fuel rod failure is formed as the clad fails by either oxidation or melting. This allows the fuel pellets to relocate as debris before the fuel melting temperature ois reached. Therefore, fuel pellet sized debris (or larger) seems realistic. Other, smaller sized, debris from the cladding, canisters, or control rods is also possible, however, much of the material from these components will melt and candle down instead of-forming debris. Once a molten core pool is formed, molten UO 2 is possible and the debris -particle size distrib> Tion suld change. The fuel pellet diameter is .0104 m so a fuel pellet will.not fall through much of the available free space below. It can fall through the space between the canisters (portions without control blades) or through the very_ centers of the fuel rod channels; but, spaces adjacent to the control rods and smaller spaces within the fuel assemblies are not wide enough to accept a - fuel pellet. Fur her, the fuel assembly grid spacers - (about every 18 inches),_if intact, v 1 prevent fuel pellets from passing. 1 About 42% of the cell volume of each completely intact cell.is occupied by a solid component, meaning that an intact component porosity of .58 or greater will not allow any debris to-enter the cell as long as this cell's components remain intact. The major uncertainties associated with _the intact component porosities _are the debris size distribution and uniformity and whether or not components, including the grid spacers', have failed. Assuming that 1) the_ grid' spacers have failed but that all other components in the cell are intact and'2) that the free space ' which can be filled by debris- includes the centers _ of ' 3-25
]
the -fuel rod channels (the largest circle possible among four adjacent - fuel rods) and the space outside the canisters not containing a control rod, ,. then the intact component porosity is .41. Assuming that intact - grid spacers _ block the free space in the fuel rod channels, so that the only free space availab12 for debris is that outside the canisters,- then the intact component porosity _is .53. The intact porosity values chosen for these calculation are .53_for the in-core potosity which will hold up most of the debris but will still allow some to enter (applied to axial cell 7 through 13) . . The intact porosity for the lower plenum, core plate, and lower non-fueled node (axial cells 1 through 6) are all set to zero for maximum settling'as these cell free volumes are much less narrow. .. If the most dense packing of spherical particles is assumed, then the lower bound debris porosity is .395. A porosity of .4 was selected for the debris at all axial levels. Radiative Exchance Factors.- Five radiative view factors are required by MELCOR to model the thermal radiative exchange within the core and between the core and the surrounding structures. These exchange factors correspond roughly to _the traditional view- factors describing the geometric orientation between two-surfaces. However, temperature' distributions must be considered to correctly calculate the' radiative heat transfer _ _ rates. Also, these exchange factors are for an intact core - geometry and doe not , 1 consider the - ef fect of core degradation on the radiative heat transfer. This is quite possibly an important limitation and a discussion of this_is found in Chapter 6. The first exchange factor is for the radiative exchange from the canister walls to the fuel rod _ cladding. The geometric view factor from the canister to itself was first estimated at .0051. The canister can view itself through the rows of rods and across the corners. .The geometric view factor from the canister to the fuel rods isL then calculated to be .995. However, since the code uses the cell averaged cladding temperatures and
'does not consider the cladding - temperature variation across ' the fuel assembly, the effective exchange factor will- be less than- . 995. An j . effective cladding temperature for -the radiative exchange will be a-temperature closer to the temperature of the outer ring of rods.next-to the canister due to the shielding of the interior rods'by.the-outer rods. A-rough estimate of the effective exchange factor indicated that the factor ranged widely depending upon assumed fuel rod Td canister . temperatures.
Therefore, this view factor.must be treated as a user selected ~ input and ' the value used based on user' judgement. A factor of .6 was selected - for - the view factor. The second exchange factor is for tha. radiative _ exchange from the control rod blades to the ' adj acent canister ' walls. The canister to control . rod geometric view was estimated at .84 from the ratio of canister perimeter next. to the' blades to total canis ter . perimeter. The control rod to canister factor was then calculated at .94 using the reciprocity rt.le. 3-26
r q f h The third exchange factoc is for the radiative exchange radil11y out6ard from the cell bcundary to the next adj ac ent cell or surrounding structure. Since the code uses the cell averaged temperatures to calculate the radiative exchange, instead of effective temperatures at the cell boundary (due to self-shielding effects of fuel rods and canisters), the appropriate view factor will be significantly smaller than if effective temperatures were used. A rough estimate of .14 was made for the effective exchange factor. The f ourth exchange factor is for the radiative exchange axially upward from the cell boundary to the next adj acent cell. This exchange factor will, in general, be quite small for intact core geometries due to large axial cell sizes (about 2 feet) compared to the fuel rod channel dimensions. For axial cells of this size, one cell does not see very far into the next cell so the radiative heat exchange rato is more a function of the temperature near the boundary than the cell averaged temperatures used by the code. User judgement is again needed for the selection of this parameter and after referring to the results of an unpublished study
- on axial radiative heat transfer within a fuel rod geometry, a value of .05 was selected for this exchange factor.
The fif th exchange factor is for the radiative exchange from the liquid pool to the core components. The exchange area for this calculation is the planar ring area and not the component surface area. This radiative exchange is deemed more important when the water level is below the core plate than when it is above because the level usually drops rapidly during the early boil-off and is below the plate during the most significant core heat t rans fe r . Therefore, this exchange factor was estimated for the water level being in the lower plenum. An exchange factor of .96 was selected. Lower Head and Lower Head Penetration - The heat transfer from the debris to the lower head and its penetrations is modeled parametrically using user specified heat transfer coefficients, areas, and masses. The lower head input was developed assuming that failure would occur at a control rod guide tube penetration. The lower head is subdivided into four rings corresponding to the rings of the core model and a penetration is modeled for each of the rings. The penetration is specified to fail when its temperature exceeds 1273.15 K (1000 oC) . Convection heat transfer rates from the debris to the lower head and from the debris to the penetrations are governed by user input coefficients of 1000 W/m 2
-K (default) for each.
- C. J. Shaffer, " MARCH-S1 User's Guide, Appendix E: Emissivities for Axial Radiative Heat Transfer Models," NUREG/CR-2536, SAND 82-0246, Unpublished Draft, May 1982.
l 3-27 i
The conduction heat transfer through - the lower head is ' modeled as being; one-dimensional with 5 axial nodes between the lower plenum control volume-and the drywell pedestal volume. The lower head ring masses and inner-surface areas consider-the spherical shape of the lower head and are Itsted in Table 3.2-6, The penetration is modeled using the lumped mass heat capacity method with convection heat transfer between the debris and penetration and conduction-heat transfer between the lower head and the penetration. .The penetration s mass, the total surface area of penetration contact 1with the debris, and the effective conduction area at the junction of the penetration and the lower head are list in Table 3.2-7-for each radial ring. The masses of the control rod housings and stub tubcs are used for the penetration masses. Upon penetration failure, the initial diameter of the failure opening for ~, ejection of molten core materials was -arbitrarily set at .1 m2 and the - discharge coefficient for ejection of debris through the failed penetration opening is 1. Since the penetration masses and the "other structure" masses for __ the lowest axial core cells are the same masses and the _. penetration masses serve only- to trigger lower head failure, both the penetration masses and areas were reduced in the input to 1% of their values to prevent double
~
accounting of these masses. This way, the punetration temperatures will be computed correctly and - the overall mass and energy _- balancos will be perturbed only slightly and most of the mass is still available for melting and relocation. Solid Debris Eiection Model - Two simple options exist in MELCOR for the ejection _ of _ debris mass from the reactor vessel lower head after a penetration has failed. This option is a user selected input and guidance is not available to su gest that one eption is preferable tolthe other. The option selected is _that: the masses of ' steel, Zircaloy, and UO2 available on the lower head for. ejection are simply _the masses of these materials melted; the masses _ of steel: oxide and control. poison materials available for-ejection are the masses of each of these materials multiplied by the steel melt fraction, based on an assumption of proportional mixing; and similarly the _ mass of Zr02 available for : ejection is the Zr0 2 mass multiplied by the Zircaloy melt fraction. Additionally,'the mass of solid UO 2 available for' ejection is the Zircaloy melt fraction times the mass of UO2 that could be-relocates with the Zircaloy as-calculated in the candling model. Other constraints that-are imposed on the . mass ejected at _ vessel failure . are that: A_ total molten mass of 5000 kg or a. melt fraction of .1-(total mass melted _ divided by total debris mass) is necessary before- debris-ejection can begin. Also, whenever the bottom lower head node exceeds the penetration failure temperature, Bross failure 'of the lower head in that ring is assumed and all debris in the bottom cell is discharged immediately. i 3-28
Table 3.2 6: Lover Head Structure Input Ring Surface Number Hagg Area (kg) (m2) 1 21330 11,77 2 6630 3.66 3 5030 2.77 4 5610 3.09 Total 38600 21,29 Table 3.2-7: Lower Head Penetration-Input Ring Surface -Conduction Number daag Area Area (kg) (m2) _(m2) 1 58.3 .376 .0179 2 15.7 .101 .0048 3 11.2 .072 .0034-4 11.7 .075 .0036" Total 96.9 .624- .0297 3-29
Other Melt Prorression Parameters - Several melt progression parameters were specified as MELCOR user defaults for-the lack of better information. These include the candling heat transfer coefficients (1000 U/m2-K each), candling secondary material transport parameters, _and minimum critical thickness-debris formation parameters (.0001 m each), The BC6 reaction model is . disabled as this model is only important for methane production and associated fission product chemistry (methyl iodide) which is not treated by MELCOR at this time. Disabling the model will not affect a calculation's thermal-hydraulic behavior. l 3.3 Core / Concrete Interaction Model l The core debris that is deposited first in the drywell pedestal of the-containment and. then in the wetwell pedestal af ter ' the failure of the separating cavity floor will attack the pedestal concrete resulting - in concrete ablation and gas generation. The cavity core / concrete interaction models predict these phenomena. Both the drywell and the wetwell pedestal cavities are modeled as cylinders with a flat base. The' inner and outer radii of the pedestal wall-are 3.08 m and 4.559 m, respectively, and the cylinder corner is modeled with a radius of .1 m. The cavity geometry is modeled using 50 rays with their origin at the cylinder center 4 . m _ above the cavity floor. The ray distribution is 25 points across the bottom, 3 in the corner, and the rest up the side. An initial steel mass of 15360 kg is added to the wetwell cavity to simulate the sump tank, gratings, pipes, and supports in . the drywell peder.tal and the concrete steel liners in the l wetwell pedestal which are expected to melt due to heat ' transfer from the core debris. The debris temperature usually exceeds the meltin5 temperature of steel. (In the initial calculational attempts , the drywell pedestal steel was initially added to that cavity, but numerical problems were encountered with activating the drywell cavity. The solution to-the numerical' problems was to initially add the drywell-pedestal steel to the wetwell pedestal steel.) The concrete used in'the construction of the primary containment pedestal is a limestone concrete
- mixed with silica sand and dolomite aggregate.
The Illinois silica sand is over 95% pure SiO 2. The composition of this -- concrete was derived from a construction mixture formula ' consisting ' of
.154, .073, .349,_and .424 cubic yards of coment, water, sand,.and aggregate, respectively, per cubic yard of concrete.
l l-
- Letter from Ceorge R. Crane, Consultant to Commonwealth Edison-Company to Arthur C, Payne of Sandia National Laboratories, "LaSalle County Station Unit 2 PRA: Concrete Data for Primary Containment,"
February 11, 1987. 3-30
In- determining the detailed composition- of the concrete, the sand was assumed to be 100% SiO 2. For the cement, the composition of Type I portland cement (for general construction) from a materials handbooks 'was used (63.2% Cao., 21.3 SiO 2, 6 Al 0 2 3, 2.7 Fe2 0 3, 2. 9 - Mgo , and 1. 8 - S03) . According to the same handbook, the dolomite aggregate is a type of limestone composed of the carbonates of calciun and magnesium (CACO MgCO 2 3) with a minimum of 45% MgCO3 . The calculation of the concrete composition assumed that the aggregate was 50% Ca0 3 and 50t MgCO3. The free and' chemically bound water fractions for limestone concrete were obtained from the CORCON code manual.6 The. S03 , which is not modeled in CORCON, was' treated as CO 2. Af ter reviewing the locations of the steel rebar in the plant drawinga, the amount of steel was estimated to be .056 kg per kg.of concrete. The resulting concrete composition, by mass fraction, is listed in Table 3.3-1. Several physical properties required by the core / concrete interaction I models, taken from References 6 and 7, are listed in table 3.3-2. The liquidus and solidus temperatures are for limestone aggregate-common sand concrete. The ablation temperature was assumed to be the solf.dus temperature plus one-third of the difference between the solidus and liquidus temperatures. 3.4 Combustion Model Combustion within the LaSalle containment during a - severe accident is unlikely because the plant is operated with the containment inerted with nitrogen gas, but combustion within the reactor building is possibic after containment failure. Containment depressurization flows following the failure will mix the hydrogen and carbon monoxide gases produced within the containment with the oxygen atmosphere in the reactor building. Combustion within the containment would require a_ containment failure situation such that oxygen from the reactor building could circulate into the containment after the completion of depressurization. The MELCOR default values -(developed in References 8 and 9) for deflagration ignition, combustion completeness, and propagation into other control volumes were used and are listed in Table 3.4-1. In addition, the oxygen mole fraction must be at least .05 for ignition and a control volume becomes inerted if the combined steam and carbon monoxide mole fractions exceed .55. The hydrogen and oxygen ignition limits were varied for some I of the sensitivity calculations. l The characteristic dimensions for the six ' reactor / turbine control volumes l are: 9.0 m - for the reactot buildings, 24.1 m for the refueling bay,'and l 12.0 m forLthe turbine rooms, The time fractions before allowing a burn to propagate were .25 for the lover Unit 2 . reactor _ building, the refueling bay, and the turbine rooms, and .10 for the upper Unit 2 and Unit I reactor building volumes. 3-31
Table 3.3-1: LaSalle Concrete Composition i i Soecies M_ ass Fraction SiO 2 .36982 TiO 2 .00013 Mno .00005 Mgo .09214 Ca0 .22255. Na20 00059 K0 2 00148 , , Fe2O3 '00210 j Al 203 .00902 i i Cr203 .00002 1 CO 2 20173 l Fe (Rebar) .05600 i H2 O (Free Vapor)- .02549 H2 O (Chemically Bound) .01888 4 Table 3.3-2: LaSalle Concrete-Properties-l- Density 2340 kg/m3 . Solidus Temperature 1420 K Liquidtis Temperature 1670 K Ablation Temperature. 1503 K Initial Temperature 305.K Emissivity .82 i Table 3.4-1: Combustion Parameters
~
Mole Fraction Values . Combustion Event Hydrogen Carbon Monoxide Ignition Limit .10 ,167 Complete Combustion .08 .148 Upward Propagation .041 .125 Horizontal Propagation .06 .138: Downward Propagation .09 .150-4 4 3-32
3.5 Decay Heat Power Madel The decay heat power is modeled using the user specified input option and implemented as a time dependent table. The reactor operating power at the time'of reactor shutdown is 3323 MW of thermal power. A comparison
- of the ORIGEN code 10 results and the ANS standardit found that the decay powers from ANS standard-calculations were substantially higher than those from the ORIGEN calculations. Several aspects of the differences between the calculations were discussed and it was' concluded that the ORICEN results wert, the most realistic. One such aspect was that a comparisonia of THOR-I code predictions with predictions from the ANS standard for a realistic LWR operation cycle indicated that the - ANS '
standard systematically overestimates the decay-power by 6 to 8% in the first 10' seconds af ter reactor shutdown. Furthermore, the ORIGEN code results agree well with the results from the THOR I code. The ORIGEN code results. are deemed more- appropriate for the LaSalle calculations than the ANS standard, since the first obj ective of _ the l working group in the 1979 revision of the ANS standard was to develop a standard appropriate for LOCA applications (cooling times up to '104 seconds) and was not intended for use in realistic estimates of decay power over the extended time periods characteristic of severe accidents. The ANS standard-covered the longer cooling times (up to 108 seconds) by using an upper bound for the neutron capture effect. l A decay power model 'which employs features of both the ANS and the ORIGEN t models was developed ** as a compromise between the models. This model'was developed from the ANS's detailed method by employing parameter _ . values calculated by the ORIGEN code. Tim parameter values included the 2390 production rate, the irradiation time, the time averaged values - for - fuel' composition (235U, aseU , 23sPu) and the recoverable energy per fission for . each fuel component. Then the scaling- factor of .9.47 was _. applied to this model such that the power curve for this model touched the curve for the ORIGEN code results at their closest approach. The results of these models as obtained from the MELCOR code' are compared in Figure' ' 3. 5 - 1. The compromise model yields lower powers similar to the ORIGEN code results but retains features of the ANS standard. r !-
- F. E. Haskin, Sandia National. Laboratories Memorandum, "Whole-Core Decay Heat Power," January 28, 1986.
l. I
- ** Paul Rexroth, Sandia National Laboratories Memorandum, " Proposal for a Revised Default Makeup Decay. Model for CONTAIN,'" May 8,1986.
l i' l-l. 1 3-33
)
R, 0.060 3 , , ,,,,,y , , ,,,,,y , , ,,iiin i . ,..y i , , , i ,.y ,i,...,
\
\
\
0.054- \ -
\
\
\
LaSalle - Compromise ome- - '
- -- MELCOR ANS Standard -
~ ORIGEN 3
0.042- '\'
\
v 0.036- N - 6 \ O \ W 0.030- N - e 'N
$ N O ', \
0.024- s - o \ Z ,\ s 0.018- 'N -
\
N N 0.012- ' ', s
.N,
.,s N 0.006- '- . ' %
0.000 , , , ,,,,,, , , ...m, , , , , , , , , , , , , ,,,,o , , . . ...y , , . . . . . . Id id id 18 Id 18 id Time Atter Reactor Trip ( Sec) Figure 3.5-1 Comparison of Decay Heat Power Models 3-34
l
^t l e t
n 3.6 Radionuclide Transport Models In addition to the 15 MELCOR default radionuclide classes (listed in Table 2.7-1), an optional sixteenth class is formed _b y chemically combining cesium and iodine to form cesium _ iodide (CsI)._ The chemistry model employed in forming cesium iodide simply assumes that; the new releases of cesium and iodine, with each time step, are combined until one of the-two is depleted. The release of fission products from the fuel within the reactor vessel is modeled using the modified CORSOR model without the surface to volume ratio option. The modified CORSOR model was chosen because it is a more recent model than the standard CORSOR model, therefore, presumably better. The MELCOR default coefficients, obtained from the CORSOR's user manual,n are used for clat.cas 1, 3 through 6, 8, and 12. The coefficients for the other , classes are input as described below either because their default- values are zero or because the use of the de f ault -- value s resulted in unrealistically rcpid releases of fission products or structural materials (discussed in Chapter 5). The model release rate correlation and coefficients used are listed in Table-3.6-1. The leading coefficient of the class 2 default values was reduced by one order of magnitude to reduce its release rate. Since the CORSOR's User Manual gives release coefficients for iron, but not for molybdenum, and iron is a member element of class 7 and iron aerosols are formed from structural steel in addition to the fission products; the coefficients of iron are used for class 7 but with the leading coef ficient reduced by three orders of magnitudes. The coefficients for class 9 are set equal to those for class 10 and the coefficients for class 11 are set equal _to those for class 12 by assuming that the release behavior of lanthanum is similar to that of uranium and that of cadmium is similar to tin, Then, the leading default coefficient for class 10 was also reduced - by - three orders of magnitude. Aerosols formed from the boron in the control rod poison would , have made up class 13 but was not modeled. The release of fission products trapped within the fuel rod cladding gaps are modeled using data from the CORSOR's user manual. The temperature criterion for cladding failure is 1173 K. The user specified gap inventory, contains 5% of the initial core inventory of cesium, 1.7% for iodine, 3% for xenon, 0.01% for tellurium, 0.0001% for barium, and zero for the-remaining classes. The aerosol dynamics is modeled with five aerosol section sizes with-diameters ranging from .5-to 512 microns resulting in section dimensions of
.5 to 2, 2 to 8, 8 to 32, 32 to 128, and 128 to 512 microns. The number of sections was limited to five due to calculational run speed considerations, but it has since been noted that the smaller diameter of- .5 microns is probably too large and this should be considered a -source of uncertainty. _i The size range probably should go down to about .01 microns.
3-35
Table 3.6 1:;CORSOR Release Coefficients-Class Coefficient Number ko 0 1 2.00E5 63.8 2 2.00E4 63.8 3 2.95E5 100.2 14 2.00E5 63.8 5 2.00E5 63.8 6 1.62E6 152.8 7 2.94E1 87.0 8 2.67E8- 188.2 9 1.46E7 143.1 10 -1.46E4 143.1 11 5.95E3 70.8 12 5.95E3 70.8 Release Rate (fraction / min)_ = k exp (-Q/1.987E-3/T) where T - Temperature 3 36
The models describing the dynamics of aerosols _ require the input of the ! solid density for the aerosol particles which is used for all particles , regardless of the particles composition. An aerosol particle may be a I conglomerate of the various . fission products, _ fuel, and structural materials. For instance, if all of the aerosol particles consistod of solid UO 2 only, then the appropriate density would be the solid density of UO 2 (10960. kg/m3) . However, the appropriate density _for a changing conglomerate aerosol associated with a core meltdown is a difficult number to realistically select and is another source of uncertainty. The nominsi solid density of 1000 kg/m3 is used. MELCOR default values are used for constants required by the aerosol dynamics models, such as the particle shape factors and slip coefficient. The aerosol coefficients were calculated for gas pressures from lE5 to 2E7 Pa and from gas temperatures from 273 to 2000 K. The vapor pressure curves were-redefined for classes 3, 6, 7, 8, 11, and 12 l. which are defaulted in MELCOR - as always existing in the aerosol form. l Since classes 3, 6, 7, 8, and 12 exist as an aerosol except at very high temperatures, these curves are not expected .to _ result in significant differences in the radionuclide transport results from those that would be obtained with the default values. Clas.= 11 (more volatile main group), however, has a pronounced vapor behavior at the temperatures associated with core meltdown. 3.7 References
- 1. "LaSalle County Station Final Safety Analysis Report," Commonwealth Edison Company, Chicago, IL, 1983,
- 2. "LaSalle Systems Training Manual," Commonwealth Edison Company,
! Chicago, IL, 1985.
- 3. L. A. Kent, " Water Released From Heated Concrete," NUREG/CR-2279, l SAND 81-1732, Sandia National Laboratories, Albuquerque, NM, March 1982.
- 4. S. E. Dingman, C.J. Shaffer, A.C. Payne, and M.K. Carmel, "MEILOR l Analysis for Accident Progression Issues,"_NUREG/CR-5331, SAND 89-0072,
! -Sandia National Laboratories, Albuquerque, NM, January 1991. I j 5. G. S. Brady_and H. R. Clauser, " Materials Handbook," Eleventh Edition, E McGraw-Hill Book Company, New York, NY, 1979.
- 6. R. K ._ Cole, Jr., D. P ~. Kelly, and M. A. Ellis, "CORCON-Mod 2: A Computer Program for Analysis of Molten-Core Concrete Interactions,"
-NUREG/CR-3920, SAND 84-1246, Sandia National Laboratories, Albuquerque, NM, August 1984.
3-37
- 7. D. A. Powers and F. E. Arellano, " Large - Scale , . Transient Tests of the Interaction of Molten - Stoel with Concreta," .NUREG/CR 2282, SAND 81 1753, Sandia National Laboratories, Albuquerque, NM, January 1982.
- 8. G. G. Weigand. et al., " Thermal Hydraulic Process - Modeling in Risk
/.nalysis: An Assessment of -- the - Relevant Systems , -Structures, and.
Phenomena," NUREC/CR-3986,- SAND 84-1219, Sandia National Laboratories, - Albuquerque, NM, August 1984
- 9. S. E. Dingman A. L. Camp, C. C. Wong, D.- B. King, and R. D. Casser, s
"HECTR Version 1.5 User's Manual," NUREC/CR-4507, SAND 86-0101, Sandia National Laboratories, Albuquerque,-NM, April 1986.
- 10. D. E. Bennett, "SANDIA-ORIGEN User's Manual," NUREC/CR-0987,-
SAND 79-0299, Sandia National Laboratories, Albuquerque, NM,-October 1979.
- 11. American Nuclear Society Standards Committee Working Group ' ANS-5.1, "American National Standard for Decay Heat Power in . Light Water Reactors," ANSI /ANS-5.1-1979, American Nuclear Society, .La Grange
! Park, IL, 1979,
- 12. J. Metzinger and H. V. Klapdor, "New Results on the Decay - Heat' of Nuclear Reactors ," International Conference on Nuclear Power = Plant Aging, Availability Factor and Reliability Analysis,. San Diego, CA, 8-12 July 1985,
- 13. M. R, Kuhlman, et. al., "CORSOR User's Manual," NUREG/CR-4173, BMI-2122, Batte11e's Columbus Division, Columbus, OH, March 1985.
f 3 38
4.0 SEQUENCE DESCRIPTION The severe accident sequences dominating the frequency of an accident resulting in core damage were selected as candidates for a MELCOR calculation. These dominant sequences and nomenclature are discussed.in-the PRA Level I report.1 A MELCOR calculation of an individual sequence sometimes is applicable to other accident sequences as well. As many y calculations were run as time and resources permitted and their order was ! in accordance with their dominance and applicability to other sequences. I Base case calculations were run for feur dominant sequences: _a high- , pressure short-term station blackout sequence (T63), a low-pressure short-term station blackout sequence (T100), an intermediate term station blackout sequence (T62), and a long-term station blackout sequence (T24). The T100 seqt.ence, which contributes 35.4% of the mean total core damage frequency, is the dominant accident sequence at LaSalle. The T24 calculation also applies directly to the T5 T10 and T17-T23 sequences or well. Likewise, the T62 catsulation applies to .the T35-T61, T67<75, T79 T87, and T91-T99 sequences. After the-T63 calculation was run,-it was deleted from the dominant accident list, but -the second most dominu t-sequence involves a control room cabinet fire (17.2% of total frequency) ' which results in-an accident very similar to the T63. Several sensitivity calculations were also run . to - - inves tiga t e the sensitivity of the release ~of radioactive materials to the. environment to certain paremeters such as the containment failure area. 4.1 High-Pressure Short-Term Statit Blackout (T63) The T63 sequence involves a transient initiator followed - by successful reactor trip, successful opening and reclosing of the SRVs, failure of all high- and low-pressure injection, and failure -to depressurize the primary system. Other than the failure of the ADS to depressurize the primary system, the T63 sequence is similar to the T100 sequence. The . dominant cut sets have a loss of.offsite power initiator.followed by loss of onsite power to the safety bt.ses caused ..by . common mode failure of . the diesel generators resulting in a station blackout. RCIC fails either immediately or delayed and core damage results bafore injection can!be- restored. One cut set of the T63 sequence was selected- for the . base case calculation. Twelve sensitivity calculations were run to investigate the-effects that the containment failure size and location and - combustion - ignition conditions have-on the transport to fission products. 4.1.1 Base Case At the initiation -of ' the high-pressure short-term -station blackout calculation, all onsite AC and DC power failed. . Subsequent to power failure, the reactor tripped and isolated, and all ECC - systems , containment heat removal capability, and the ADS also failed. 4-1
The failure of all ECC systems resulted in the boil off of the reactor. vessel water, exposed the reactor core, allowed the fuel to heat and ultimately to melt and. relocate to the vessel's lower plenum. The failure of the ADS resulted in the reactor vessel remaining at high pressure with pressure relief through the - safety / relief . valves until _ lower head penetration occurred due to the interection of the high temperature core , debris with the control rod housing and stub tubes. Immediately ef ter vessel . f ailure , the primary system depressurized' into the drywell pedestal, and, after a delay, the core debris became fluid enough to pour out onto the drywell pedestal floor. 3 Twenty minutes after the Cire d-bris was ejec ad from the reactor vessel, the two four-inch diameter drain pipes passing through the pedestal floor 1 failed (refer to Chapter 3). Subsequently, the debris passed through the ) holes to the wetvell pudestal. The containment pressure increased during this process and continued to increase due to steam and noncondensible l gases produced by the interaction of core debris with pedestal concrete-and by concrete degassing. The containment failed-due to overpressurization at the estimated failure , pressure of 195 psig with a leak in the wetwell wall located about midway between the suppression pool surface and the ceiling. The containment 4 then depressurized into the reactor building resulting in - releases of radioactive - materials to the environment. Hydrogen ' accumulation in the
- reactor building created flammabic conditions resulting in-the prediction of hydrogen burns which increased the ' amount of radioactive materials released.
4.1.2 Sensitivity Cases 1 i Twelve sensitivity calculations were done for the high-pressure short-term station blackout sequence from a restart written during the base case calculation at the time of predicted wetwell wall failure. The varied parameters were the containment failure location,- area, and flow l resistance and the combustion minimum ignition mole . fractions. These cases are listed in Table 4.1-1. d The containment failure locations included the wetwell wall- about midway between the suppression pool surface and the ceiling, the- drywell vall near the drywell floor, and the drywell - head. The flows through the drywell head went directly to the refueling bay from the containment and the wetwell and drywell wall flows entered the lower portion of the reactor building. Since the flows to -. the refueling bay bypass the majority of the reactor building, the decontamination of these - flows by-the building was less than for flows _ entering lower in the building. The amounts of radioactive materials leaving the containment from the drywell-differed from the those leaving the wetwell. The effect of containment failure location on radionuclide- transporc is _ shown through ' the cocparisons- of cases 4, 7, and 8; cases 2, 5, 10; and cases 1, 3, 9, and' 9A in Chapter 5. 4-2
Table 4.1-1. Sensitivity Cases for the High-Pressure Short-Term Station Blackout Sequence
, i Case Breah Break Area Minimuta Ignition i Number Location Area k+f1/d /SQRT(k+f1/d) Mole Fractions ft2 ft2 m2 g, o, Base Wetwell Wall .1 136.6. .0086 .0008 .10 .05 i 1 Wetwell Wall 1. 1.011 .995 .0924 .05 .05 l 2 Wetwell Wall .1 1.011 .0995 .00924 .05 .05 3 Drywe'11 Wall 1. 1.011 .995 .0924 05 .05 1 4 Wetwell Wall .1 136,6 .0086 .0003 .05 ,05 l 5 Drywell Head .1 1.011 . 09R5 .00924 .05 .05.
! 6 Wetwell Wall 1. 1.011 .995 .0924 1. .05 7 Drywell licad ,1 136.6 .0086 .0008 .05 .05 8 Drywell Wall .1 136.6 .0086 .0008 .05 .05 9 Drywell Head 1. 1.011 .995 .0924 .05 . 05 9A Drywell Head 1. 1.011 .995 .0924 .05 .05/.04 10 Drywell Wall .1 1.011 .0995 .00924 .05 .05 11 Wetwell Wall 10, 1.011 9.95 .924 .05 .05 i l
.1 1
l l u l l l i 4-3 L
The flow areas used were .1, 1, and 10 ft2 with two different flow resistances for the smaller area. The flow areas and resistances were combined by dividing-the areas by a term to account for the ef fect of the form and friction losses, i.e., the area was divided by the square root of the sum of: (1) the loss coefficient and (2) tha characteristic friction factor multiplied by the flow length and divided by the hydraulic 4 diameter. These resistance adjusted areas are .0086, .0995, .995, and U' 9.95 ft 2. The effect of the flow area and resistance on radionuclide "' transport is shown through the comparisons of caras 1, 2, 4, and 11; cases 3, 8, and 10; and cases 5, 7, 9, and 9A in Chapter 5. The minimum hydrogen and oxygen mole fractions required for a deflagration - burn were varied in two cases. The hydrogen mole fraction in case 6 was set high to completely prevent combustion during the entire calculation. A comparison of cases 1 and 6 will show the effect of reactor building burns on radionuclide transport. All of the other sensitivity cases used
.05 for the hydrogen mole fraction but the base case was run --ing a mole fraction of .1. The .05 mole fraction was deemed more applicable af ter the base case was completed. A comparison io made for the base case and sensitivity case 4 in Chapter 5. In case 9A, the oxygen frar. sn was reduced on a resta-t during the calculation from .05 to .04 to iraciate a burn that otherwise would not have occurred. A comparison of cases 9 and 9A, in Chapter 5, illustrates how a small change in input had a large effect on the transport of radioactive materials to the environment.
4.2 Low-Pressure Short-Term Station Blackout (T100) The T100 sequence, as described in the Level 1 report, involves a transient initiator followed by successful reactor trip, successful opening and reclosing of the SRVs, failure of all high pressure injection, successful depressurization of the primary system, and failure of all low pressure inj ection. The failure of injection can be either -immediate or delayed depending on the particular cut set; however, the dominant cut sets have immediate failure. The dominant cut sets have a loss of offsite power initiator followed by loss of onsite power to the safety buses , caused by common mode failure of the diesel generators resulting in a station blackout. RCIC fails either immediately or delayed and core damage results before injection can be restored. An immediate failure cut set of the T100 sequence was selected for the base case calculation. One sensitivity calculation was run to investigate the potential impact of a failed lower pedestal wall due to debris concrete interaction. 4.2.1 Base Case At the initiation of the low-pressure short-term station blackout calculation, all onsite AC power failed with the subsequent trip and 4-4
isolation of.the reactor, and failure of all ECC systems and containment heat removal capability. The ADS continued to function normally N this sequence with DC power supplied by the batteries. This sequence progressed.in a similar manner to the high-pressure short-term station blackout sequence. except that the ADS (with the operator control procedures simulated) slowly depressurized the reactor vessel at the rate that corresponds to a decrease in the saturation temperature of 100 oF per hour until a low core water level indicator caused a rapid completion of the depressurization. This resulted in a low pressure core meltdown, a low pressure lower plenum boil off, and a low pressure vessel failure. 4.2.2 Sensitivity case One sensitivity _ calculation was done for the low pressure short-term station blackout sequence from a restart written during the-base case at the time when failure of the lower pedestal-vall was predicted. The wall was predicted to fail when -the radial concrete ablation exceeded the initial wall thickness. The sensitivity calculation allowed suppression pool water '.o enter the pedestal and make contact with the core debris. The additional steam generated from this contact accelerated containment failure and increased the transport of radionuclides to the environment. 4.3 Intermediate Term Station Blackout (T62) The T62 sequence, as described in - the Level I report, has a transient initiator followed by successful reactor trip and SRV operation. All high pressure inj ection except RCIC fails .and containment and primary system heat removal fail. The ADS system works but the. low pressure systems are failed. The overall time available_ to the operators to perform their recovery actions is approximately 2 hours from the time that RCIC-fails. In some cases (e.g., restoring offsite power when a diesel generator has run for some period of time) more time is available._-The amount of time available depends on the failures _ that constitute . the cut- set and what recovery action is being considered. A cut set of the T62 sequence involving about 9 hours of cooling was selected for the base case calculation. The llPCS models were used instead of explicitly modeling the. RCIC -system since - the llPCS model could-- be - adjusted to- simulate the response of =.RCIC . for this sequence and also be used . to ' model -. the long-term sequence. At- the initiation Lof the intermediate-term station blackout calculation, the reactor tripped and isolated and all ECC failed except llPCS which operated normally until: after the ~ suppression pool-became saturated. Also, the -- ADS operated-normally; but all containment heat removal systems-failed. I IIPCS injected water into the reactor vessel upper' plenum. with the water supply. coming from the CST until the suppression pool . level exceeded r 4 . .. . . _. ..
two inches above ' the normal level, then the water was pumped from the suppression pool. The HPCS flow rate.was governed by the reactor vessel water level. The ADS reduced the primary system pressure at a rate corresponding to'a decrease in the. saturation temperature of 100 'F per hour until the system was fully depressurized. At 9-hours, ADS-and HPCS (i.e., RCIC) fail due to battery depletion and loss of DC - power, The primary ' system ' then repressurized until the pressure was relieved through the safety / relief valves. The primary system water boiled away, followed by core meltdown, lower head failure, and debris ejection. During the repressurization of the reactor vessel, the entire vessel water mass (except the water trapped in the recirculation piping) heated as the saturation pressure increased, significantly delt '.ng the time to core meltdown. - Natural -circulation' ] , flows kept the water well mixed, j i 4.4 Lonn-Term Station Blackout (T24) i The T24 sequence, as described by the T18 sequence in the Level I report, also has a transient initiator followed by successful reactor trip and SRV operation. The-main feedwater system fails but HPCS and one train of the CRD system work providing high pressure injection. The normal containment and primary heat removal systems fail, and venting fails. -Containment pressure increases until a leak develops. Depending-upon its location, this leak will produce an environment which could -cause injection. systems that are operatin$ or that = may be able to operate to fail. The overall time available to the operators to perform _ their recovery actions . is approximately 27 hours. In some cases - (e. g. , venting). less time is available. The amount of time available depends on the ' failures- that constitute the cut set and what recovery action is being considered. A cut _ set of the T24 sequence involving successful core. cooling until the containment failed was selected for the base case calculation. At the initiation of the long- term .- s ta tion blackout calculation, the L reactor tripped and - isolated and all ECC failed except HPCS which operated nc.rmally until the containment failed by .overpressurization. Also, ADS operated normally, but all containment heat. removal systems failed. (This calculation was identical to the intermediate-term station blackout calculation up to nine hours). The HPCS system in s ected water into the reactor vessel upper plenum with the water supply coming from the CST until the. suppression pool ~' level exceeded two inches above the normal level, then water was pumped from the suppression pool. The HPCS flow rate was governed. by the reactor vessel water level. The ADS system was used to reduce the primary system pressure .at a rate corresponding to a decrease- in the - saturation temperature of 100 oF per hour until the system was fully . depressurized; but ' then, - the ADS failed when the drywell pressure exceeded 100. psia and remained failed. When tha . ADS failed, the primary system repressurized until the pressure was relieved through the safety / relief valves. 4-6
1 The core was maintained completely cooled until containment failure caused
~
the failure of HPCS, then the water boil-off in_the reactor-vessel began. The core meltdown and vessel failure occurred at high pressure. The containment depressurized to the reactor building through a wetwell break of .995 - f ta and was completely depressurized when the reactor vessel failed. With the containment already failed, the transport of radioactive materials to the environment was caihanced as the reactor vessel depressurized. 4.5 References
- 1. A. C. Payne, Jr., Analysis of the LaSalle Unit 2 Nuclear Power Plant; Risk Methods Integration and Evaluation Program (RMIEP),
Volume 1: Summary," NUREG/CR-4832, SAND 92 0537, Sandia National Laboratories, Albuquerque, NM, March 1992.
~
c 4-7
.l 5.0 RESULTS l The calculational results are discussed in this chapter. A detailed discussion of the uncertainties associated with - these calculations is found in Chapter 6. The results of the high-pressure short-term st&. tion blackout calculetion are discussed in more-detail than the results'of the i
other calculations- to demonstrate the use . of the MELCOR models and their capabilities. The remainder of the calculations are then- presented - as variations of this calculation. The reported results concentrate on the results most pertinent to the PRA. These results. include- (1) the timing of events occurring as the accident sequence unfolds (e.g., the beginning of- core damage, the failure of the reactor vessel, and the failure of the containment),- (2) selected- thermal and hydrodynamic results at locations throughout the - plant (e. g. , pressures, temperatures, and interactions of water and core debris producing steam), (3) the-generation of hydrogen and other noncondensible gases from the oxidation of structural metals and from the interaction of core debris with concrete. (4)- results indicating the possible failure of structural components (e.g,, the reactor pedestal wall from ablation and degassing caused by molten core debris), (5) reactor building combustion results relating to possible equipment - failure '(e.g., the_ peak burn pressures and temperatures and combustible gas - mole _ fractions), (6) the release, transport, and deposition of fission products throughout. the plant, and finally, (7) the fission product time-dependent source terms to the environment. An overview is presented for each calculation both as an introduction and as a summary for readers not interested in the details. -For some, reading the overview and flipping though the tables and . figures will be sufficient. The detailed results will be the most useful to people-performing severe accident calculations, or interested the performance of particular MELCOR models. Accident analysts using these MELCOR results should be aware that the core meltdown progressed- differently than the typical meltdown calculated by one of the less detailed predecessor codes. A sampling of the results is included from each of the many facets of the complete severe accident calculat;" The detailed results will also aid-those wishing to assess the_credibtlity of these calculations. Because of the large number of figures and tables and to ease comparison L between the different calculations, all the figures and tables for each l calculation are presented at the end of each section. 5.1 High-Pressure Short-term Station Blackout 5.1.1 Base Case 5.1.1.1 Overview The results for the high-pressure short-term station blackout calculation with failed emergency core cooling systems, automatic depressurization 5-1
system, and all containment _ heat removal capability, are discussed here. A detailed description of this sequence is found_in Section 4.1.1 and the progression of the accident is illustrated with the list of key events and the timing of those events found in Table 5.1-1. After the accident was initiated, the core decay heat caused boiling in the core region, generating steam and further pressurizing tha_ primary system. This pressure was relieved by the safety / relief valves _ which cycled open and closed. Steam and hydrogen- flow through tho valves was piped into the suppression pool whet, virtually all of the steam was condensed, heating the pool. The vess.1 water level (collapsed) dropped to the top of the active fuel at 0.59 hours at-which' time core uncovery began, followed by heating of the exposed fuel. The first fission products were released from the fuel rods at- 1.26 houn when the cladding was predicted to fail in the inner core (ring 1). The 1 first cort material relocation (by melting and candling) occurred in the l upper central portion of the core at 1.48 hours and the first core channel blockage -(also ring 1) at 1.79 hours . As_the core meltdown progressed, molten and solid debris accumulated on the. core plate, heating the plate to its specified failure temperature of 1273 K, thus causing the failure of the plate at the center at 3.88 hours, This debris then fell into the vessel lower plenum, which was nearly full of water, generating a large quantity of steam as the - debris partially quenched. The primary system pressure was relieved by the safety / relief valves as the lower plenum water quickly boiled away. The lower plenum pool boiled away 25 seconds after the core alate center failed. The partially quenched hot debris, deposited on.the lower head, heated the lower head and penetrations causing a lower head penetration to reach its estimated failure temperature of 1273 K at 3.92 hours (156 seconds after core plate failure). The primary system depressurized rapidly _ into the drywell pedestal after the penetration failed. The core debris then continued to heat up and whet. the required molten condition was reached: the debris poured out of the vessel onto the drywell pedestal floor at 4.08 hours. During this time, the remainder of the core - continued to degrade as indicated in the list of events, The drywell pedestal floor drain pipes- failed . at 4.41 hours (20 minutes af ter the first debris ej ected, see Section 3.1.2) with the subsequent passage of this debris through the holes into- the _ wetwell pedestel.- The ejected debris attacked the pedestal concrete producing gases which contributed to the pressurization of .the containment. . Debris ejection continued, af ter the first ejection at 4.08 hours, _ with the_ final ejection -
~
from the outer ring at_9.48 hours (over a period of 5.4_ hours). Portions of the outer ring reme ced in the core _ including one third of the outer ring fuel. The total in-vessel hydrogen produced by oxidation was .1138 kg. 52
i-The containment exceeded 60 p-( (the venting pressure criteria)-_at 6.13
- hours and continued to pressurize,-_ reaching pressures of 85 -(ADS failure -
,- pressure),145 (lower limit of containment failure pressure), and 195 psig (median containment failure pressure) at 8.14, 10.30, and 18.18 hours, j respectively. When the wetwell reached 195 psig pressure, the containment failed with a leak through the wetwell vall about midway between the
- . suppression pool surface and the ceiling. Degassing of concrete within
- the containment contributed significantly to the - pressurization of - the
containment with the majority of the gases produced ftom the pedestal wall j above the surface of the molten debris pool within the wetwell; pedestal. 4 The suppression pool remained subcooled throughout the calculation but was i gradually approaching the boiling temperature of unpressurized water (100 l oC) at the end. The suppression pool might have boiled due to the decay. i energy of the fission products deposited within the pool, had the calculation been extended, with implications of enhanced fission product transport from the containment later in accident. The flows from the containment depressurization entered the reactor ! building in the upper Unit 2 volume. Two deflagrations of hydrogen and carbon monoxide occurred in the reactor building during the containment depressurization, with further burns possible beyond the termination of the calculation. The calculation was terminated beiore the j depressurization .was completed at 32.14 hours because of the slow .run speed caused by convergence problems in the heat structure and 4 j core / concrete interaction models. i The maximum radial' penetration calculated by the core / concrete interaction models exceeded the wall thickness of the wetwell pedestal wall at 19.16 hours (about 1 hour after containment -failure). This implies - the ' possibility of structural failure - of the pedestal and support for the reactor vessel and the shield wall and possible interaction of the suppression pool with the molcen debris either by water flowing into the l pedestal or the debris flowing into the suppression pool. If the pedestal . wall were to fail before. the containment failed and depressurized, then, the impact-of the potential relocation of structural components and/or the
- pressure produced by steam generated from the interaction of_ the ~ core_
i debris and water, could be a factor in the timing, location and' severity
- of - the containment- failure. Mul tiple ' failure locations, for instance,
- might allow a natural circulation loop to develop between the containment '
- and the reactor building after the depressurization completed.
4 The radionuclides released- and transported included the - radioactive, fission products released from the fue'l as a vapor or _an aerosol,
- . aerosolized radioactive uranium fuel, and non-radioactive aerosolized E ' structural materials. Theses materials - were released both in vessel-during the core meltdown and ex-vessel from the core / concrete interaction.
The first in-vessel releases occurred with the failure of the inner ring cladding at 1.26 hours and the ex-vessel-releases after the first debris ejection at 4.08 . hours . The aerosolized - non-radioactive structural materials include steel,-zircaloy, and concrete. For some_ classes (e.g., xenon) most of the fission products were released from the fuel, while for ! others-(e.g., cerium) only a small fraction of their initial inventory was released. These results are found in Section 5,1.1 6. ] 5-3
', l.1.2 Primary system flydrodynamics The primary system pressure for the hi6 h pressure short term statha blackout sequence remained high as shown in Figure 5.1 1 until the pressure integrity of the reactor vessel was losc due to the failure of a penetration in the lower head. Prior to vessel failure, the pressure was relieved through one safety / relief valve opening at 6.936 HPag (1006 psig) and reclosing at 6.178 MPag (896 psig), except during the rapid boil off -
of the lower plentun water immediately af ter the first core plate failure, which opened all 18 valves. The one valve cycled open and closed a total of 72 tirnos during the calculation (but the plot frequency of every 100 seconds during the core meltdown was too infrequent to exactly reflect the valve operation in the figure). A pressure spike of 9 MPa (1300 Psia) occurred at 3.88 hours (13980 seconds) which was caused by the rapid heat transfer and steam generation in tha lower plenum from the contact of the molten core debris with water. The accuracy of this pressure spike is quite uncertain because of the lack of a dryout heat transfer model as noted in Section 3.2. A discussion of the uncertainty in the inagnitude of this pressure spike is. found in Chapter 6. After the primary system depressurized through the lower head to the drywell pedestal, the primary system pressure followed the contairunent pressure for the remainder of the calculation. The partial pressure for hydrogen that was produced by ox'.dat!on within the core is ' also shown. The hydrogen inole fraction reached as high as .97 in the core channels at 1.8 hours (6400 seconds) during the meltdown. The reactor vessel water levels (uncc11apsed) are shown in Figure 5.1 2. The level at the beginning of the accident was within the upper plesnun/ separators and downcomer control volumes. The water level dropped below the top of the activo fuel at 35.3 minutes (2117 seconds). (In the figure, when a level drops to the bottom of a control -volume, the curves-remain at the lower elevation of that volume.) As the core was further exposed,=the downcomer level followed the core levels until the level dropped below the. elevation of the jet pump nozzlea , af ter which, the remaining downcomer water became trapped until it was raporize$ by either flashing or heat transfer through the core shroud. The downcomer eventually dried out at 12.8 hours (40.90 seceae) and se vaporization of the downcomer water contributed ta th.o cont 6 .na pressurization. The core volumes dried out at 2 .1 ours but ni reflooding did occur twice ' af terwards; once af ter the fi. cst opening n a safety / relief valve caused downcomer and lower plenum water to flash and once after the core plate failed when rapid boiling occurred in the lower - plenum. Most of the water in the recirculation loop (5.5% of initial primary system water mass) remained at the end'of the calculation. However, this water did make a minor contribution to this calculation because 14.1% of the water in the recirculation loop piping flashed during vessel 54
- i,r
--- w--1 -m. w .- rv - .v -- .s
depressurization, slowly vaporized due to heat transfer through the pipe 3 wall, or leaked through the pump seals to the drywell pedescal. The pump ; seals leaked 1570 kg of water by the time of vessel failure with an additional 1200 kg leaked by the end of the calculation. In addition to the pump seal leakage, the control rod drive seals leaked 1910 kg before vessel failure. This leakage was much more important in the long term station blackout calculation than in this one since more water accumulated i I in that calculation. Since the lower i enum volume included the jet pump diffusers, its water level fluctuated between the elevations of the jet pump nozzles and the bottom of the active fuel, until the level drapped below the .attom of the active fuel, at which time the level became teore stable. The level then , decreased slowly until the core plate failed and dropped core debris from the inner ring (containing 60.26 of the core) into the lower plenum water, i after which the lower plenu'1 water quickly boiled away. The time elapsed I between the core plate failure and the dryout of.the lower plenum was only 24.3 seconds. This tirue interval - was dependent upon the input selected for modeling the lower plonum heat transfer and 19- quite uncertain as will , be discussed in Chapter 6.
'l primary system ' integrated flow rates are shown in Figures 5.1 3 and 5.1 4 for steam and hydrogen flows, respectively. The flow pattern dtring the first 5000 seconds, before significant hydrogen generation, was dorninated by steam flowing from the core -through - the upper plenum, separators, and dome to the safety / relief valves. After the water icvel dropped below the core plate, a natural circulation loop was established between the core channel and the core bypass volumes with the flow go'.1g up the channels and down the bypass. This loop is best shown by - the hydrogen flows for the entrance and exit to the channel and-bypass volumes.
High-flow rates of short duration occurred-during the rapid boiling of the lower plenum water with the flows moving from the lower plenum upwards thro gh the core, upper plenum, separators, and dome to the safety / relief valves. There was also a strong flow through the' separator drains to the downcomer and- then from the downcomer to the ' dome. -These flows lasted i until the lower plenum completely dried out, , The failure of the reactor vessel at a lower head penetration was predicted 132 seconds af ter the lower plenum dried out (1561 seconds after core plate fallure). Chapter 6 contains a discussion of the uncertainty. l associated with the modeling of the penetration failure. The flows during , the vessel depressurization were downwards in the vessel and out' through - the breach in the lower head. With the jet pumps now fully ; exposed a 4 l significant flow through the jet pumps was also established and the flows, l after complete- depressurization, were dominated by a natural- circulation loop between the downcomer.and the core. The primary system fluid temperatures are shown in Figure 5.1 5 which shows the atmospheric temperatures for the core channels and bypass, upper 5-5 i
. -a- - . . . . - . , , . ., -n . .i,-..- :.. -, ,. . .. ., , ,, _ - z 4:, - . .-,:
i plenum / separators, and doree volumes and the channel volume saturation
- temperature. The core channel temperature, as expected, was the highest in vessel vapor temperature calculated. This temperature shows tr.e effect of convective cooling to the core when large steam flows were generated.
! For example, shortly after the channel temperature reached 1534 K at 5500 seconds, the first core debris was deposited on the core plate where it j quickly vaporized most of the remaining water above the plate. j Vaporization continued as water flowed frorn the lower plenum and provided , convective cooling ta the core. A second dip in this teroperature at 8500 l seconds followed the opening of a safety / relief valve causing the lower plenum to flash and provide cooling to the core. The maximum channel , temperature of 1681 K at 13400 seconds occurred just before the failure of the core plate which caused more convoctive cooling. This channel temperature represents the core wide average temperature as calculated by the hydrodynamics roodels, whereas, the core models calculated and used localized fluid tersperaturc s in the channels (see _ Section 5.1.1.3). Further, the plot frequency of this figure was once por l every 100 seconds so the exact maximum temperature and the fine details of the calculation were not captured. The upper plenum vapor temperature was generally cooler than the core . j channel ternperature due to heat transfer to the colder surrounding l structures. The upper dome was even cooler. The downcomer vapor temperature (not shown) was sirnilar to the upper plenum temperature but generally about 100 K cooler. The lower plenum vapor temperature remained , much cooler until after the pool was dried out due to its position below the core and to the pool. W The liquid temperatures within the primary system followed the saturation tempetatures with only reir or subcooling present in the water between openings of the safety / relief valves, with the exception of the water trapped in the recirculation loop. This water was about 20 K subcooled at the time of vessel failure due to heat losses through the piping to the containment. The in vessel vapor temperatures, after the core debris was ejected from the vessel, generally became more stable and uniform. The upper plenum, downcomer, channel, and bypass volumes slowly heated to about 1200 K at the end of the calculation. The lower plenum finished at auout 1100 K and the vessel dome .1t about 780 K. Selected reactor vessel and internal structure surface temperatures are shown in Figure 5.1-6. The maximum structure temperature calculated was
~
1407 K or f the upper shroud head, at the time of first core plate failure. iii. ettr eure, located direct 1v above the core, received energy frors the hot gase flowing from the core to the dome and frorn heat radiating upwards from the core. As this was a structure average temperature, the upper shroud head center m1 6 ht have been predicted to - melt if more detailed modelirig had been employed. After core plate failure, the upper P 56
i shroud head was convectively cooled by the flow of steam from the lower i plenum to the dome and its temperature at the end of the calculation was . 1254 K. The relocation of shroud head steel was not modeled but since the ! pred'.cted temperature exceeded the temperature where steel losses its strength and was approaching the snelting temperature of steel (1700 K), the possibility of relocation has been shown. Portions of the reactor vessel heated to high temperatures as well. The inner and outer temperatures for the cylindrical portion of the vessel vall were 1202 and 1161 K, respectively, at the end of the calculation. The reactor vessel lower head, lower plentun shroud, core shroud, and upper shroud reached similar temperatures to these whil e the upper head and dryers remained considerably cooler. These temperatures continued to increase after the majority of the core was ejected from tt.o vessel (final-ejection at 9.48 hours) due to decay heat from the remainin5_ fuel (4% of core) and from fission products deposited on the structures. These ; structure temperatures would have continued to increase until the heat + transported (primarily conduction through the vessel walls which was domNted by conduction through the vessel insulation) equaled the decay i he g w e. l These temperatures suggest the possibility of sltur. ping or relocation of the reactor vessel internals or even the reactor vessel itself. The major uncertainties associated with these temperatures are the in vessel fission products release models and the deposition rnodels, which determine the decay heat po9er within the vessel late in the accident, and the simple vessel insulation model (Section 3.1.1). 5.1.1.3 Core Meltdown The physical destruction of the reactor core began with the failure of the fuel rod claduing to contain trapped fission product gases located in the fuel rod gap. The cladding failure was . predicted by MELCOR when the > cladding temperature of an axial cell (for ecch radial ring) first exceeded 1113 K as shown in Figure 5.1 7. The first failure for each ring occurred in the top core cell containing fuel. When the cladding melted or formed debris, the temperatures were set to zero by MELCOR to indicate that intact cladding no longer existed in that cell.- The-cladding temperatures were increasing rapidly' at the time of failure (.46, .39,
.33, .10 K/sec) implying that the time of failure was not very sensitive ,
to the specified failure temperature. The releases of the fission product ! gases, which were dependent upon the ring volume fractions and the radial power oistribution, are discussed in Section 5.1.6. The first materia 1' relocation occurred at 1.48 hours when the control rod steel in cell-112 melted. The first cladding to melt (103 seconds later) was also in cell 112. As the cladding in each celi reached the zircaloy melting temperature, the cladding and the fuel formed either liquid metal which candled downward .or debris which filled the available space. The l 'l l i 5-7 l l
-- . . - . . - - _ - - -- - . . ~ -- - - . --
, inner ring claddin6 temperatures (cells 106 through 113) are shown in Figure 5.1 8. (An artifact of the infrequent plot frequency was that not all of the temperatures in the figure appeared to reach the stelt temperature). The order in which the axial cells failed var from top to bottom; the top fueled cell first, then the next one below, and on down to the bottom of the core. The upper non fueled cell reached a maximum cladding temperature of 1357 K before the cell below completely vacated allowing it to move downwards while still intact. The upper core cells failed first, despite their lower power factors, because of heating from the hot gases flowing upward and because they were uncovered first. The cell immediately above the core plate teached th- meltin6 temperature at about the time the core plate failed. The other three core rings behaved similarly except for the effects of convective cooling caused by vaporization of the lower plenus water by the inner ring debris. A core wide time dependent view of the core meltdown is shown in Table 5.1 2. This table shows component temperatures for each core cell at nine different times corresponding to key events. A termpe rature of zero indicates that the component was missing.and non+ entry indicates that the component did not originally exist at that position. Axial levels 6 and 13 do not contain fuel . The core plate was axial location 5 and intact fuel, cladding, and canister components did not exist in axial cells 1 through 5. At 1.26 hours, first cladding failure occurred when tse - temperature- of core cell 112 exceeded the failure temperature of 1173 K and all Considerable melting ant; debris formation components were still intact. had occurred in ring 1 at 1.79 hours with core blockage and debris bed formation on the core plate. At that same time,.the steel control roda in ring 2 had just begun to melt and rings 3 and 4 were still intact. At 3.33 hours, debris bed formation on the core plate had extended into the third and fourth rings leaving only ring 4 intact. The inner ring core plate failed at 3.88 hours, dropping the accumulated deoris into the lower plenum. Partial quenching of this debris generated steam which convectively cooled the remaining core as seen in the next time edit. The core intact components and debris continued heating after the lower plenum dried out resulting in lower head penetration failure at 3.92 hours and in debris ejection from the lower head at 4.08 hours. -When the calculation was terminated, the inner three rings contained' only structural steel in the lower plenum and the outer fourth ring conceined partial fuel (cell 408 and 411), cladding, canisters, and control ' rode sbove the core plate and structural steel below the core plate. The core plate steel in the first three rings and some of the control rod guide tubes in the lower plenum also melted and left the vessel. The ternperature history of. a typical inner core cell in shown in Figure 5.1 9, which shows the temperature. for each cell component for cell 109 along with the local fluid temperature and the material melting temperatures. The first debris appeared in the cell'at 5500 seconds from 5-8
i l debris formed above the cell. The control rod reached its melting temperature of at 6200 seconds and the zircaloy cladding melted at 6400 j seconds. The oxide shell of the canister kept the canister intact after its aircaloy melted until the temperature reached the zircaloy .- oxide melting temperature at 3.60 hours. The molten debris at this time was prevented from downward movement because the cells below were full. Both l the canister and the debris relocated downward with the occurrence of core plate failure at 3.88 hours leaving cell 109 completely empty. This meltdown process is further illustrated in Figure 5.1 10 which shows the masses in cell 109 compared with the cinister temperature.- The masses first increased with the arrival of melt or debris from above and then decreased when the melt or debris moved to the cells below. Metal. oxidation resulted in the continuous increase of zirconium and steel oxide masses. l The temperature gradient of the debris resting on the inner enre- plate before plate failure is shown in Figure 5.1 11. These temperatures help L lllustrate the heat transfer processes which heat and fail the core plate, releasing - the debris to the lower - plonum- . ' The - debris. and - steel temperatutes are - shown for the first three axial cells above the core plate, 'and for- the core plate and the control rod guide tubes below the , core plate. The debris temperatures were higher in the upper cells with ; cell 108 hot enough to melt the fuel. The debris temperature for cell 106 was, in general, higher than for the steel because _ the s teel _- was - conducting heat to the core plate, whereas, the conduction from the debris to the' core plate was not modeled, lleat transfer to the core plate was by radiation from the debris,-by conduction through-the intact steel, and by convection from the hot gases. The steel in cells 107 and-108 melted, the steel from cell 108 flowed downward into cell 107, and:the molten steel in cell 107 was trapped until core plate failure at-3 88 hours. The inner . ring core plate situation is further illustrated with Figure 5.1-12 showing the component masses for cell 106 above the plate. Since : this cell initially started without a fuel mass, the fuel mass shown is accumulated debris. The steel increased to approximately double its initial mass until all the steel in this cell moved downwards at 5.17 hours, lorg after core plate failure. The core plate itself also tuoved downwards at this time following the melting and movement of the lower control rod guide tubes. The core plate temperatures and - failure times are compared in Figure , 5.1 13_ for each radial rin6 The convective cooling that followed the failure of ring 1 delayed the failures of the other rings. , The- lower plenum heat transfer and- lower head penetration failure for the - inner ring are illustrated i_n Figure 5.1 14 which. shows the temperatures of-the debris, the lower head,7 the lower head. penetration, and the-_ vapor and in Figure 5.1 15 which ~ shows - the component: masses for cell 101. The
- core debris . entered the lower plenum at the temperatures of 2684,'3113, L
59
, . - , , ~ ..2- - . - --=% --r' e w r me +er' 2%--,w ryrs r et14 w v e- w'
and 3113 K (fuel melting temperature) for cells 101, 102, and 103, respectively, and completely filled the free volume of the lower two axial cells and part of the third. The debris then partially quenched until the lower plenum boiled dry. Following dry out, the penetration temperature quickly exceeded the assumed failure temperature at 3.92 hours, but the first debris ejection did not occur until 567 seconds later when the debris ejection criceria were met. The first ejected mass consisted of molten steel as the contrcl rod steel in cell 101 reelted. The space occupied by this steel was refilled by debris from above. The next debris ejected frote the inner ring vss at 4.94 hours when the zircaloy contained in ec11 101 melted and it, along with the zirconium oxlde, was ejected. The space was refilled with fuel from above. The ; final ejection occurred at 5.17 hours and included all the remaining mass l in cell 101 and all of the fuel, zirconiurn, and zirconium oxide masses and part of the steel irom cella 102 and 103. Following the last ejection, intact control rod drive steel moved down.to fill the space freed by the ejection. .j At 4.83 hours, the inner surface of the lower head had begun to inolt,- the-outer surface temperature was at 1070 K and more than half of the lower head thickness was above the ' structural failure temperature of 1273 K. By-about 6.7 hours, the entire center three rings were well above 1273 K and structural strength should be considered lost. The relocation of the lower head was not modeled. l The temperatures and the predicted failure times for each of the four lower head penetratiens are shown in Figure 5.1 16. Once core debris was deposited on the luwer head, all the - penetrations rapidly exceeded the i i failure temperature. The-overall process and timing of melt ejection is illustrated in Figure 5.1 17 which shows the lower plenum fuel mass for each of the lower plenum-cells in all four rings. The largest debris. masses were from ring-1 because this ring contained 60% of core mass. Although the lower pier.um contained debris from the inner three rings at the same time, the order of ejection was still from the center out, except, fur instance, that some molten steel was ejected from -ring 2 before the fuel - from ring 1. .The-core plate failure and debris ejection of ring'4 happened much later than the inner three rings and did not overlap the other three. The metal oxidation' process is illustrated in Figure - 5.1 18 which 'shows the total core masses. of zirconium, . zirconium ' oxide, steel, and steel' oxide and . the mass of hydrogen produced in the. primary system. The initial masses of zirconium and steel- in the - core were reduced by oxidation with steam creating the oxide masses. All < the masses - were_ reduced due to melt / debris ejections. A hydrogen mass of 882.2 kg. was produced in vessel by the time of reactor vessel failure and 1137.6 kg by the end of the- calculation. The - percentage of hydrogen, - produced in.. vessel after vessel failure occurred, was 22.5%. 5 10 l
_ _ _ _ . . _ - _ . . _ . . _ _ _ _ _ _ . _ _ _ - - - ~ _ _ _ _ . . . - - The local channel fluid temperatures at the top of the core for each radial ring are compared to the channel control volume temperature in Figure 5.1 19. The local fluid temperatures were used by the core models to calculate the convective heat transfer from the core components but the control volume temperature was used by the hydrodynamics models to calculate heat transport to the other control volumes. The local core exit fluid tstsperatures were much higher than the control volume ; temperature. For instance, the average core exit temperature at 3.33 ; hours (12000 seconds) was 2455 K compared to 1274 K for the control volume and the average local temperature over the entire core was 2431 K. An inconsistency exists between the core and the hydrodynamic models in that the heat transport from the core calculated by the hydrodynamics models , did not agree with the heat transport indicated by the core models (energy ; was conserved) . This inconsistency affects upper plenum heat transfer, , and the_ transport of fission products, therefore, it is another source of ! uncertainty. ; s 5.1.1.4 Containment flydrodynamics I j The containment contained the released fission products until it failed i due to over-pressurization caused by the addition of steam, non* condensable gases, and heat. Steam van produced by the vaporization of ' water and from the decomposition of concrete. Non.condensible gases were produced from the oxidation of structural metals, from the decomposition of concrete and from the chemistry of the debris / concrete interactions. Ileat was produced both by the radioactive decay of_ tission products and
.from exothermic chemical reactions. lleat transfer into the steel and '
concrete structures and steam condensation both tended to reduce the containment nressure. The containment pressures are shown in Figure 5.1 20 for both the wetwell and the drywell. Initially, the pressure increased primarily due to the steam and hydrogen gases released from the primary system through the safety / relief valves into the suppression pool (the steam was mostly ' condensed). The rapid pressure increases, following lower head failure, were .236 MPa (34.2 psi) and .195 MPa .(28.3-~ psi) for the drywell and wetwell volumes, respectively, peaking about 15 'minutet af ter vessel failure. The peak differential pressure across the drywell floor was .049 MPa (7.1 psi). The containment pressurization continued af ter primary system depressurization due to the boiling of water in the reactor pedestal volumes and the ' production of steam and non condensable gases l from the core / concrete interaction' and concrete degassing. Containment failure-was predicted at 18.2 hours with a leak through_the wetwell vall. Fission products--were transported from the containment into ; the = reactor i building, following failure, first by the depressurization flows and t then by the flow of gases produced continuously by oxidation and concrete j decomposition. The temperatures of the drywell atmosphere and surfaces are shown in Figure 5,1-21. .The' temperature of the miscellaneous steel within the 5 11
. _ ,-., _ _ ._, _ _ _ a,- - ~ , _ _ _ . , .. ..._--,__--a ~
- - ~ _ - - - _ _ - -. . . _ _ . - - . - _- ~ _ - __._ __ - - - .
drywell closely followed the atmosphere teitperature except during the highly transient period immediately after vessel failure; but the concrete surface temperatures lagged behind the atmosphere temperature as heat was transferred to the interior of the concrete walls. The average temperature for the dryvell containment wall increased by 17.2 K. The temperatures of th; wetwell a t tnosphe re and surfaces are shown in Figure 5.1 22. The t erope ra ture peaked af ter the concrete ablation rate and debris teroperature within the wetwell pedestal peaked. The average temperature for the wetwell contairwet wall increased by 18.2 K. The hottest containment temperatures were in the wetwell pedestal volume where the concrete was ablated by molten core debris. The atinosphere and surface temperatures for both of the pedestal voitunes are shown in Figure 5.1 23. The peak atmosphere and surface teroperatures (1815 and 1765 K, respectively) both exceeded the melting temperature of steel verifying the initial input aestunption that the steel would melt. The first core debris j arrived in the wetwell padestal at 4.41 hours and the overlaying pool of water boiled away in anot.her 2.22 hours. (The fluctuations seen in this figure were dite to additions of water frate the drywell pedestal.) l The surface teroperatures for concrete above the debris exceeded the input temperature for concrete ablation suggesting this concrete could ablate away and fall into the molten pool. Ablation of concrete above the debris is not modeled by MELCOR. The impact that this ablated concrete would have on the rate of non condensible gas generation, the molten pool level (possible overflow into ti.e suppression pool), or pedestal wall failure is another calculational uncertainty. Concrete dehydration, which is roodeled, released large quantities of-ateam and carbon dioxide gases from the concrete above the molt. The drywell pedestal, shielded from the noiten debris .by the concrete 11oor, remained much cooler than the wetwell pedestal but about 200 K hotter than - the drywell. Steam was released from the drywell pedestal concrete but at smaller quantities than in the wetwell pedestal. Stearn condensation within the contairunent was an important factor in the containment pressurization and is illustrated in Figure 5.1 24, The condrasation mass fluxes for two drywell surfaces and one vetwell surface are shown. The most rapid condensation occurred 1n the drywell just after reactor vessel failure and in the wetwell when the water over laying the rnolten debris boiled away. The condensate in the wetwell entered into the suppression pool and the condensate in the drywell fell to the floor, then drained into the pedestal- where it was revaporized. The total dryvell condensate is also shown in the figure, as measured by the - flow- through the drywell drains. The water rnasses for pools accumulating in the drywell, drywell pedestal, and vetwell pedestal are shown in Figure 5.1 25. The initial water accumulation was leakage from the pump seals and control rod drives. The l 5 12 l' 1
drywell pedestal drained into the vetwell pedestal after the failure of the drain pipes. Later, the water flow between the drywell and vetwell pedestals fluctuated due to countercurrent flow through the drain pipe holes. The suppression pool mass increase and temperature are shown in Figure 5.1 26. The pool mass increased due to condensed steam flow from the safety / relief valves until the primary system failed, then continued to increase due to condensation on the heat structures and the pool surface. The largest step increase, 68770 kg, occurred following the first core plate failure when molten debris fell into the reector vessel lower plenum and vaporized the remaining water. After primary system depressurization, about 81% of the initial primary system water mass resided in the suppression pool. The suppression pool continued to heat due to the addition of warmer condensate and decay heat from fission products deposited within the pool. A linear extrapolation of the pool temperature predicts that the pool will boil about 2 days after the accident started if the containment is fully depressurized by that time. The downcomer and wetwell suppression pool water levels are shown in Figure 5.1 27 and the clearing of the vent downcomers following the reactor vessel failure is illustrated. The initial downcomer level deprecolon was a result of drywell heating following the loss of the containment heat removal systems at the _ initiation of the accident. *;'he increase in the suppression pool level associated with lower plenum dry ouc and reactor ver.sel depressurization was .2 meters and the total increase at the end of the calculation wac .9 meters. The final level was still .9 meters below the bottom of the ports into the wetvell pedestal volume containing the molten debris. The first mcl. ten core debris poured from the reactor vessel lower head into the drywell pedestal at 4.08 hours (14670 seconds) and consisted of liquid steel along with some steel oxide. This first pour and a smaller one, shortly af ter the first, were cooled by- the -pool of water and refroze. Twenty minutes later, the drywell pedestal floor was modeled to failed and the debris completely passed through the failed pipe openings to the wetwell pedestal. The masses and temperatures of this steel and stool oxide are shown in Figure 5.1 28. The input model for the failure of the cavity floor drain pipes was based on expert opinion (ses model description in Section 3.1.2) instead of the actual physical processes due to the lack of a phenomenological model. An initial pour of liquid steel from the vessel was not anticipated and the expert opinions were probably based on a much larger pour which included fuel, as is characteristic - of a STCp calculation. Other scenarios art possible; for instance, the drain pipes might not have failed until a much larger fraction.of the core poured from the vessel or.the pipes could have failed allowing the water to drain but leaving the debris in the drywell pedestal until it became molten again. This - issue is another source of uncertainty. 5 13
. . - - _ _ ~ _ -. . -- . _ _ _ _ - - - - . .. _. -
Additional molten debris from the vessel, after the drywell pedestal floor failed, was modeled to pass directly through the drywell pedestal to the wetwell pedestil. The concrete ablation in the drywell pedestal was insignificant compared to the wetwell pedestal. The wetwell pedestal debris masses are shown in Figure 5,1 29. The first debris arrived af ter the drywell cavity drain pipes failed (the wetwell pedestal was initialized with 15450 kg of steel, see Section 3.3), followed by additional pours from the reactor vessel seen as step increases in the total debris mass. The first fuel arrived 46 minutes after the pipes failed. The gradual increases in the oxide masses were from the addition of ablated concrete and, as lighter oxides from the concrete accumulated in the heavy oxide layer, its density steadily decreased until it became lighter than the over laying layer of liquid metals, upon which, a computational layer flip occurred (referred to the core / concrete interaction models). When the layers flipped, the metal layer was ; transferred to the bottom of the cavity and the heavy and light oxide j Jayers were joined together into a single upper layer. The metal layer i consisted mostly of steel and zirconium, however, the zirconium rapidly oxidized where it was computationally transferred to the light oxide layer. The metal layer is seen to increase due - to additional pours but then decrease due to the oxidation of zirconium until only steel remained. After which, steel was slowly added to the layer from the concrete rebar and removed by oxidation. The wetwell pedestal molten debris temperatures are shown in Figuxe 5.1 30. The temperatures peaked during the period of rapid exothermic zirconium oxidation. The peak temperature of 2193 K was in the heavy oxide layer upon the arrival of the first fuel. The debris temperatures at the end of the calculation were slowly decreasing but still above the concrete ablation temperature. The top and bottom elevations of the molten pool are compared to the elevations of the structure and the suppression pool aurface in Figure 5.1 31. The calculation predicted that the surface of the molten pool came within .15 meters of the ports connecting the wetwell pedestal to the wetwell during the period of rapid zirconium oxidation. Therefore, the possibility that molten debris could flow through the ports into the suppression pool is an uncertainty. During this period, the peak void fractions within the molten pool were 13.1, 9.5, and 9.1% for the heavy and light oxides and metals, respectively. The' level of the water pool overlaying the molten pool is also shown. The maximum redial and axial molten pool penetration- distances into the concrete structure are shown in Figur_e 5.1 32, _ and a two dimensional axisymmetrical cavity shape drawing is shown in Figure 5.1-33. The concrete ablation rate was the most rapid during the rapid zirconium oxidation period. The end of calculation penetration distances were 1,95 5-14
meters radially and 1.77 meters axially. The radial penetration distance exceeded the pedestal vall thickness of 1.47 meters at 19.2 hours (about i hour after containment failure) suggesting the possibility of structural failure of the wall which supports the reactor vessel and shield, and the possibility of interaction between the molten debris and the suppression pool. When the pedestal wall thickness was exceeded, the calculation continued as if the wall was semi-infinite, providing results for one potential outcome of the wall failure issue. Carbon dioxide, <.erbon monoxide, hydrogen, and steam gases were produced as by products of the concrete ablation and the resulting chemistry of the molten pool, and were major contributors to the pressurization of the containment. The integrated masses of these gases are shown in F1 6ure 5.1 34. The carbon monoxide production dominated _ during the rapid zirconium oxidation period but was later exceeded by the carbon dioxide production. The end of calculation totals were 57110, 25610,-12610, and 1031 kg for carbon dioxide, carbon monoxide, steam, and hydrogen, I respectively. The hydrogen gas produced by the molten pool (ex vessel) was an-important l contributor to the later reactor building combustion and fission product transport as was - the hydrogen produced by the core meltdown (in vessel) . The in vessel and ex-vessel hydrogen production are compared in Figure 5.1 35. A total of 2168 kg of hydrogen was produced (1138 kg in-vessel and 1031 ex vessel) . The majority (72%) of the ex-vessel hydrogen was produced during the period of zirconium oxidation within the molten poo'.. In addition to the gases produced by the molten pool interacting directly with concreto, steam and carbon dioxide gases were released within the containment from heated concrete. The integrated masses cf these gases are shown in Figure 5.1-36 along with the total release rate. A total of 36750 kg of steam and 16290 kg of carbon dioxide were released by this mechanism with the majority coming from the _ pedestal wall and drywell cavity floor located immediately above the molten pool. Further, the steam release distribution by structure was 60,0, 26.8, 6.9, and 6.3% for the pedestal wall, reactor shield wall, drywell floor, and drywell cavity floor, respectively. 84.2% of the-carbon dioxide came from the pedestal wall and 15.7% from the drywell cavity floor. The ratio of moles of gases produced by degassing to moles produced by the core / concrete interactions-provide a method of estimating the relative importance of modeling concrete degassing relative te - core / concrete interactions. The ratios were .136 for non-condensable gases, 2. 915 - for steam, and .704 for all gases. The mass and mole fraction distribution of hydrogen within the containment is _ shown in Figure 5.1-37. Prior to the _ reactor - vessel -- failure , the , hydro 5en was concentrated in the wetwell because the hydrogen entered the l containment through the safety / relief ~ valves to the suppression.. pcol. , After vessel- failure, ex vessel hydrogen production and . drywell l condensation concentrated more hydrogen in the drywell until just before ' E 5 ! l l . . . . - - - , - _ - . .
containment failure. At containmont failure, the wetwell and dryvell hydtcgen masses were 1011 kg and 854 kg and the mole fractions w;,re .282 and .189, respectively. The pedestal and downcomer volumes contained much smaller quantities in accordance with their smaller volumes. Containment flows are illustrated in Figures 5.138 through 5.141. The drywell pressure relief through the vent /downcomers during primary system depressurization is shown in Figure 5.1-38. The plot frequency during this portion of the calculation was not frequent enough to capture the details of the vent clearing but it took between 5 and 35 seconds after the reactor vessel failed to clear the downcomers of water. The highest flow rate recorded was 111 kg/see but the peak rate was probably missed due to the infrequent plotting. The downcomers refilled again preventing further gas flow. The flow rate through the vacuum breakers between the wetwell and drywell is shown in Figure 5.1 39. An input error for the vacuum breaker model was located as a result of the negative flow shown, because the breakers are intended to operate as check valves. The error eliminated the small hydrostatic head term of the wetwell atmosphere in the calculation of the pressure differemial ef fectively allowing the - vacuum breakers to leak-from the drywell to the wetwell. The. leak did not occur prior to reactor vessel failure during the operation of the safety / relief valves. The negative flow rate included a small- amount of drywell floor _ condensate which dominated over the gas leak flow. This input error was deemed to have little effect on the fission product transport. Two other flow paths existed between the wetwell and the drywell besides the vont/downcomers and the vacuum breakers. These were the two failed four-inch pipes between the wetwell pedestal and the drywell pedestal and the half-inch nitrogen system bypass line. The bypass line flows were small and had little effect on the course of the calculation. The-flows through the failed four-inch pipes are shown in Figure 5.1-40. These flows were unstable after the pipes failed due-to the irregular condensate flow from the drywell. The flow rates were generally positive during containment pressurization and negative during containmenc depressurization. A composite of the contairment flows is illustrated by the integrated hydrogen flows-shown in Figure 5.1 41. A strong natural circulation flow existed through most of the calculation between the drywell _ and the drywell pedestal which had a significant effect on the drywell pedestal temperatures. The circulation path was from the drywell through the - drywell drains to the pedestal and then through the pedestal wall-hatches and openings back to the drywell.- The integrated hydrogen vacuum breaker-flow - was predominantly positive to about 10 - hours where a - weakened negative flow began which became n.uch more negative after containment failure. The hot core / concrete interaction flows from the wetwell - pedestal went predominantly to the watwell where a portion reached the drywell by means of the vacuum breakers.
'5 16
5.1.1.5 Reactor Building Hydrodynamics The hydrodynamic characteristics of the teactor building after containment failure determine the transport of fission products entering the reactor building, by means of the containment break, to the environment. The hydrodynamic characteristics include the containment depressurization, hydrogen and carbon monoxide combustion, and natural circulation flows. The containment failed with a leak in the wetwell wall at 18.2 hours. The containment depressurization total and hydrogen flow rates are shown in Figure 5.1 42. The hydrogen mole fraction of the depressurization flow is also shown. These flow rates were nearly independent of the reactor building hydrodyntmics, however, the two short dips in the total flow rate were the result-of reactor building combustion. The dynamics of hydrogen and carbon monoxide combustion retalted in spikes in the pressures, temperatures, and flow rates which were important to fission product transport as will be shown ir the Section 5.1.1.5. The reactor building atmospheric temperatures are shown in Figure 5.1 43 -and the peak pressures and temperatures are listed in Table 5.1-3. The second of the two burns was the most severe with- a peak pressure increase (over atmospheric pressure) of .136 MPa (19.7 Psi) in Unit 1. The pressure-spikes were similar for tha Unit 1, the Unit 2, and'the refueling floor volumes but the steam tunnel and turbine volumes were relatively isolaced and the - burn effects much smaller. Both burns -originated in the upper section of Unit 2, however, the second burn propagated into Unit 1. The peak temperature spike of 1308 K (1894 *F) was calculated in Unit 2 for , the second burn. The combustion dynamics are further illustrated by Figure 5.1-44 which shows the mole fractions for the gases in upper Unit 2. The hydrogen and carbon monoxide mole fractions ' increased prior - to each' burn and then sharply went to zero as they were consumed during the burn.- These gases were then replenished by the containment' depressurization flows. . The oxygen mole fraction was reduced during each burn - and increased between burns - as air was circulated back into the reactor building. Ti.3 corresponding sharp increase in ' the- water vapor- . mole fraction is also shown. The hydrogen mole fractions are compared in Figure 5.1 45. The first burn was limited to the upper Unit 2 volume where the containment leak occurred but the second burn propagated into Unit 1 and 'the refueling floor as-well. The lower Unit 2 and the two steam tunnel and turbine volumes did not - accumulate sufficient hydrogen to burn. Had the calculation been continued, a third burn would have been likely. The reactor and turbine building flows .are illustrated -in the Figures 5.1 46 and 5.1-47. Figure 5.1 46 shows the flow rates at the end of the calculation at each of . the flow . paths located on a schematic. of the building. These flow rates include both the late containment 5-17'
depressurization flows and natural circulation flows. Figure 5.1 47 shows the time integrated flows at the end of the calculation which includes all the flow effects including combustion driven flows. The containment depressurization flows entered the upper Unit 2 and flowed predominantly upwards to the refueling floor and then to the outside. All flow paths modeled as blow out panels to simulate doors and hatch covers were opened completely by the first burn and remained open. The flow paths from Units 1 and 2 to their respective steam tunnels were modeled as flap covers which reclosed again with the loss of pressure differential. These flaps only opened during the two burns and limited the transport of fission products into the steam tunnels. A strong natural circulation flow loop was established among the upper Unit 2, the refueling floor, and the Unit 1 volumes with flows leaving the top of Unit 2 entering the refueling floor, flows downwards from the refueling floor to Unit 1, and flows from Unit 1 to Unit 2. Smaller, but important, natural circulation flows involved flows from the outside entering both Units 1 and 2 through building leakage near the ground level and leaving the building at the refueling floor. These flows fron, the outside restored the oxygen concentrations within the building after - ombustion depleted the oxygen. 5.1.1.6 Radionuclide Transport A large amount of detailed radionuclide transport data is available in the MELCOR results. A aummary . of these results is included along with examples of time dependent details. These results are illustrated in s Tables 5.1-4 through 5.1-25 and Figures 5.1-48 through 5.1 61 and are referenced in the following discussions of particular res its. The default classes of radionuclides are listed in Table 2.7 1 and will be referred to by the class representative, for instance, the class containing the noble gases will be referred to simply as xenon. Non-default class 16 was created for cesium iodide. The predicted radionuclides released and transported by MELCOR-include the radioactive fission products released from the fuel as a vapor or as an aerosol, aerosolized radioactive uranium fuel, and non-radioactive acrosolized structural materials. These materials were released both in-vessel during the core meltdown.and ex vessel during the core / concrete interaction. The first in vessel releases occurred with the failure of the inner ring cladding gap at 1.26 hours and the first ex vessel releases occurred after the first debris ejection at 4.08 hurs. The non-radioactive structural materials include steel, zircaloy, and concrete. An end of calculation summary of the predicted radionuclide releases by class is found in Table 5.1-4. The in-vessel and ex-vessel release masses are given in Tables 5.1-5 through 5.1-17.(one tabla for each class) at the time of reactor vessel lower head ' f ailure . after reactor vessel depressurization,.at the. time of containment failure, and at end of the ; calculation. 5-18 q l
For some classes such as xenon, tuost of the fission products inventories were released from the fuel, while others such as cerium, released but a small fraction of their initial inventory due to their relatively low volatility. Some classes showed high releases of non radioactive materials. These were ruthenium which includes nickel from the steel, inolybdenum which include iron and chrotnium from tl.e steel, and cerium which includes zirconium frorn the cladding and canisters. An uncertainty exists for the in vessel release rates for cesium, molybdenum, and uranium. The leading coefficient in the CORSOR release correlation was reduced (refer to Section 3.6) to counter extrernely high predicted release rates causing unrealistically dense in core aerosol densities. These high densities occurred. when core temperatures < approached or possibly exceeded the upper teroperature lirait for the correlation, and contributed to other numerical problems (refer to Chapter
- 6) resulting in radionuclide rnasse s being numerically lost from the calculation. The reduced release rate alleviated the mass loss problem to a rate that could be accepted. The impact of reducing the r.ste for cesium was to spread its release over a longer period, since the majority of the cesium was released in vessel anyway. The total releases of molybdenum (which contain the aerosols formed from structural steel) and of uranium likely were underpredicted; their leading coefficients were reduced by 1000.
The radionuclide releases within the reactor vessel were in general much larger than the ex vessel releases; the exceptions were molybdenurn and uranium and these could be due to the reduction in their release coefficients. There are several explanations for the greater in vessel releases: (1) the majority of the. initial inventory was released in vessel for some classes leaving little to be released from the ex vessel- fuel, (2) much of the in-vessel fuel reached considerably higher temperatures than it did when it was ex-vessel, and (3) the in vessel and ex vessel release models were completely cifferent. Cesitun iodide was formed from cesiuu and iodine upon release from the fuel (refer to Section 3.6). Since the initial inventory of cesium had 12.2 times as many inoles as did iodine, there was usually but not always, more cesium released in each time step than iodine. This resulted in 8.1% of the released cesium forming the halide which consumed nearly all of the iodine. Only a small amount of the iodine remained the elemental form. Three classes are modeled by MELCOR as forming molecular compounds upon release. Cosium is modeled to forin a hydroxide (Cs0H); telluriurn an oxide (Te0); and uranium an oxide (UO2 ). The ratios of total- to radioactive snasses for these classes are siinply the ratios of their elemental to compound inolecular weights. Two examples of the timo dependent releases are shown in Figures 5.148
, and 5.1 49. Both the in-vessel and the ex vecsel releases of both radioactive and non-radioactive materials represented by cerium are shown k.
5-19
in Figure 5.1 48. The in vessel releases of cerium began during the tiene of core relocation and ended when the core plate failed resulting in the core being convectively cooled by the steam flow from the lower plenum. The ex vessel releases of cesium began when the core debris started to significantly ablate the concrete and continued throughout the calculation with the largest releases occurrie- durin6 the rapid oxidation of the zirconium. Figure 5.1-49 shows the ex vessel releases of non radioactive structural acrosols. The release rate of steel and concrete aerosols vas related to the rate of concrete ablation. While these non radioactive 1:terials were not important from a radiological consequence standpoint, they are involved in the aerosol dynamics and did contribute to the transport of , the fission products. Once the fission products were released f rom the fuel, the next retention barrier was the primary or reactor cooling system. The retention of the- j pritnary system is shown first by an end of calculation retention factor in l Table 5.1-4. This retention factor is defined as the mass of the I radioactive radionuc1!Jeb (by class) retained within the primary system l divided by the radioactive mass released from the fuel within the priinary - system. Figure 5.1-50 shows the time dependent retention factors - for ' several of the classes (classes not shown behaved similar to ruthenium which was typleal for the aerosols). The radioactive masses contained within the primary system (at four significant times) are listed in Tables 5.1 5 through 5.1 17. . Table 5.1 18 lists data illustrating early versus late transport from the primary system. Detailed distrioution information for fission products within the primary is available in the MELCOR output but is not included here. The lowest primary system retention was for xenon which could not deposit on structures. The final fraction was only .0045. These gases circulated within and transported from the primary system with the steam and hydrogen gases. The aerosol classes in general showed typical aerosol behavior. The highest retention factor of .82 was for molybdenum. After the prinary system depressurization was complete, their primary system retention rernained relatively constant with some exceptions. The fraction of the radionuclides transported from the primary system af ter vessel depressurization is given in Table 5,1-18 along with its relative irnpac t on the releases from the containment. For certain classes, a significant fraction of the transport from the primary system occurred after vessel depressurization which illustrates the importance of doing fully integrated calculations. For example, . 53% of the iodine ~, leaving the primary system, left after the vessel depressurized. The molybdenum, uranium, and tin classes showed a net negative late transport-from the primary systein implyin5 some radionuclides were transported into the prirnary system when the contairanent pressurized which occurred after the vessel had depressurized. 5 20
i The relative impact factor in Table 5.1 18 indicates the effect of the i late fission product releases from the primary systern on the amount of l fission products escaping from the containment. This factor is defined as j the total mass of fission products transported from the primary system for the entire calculation plus those released externally divided by the early transported mass from the primary plus the ex vessel releases. The early and late transport periods correspond to the time periods before and after the completion of reactor vessel depressurization. For example, 26% of the tellurium released in-vessel was released late in the calculation and the corresponding relative irnpac t factor of 1.3 indicates that the containment tellurium class fission products were about 30% greater than if the late releases were not modeled. An advantage of doing fully integrated calculations is being able to continue the in. vessel fission product transport throughout the entire accident. The iodine and cadmium classes vere in the vapor phase at the elevated temperatures of the reactor core. Both classes showed a sharp drop in primary system retention after containment failure. Late in vessel releases of iodine occurred at about 16 hours as indicated by an increase in primary system retention in Figure 5.1 50. The retention and distribution of fission products in the containment, the next retention barrier after the primary systern, is illustrated by Tables , 5.1 5 throuSh 5.1 17. The fission product masses are given for the drywell, the wetwell, the drywell and vetvell pedestals, the suppression pool, and the containrtent total at times corresponding to reactor vessel ; i I failure, after vessel depressurization, containment failure, and the calculation end. Table 5.1 19 lists the fractions of the fission products released from the fuel which were located within the -containment (excluding the primary system) at the end of the calculation. The table , also gives the distribution of these fission _ products within _ the containment. Examples of time dependent containment fission product - l masses are shown for the xenon and cesium classes in Figures 5.1 51,_.and' j 5.1 52, respectively. , t ! The majority of the xenon class gases had escaped the containment at the l end of the calculation, only 19.4% remained. 57.7% of the remaining xenon l was located in the drywell and 42.3% in the wetwell. The suppression pool l model does not allow these gases to be retained in the pool, t l Most of the cesiurn class was released within the primary system (86.7%) I and a large portion of these fission products was deposited in the (_ suppression pool by means of the safety / relief valve flows. 88% of the containment cesium class fission' products were located in the suppression pool. Several other aerosol classes, such as Ruthenium for example, showed similar behavior but the suppression pool-retained smaller portions of fission products for classes with higher late in-vessel releases 'or higher ex vessel releases. The final barrier for retaining the fission products was the . reactor / turbine building. The end' of calculation fission product masses l' 5-21
located within the reactor building, the turbine building, and the environment are given in Tables 5.1 5 through 5.1 17. A series of figures, Figure 5.1 53 through 5.1 58, show the time dependent fission product masses escaping from the containment and the masses entering the environment, for each class. The retention of fission products by the reactor / turbine building is illustrated by using decontamination factors. The decontamination factor is defined as the total fission product- mass (by class) outside of the containment divided by the corresponding mass in the environment. The end of calculation decontamination factors are listed in Table 5.1 4 and tino dependent decontamination factors for selected classes are shown in Figure 5.1 59. The environmental source terms, the time dependent rate at which fission products enter the environment, were influenced 'by a number of factors. The hydrogen and carbon mono; tide burns within the reactor building (refer to Section 5.1.1.5) dramatically enhanced the transport of fission products through the building. The source terms increased by step increments (Figures 5.1-53 through 5.1 58) as a result these burns. Then, after a burn cleared the building of airborne fission products, the source terms significantly decreased again, The source terms were also affected by the depressurseation rate. The- containment escape rate Increased dramatically towards the end of the calculation for some classes (e.g., tin and uranium), due to increased fission product releases from the ex-vessel core debris. A review of the coro/ concrete results did not indicate the cause of the increased release. The transient nature of the envirotunental source terms is further illustrated by the time dependent decontamination factors (Figure 5.1-59). The aerosol site distribution of the reactor building is shown in Figure 5.1-60. The distribution was domin.sted by particles in the 2 to 8 micron size. The smaller sized particles agglomerate to the next larger group and the larger sizes were more .etrongly influenced by gravitational settling. The radioactive decay power distribution is sbwn in Figure 5.1 61. The , total decay power for the entire plant was 199.4 MW at time - zero but initially dropped very quickly, The decay powers associated with the.in-vessel fuel and the ex vessel fuel are also shown. During lower head debris ejection, decay power was transferred from the in vessel _ fuel to the ex vessel fuel as the fuel was transferred. The decay power ~ not associated with the fuel belonged to the fisnion products released' from the. fuel . These fission products were either airborne or deposited in a pool or on a structure, The end of calculation distribution for airborne fission products, versus. fission products deposited in a pool or on a structure is given (by class) for the primary system, the containment, the reactor / turbine building, and-the entire plant in Tables 5.1 20 through 5.1 23. Most of fission products contained in the primary. system at the end were deposited on the reactor vessel and internal structures. Small masses were -deposited - in 4 5 22
the recirculation loop piping water and still smaller masses were airborne. -The decay power from these deposited fission products was responsible for the high in-vessel structures temperatures (section
$.1.1.2). The nenon class gases were, of course, entirely airborne.
In the containment, the xenon, uranium, and tin classes still had relatively high airborne fractions (100, 11.1 and 28.3% , respectively). Most of the fission products were deposited in the suppression pool for some classes (c.g , cesium, tellurium, ruthenium, cerium, lanthanum). The reactor / turbine building distributions tended towards high airborne fractions, and for many classes, there was a higher deposition in the refueling pool than on all the structures. However, a MELCOR coding error was discovered after the calculations . were run, that affected the de,,osition on the structures in the reactor building. Many structures 6 the building were modeled (user input) as multiples of one structure. With the coding error, the structure multiplicity was not-included in the deposition computations. Multiplicity was used extensively in the reactor building but- not inside the containment, therefore, the result of this error is_that the fission product deposition within the reactor building was underestimated, probably by a large amount, but the error hcd little , effect inside the containment. The total fraction of all released fission products deposited on structures for the entire plant is given in Table 5.1-23 and is further broken down into deposited mass fractions . for - the primary system, the containment, and the reactor / turbine building in Table 5.1 24. The largest deposits, .for most classes, were in the primary system, then the containment, and finally the reactor building. The iodine class, which existed in the vapor form within much of the primary system,. deposited most heavily in the containment. Once fission products were deposited on a surface, they remained there unless the surface became hot enough to revaporize them. Fission product resuspension is not currently modeled by MELCOR. The only class of fission products which currently can emerge from a pool is the iodine class unless the pool completely vaporizes. The in vessel water in the core, downcomer, and lower plenum released radionuclides by drying out and the water in the recirculation loop- released small-quantities of iodine as flashing occurred during the primary _ system and containment depressuriz;tions. The containment suppression pool and the reactor building refueling pool remained subcooled ' throut,h out - the calculation, thereby, retaining all fission products deposited in.them. The fission prodoet distribution between vapor and aerosol forms is shown in Table 5.1-25 as the ratio of atmospheric . vapor mass to atmospheric aerosol mass-(by class) for the primary system, the containment, and the reactor / turbine building. Some. classes existed predominantly in the vapor form in the primary system, (e.g., cesium and iodine), but in the 5-23
containment and reactor building formed an ae rosol - because of cooler temperatures. Other classes exfsted predominantly in the aerosol form even within the reactor vessel. a the reactor / turbine building, the only class existing in the vapor form, besides xenon, was a small portion of the iodine. 5.1.2 containment Failure Sensitivity Calculations The twelve sensitivity calculations based on the high pressure short term station blackout calculation described in Chapter 4 are discussed here. These calculations examine the sensitivity of fission product transport through the reactor building to the containment failure and the hydrogen ignition parameters. The parameters varied in these calculations (refer to Table 4.1 1) include the containment failure area and location, and the minimum ignition concentrations. All of these calculations were run.from a restart file written at the time the wetvell was predicted to fail. The break flow resistance was varied to simulate break sizes ranging from a small leak simulating a crack in the concrete wall to large . ruptures. The flow resistance is characterized- by a modified flow area .(the actual flow area divided by the. square root of the sum of the loss coefficients and the friction loss term). The four modified flow areas used in these calculations (7.99E 4, .00924, .0924, and .924 m2) are listed for each case in Table 4.4 1. 4 The larger break area calculations were continued beyond the completion of the containment depressurization to predict the rapid transport of the fission products during the depressurization and the slower transport due to the production of gases from the continuing concrete oblation. The smaller break calculations, the small leak calculations in particular, were terminated before the containment was completely depressurized, due to resource constraints. The calculation termination times (relative to both the beginning of the calculation and the failure of the contairaent) and the percentage of depressurization completion are listed in: Table 5.1 26. The depressurization rates are also illustrated by the wetwell pressures for. Cases 1, 2, 4, and 11 in Figure 5.1 G. Case 11, with the largest break, depressurized the containment in about .' minutes, and the next largest break calculation, Case 1, took obout 20 mt.iutes . Cases 2 and 4 were terminated at 90.1% and 16.4% completion, respectively. Tables 5.1-27 through 5.1 34 present end of calculation values for normalized fission prc iuct masses escaping the containment, noractized releases to the environment, and containment decontamination factors. The masses were normalized to one case in each comparison. 'lhese comparisons show trends in the transport of fission products through the reactor building, however, caution'must be exercised in using the values in these tables becauseithe calculations did not all end at-the same time and they were not all fully depressurized. Selected time dependent resulte e.re presented in Figures 5.1-63 through 5.1-72. 5-24
Most of the entries for the iodine class are suissing in the tables because the quantities of this class escaping the containment were minute. One reason for the small masses is that the majority of the iodine released from the fuel combined immediately with cesium upon release. Most of the remaining iodine was deposited on structures or in pools leaving only trace amounts available for transport to the reactor building. The results of these sensitivity calculations were also affected by the multiplicity error discussed in Section 5.1.1.6 and the fission product deposition within the reactor building was underestimated. 5.1.2.1 Break Flow Resistance Sensitivity Tha sensitivity of the fission product transport to the containment break size is shown by the comparisons of Cases 1, 2, 4, and 11 for a vetvell wall break (Table 5.1 27), Cases 3, 8, and 10 for a drywell vall break (Table 5.1 28), and Cases 5, 7, and 9 for a drywell head break (Table 5.1 29). For all of these cases, the minirnurn hydrogen and oxygen rnole fraction ignition limits were set at St. Time dependent fission product release masses (Cases 1, 2, 4, and 11) are shown in Figures 5.1 63 and 5.1 64 for the xenon and bariurn - classes, respectively. The magnitude of the releases generally increased as the break size increased. The largest break calculation (Case 11, .924 in2) , however, predicted smaller environmental source terms than the next largest break calculation (Case 1, .0924 m2). The smaller source term was partially due to less severe hydrogen burns in Case 11 than in Case 1. The burns tend to sneep the reactor building of airborne fission products resulting in step increases in the environmental source terms. In Case 11, more of the combustible gases (hydrogen, carbon monoxide, and oxygen) passed through the reactor building to the environment without burning. The majority of the xenon class gases released from the fuel also escaped the containment in the fully depressurized cases. Virtually all of these gases would eventually enter the environment. The barium class fission products, which existed predominantly in aerosol form, strongly deposited on surfaces or in the suppresrion pool. Only a small fraction - of the bariuta, released from the fuel, escaped the containment as shown in Figure 5.1-64. 5.1.2.2 Break Location Sensitivity The sensitivity of the fission product transport to the containment break location is shown by a comparison of Cases l', 3, and 9 for_a break = size of
.0924 m2 (Table 5.130), Cases 2, 5, and 10 for a break size of .00924 m2 (Table 5.1 31), and Cases 4, 7, and 8 for a braak - size of .000799 m2 '<able 3,1 32). Selected time _ dependent results for break location-x tsitiQths are shown in Figures 5.1 65 through 5.1-70, For all of se cases, the minimum hydrogen and oxygen mole fraction ignition liraits ce set at 5%.
5 25 l
. a $# ,
__ _.__.______._____.____..._.__-____.____m The results of these sensitivity calculations were both fission product clase and containment break size dependent. The transport of the xenon class gases to the environment was not dependent on the break location for the large break sizes as illustrated in Figure 5.1 65 because most of these gases were released from the containment. But for the small leak calculations (not shown), the drywell wall and drywell head leaks released the xenon faster than the wnwell wall leak because of a higher concentration of gases in the drywell. The transport of cesium and barium is illustrated for a large break in
'.tgures 5.1 66 and and 5.1 67, respectively. The escape of cesium from the containment and release to the environment were higher from the drywel. breaks than for the wetwell break, but for bariu:n, the releases were much higher for a wetwell break. The reason for this apparently opposite behavior was not immediately clear and further analysis was beyond project resources. 1 Further examples of fission product transport behavior are illustrated in Figures 5.1 68 through 5.1 70, The first two figures are for cesian and barium with a small brcak of .00924 ma and the third figure is for barium with a small leak of .000799 m2, t
It was expected that the decontamination factors for a drywell head break would be lower than for a drywell or vetvoll wall break because the flows from a drywell head break only had to flow through the refueling bay before leaving the reactor building. But this was not true for all classes when a large break was involved. The large drywell head break calculation, Case 9, predicted higher decontamination factors for some classes. Case 9 predicted two small burns shortly af ter the containment failed, which were relatively ineffective in the overall transport of fission products, compared to much more substantial burns predicted in the j corresponding wetwell and drywell walls breaks. 5.1.2.3 Combustion Ignition Limit Parameters , The sensitivity of fission product transport to the combustion ignition limits is shown by the comparisons of Cases 1 and 6 in Table 5.1-33, of Cases 9 and 9A (also in Table 5,1-33), and of Case 4 with the base case in Table 5.1 34 In Case-6, the deflagrations were deactivated by setting the hydrogen ignition limit'at 1006. In Case - 9A, the oxygen ignition limit was lowered from the usual.5% to 4%. The base case calculation used a 10% ignition hydrogen limit. The . dramatic effect that combustion has on the transport of fission products through the ' reactor building is clearly shown in Figure 5.171. The explosive flows following a burn expelled much of the reactor building air carrying the airborne fission products out with it. - Revaporization of deposited aerosols during a hydrogen burn may have contributed to the rapid transport for the cesium, iodine, and cadmium classes, but i W 5-26
-~w --, .. , . . - - . - - , , , - . _ ,, .-%;. , , , , -m-,. e,-.,i. ,+ y- ., e ,-.
insufficient data was retained in these calculations to assess the importance of revaporization during the burns. Resuspension of deposited aerosols not currently nodeled but might have also contributed, had the model been available. During the large drywell head break calculation, Case 9, following two small early burns, further combustion was prevented by a lack of oxygen within the refueling bay. The minimum oxygen ignition limit was set at St but the oxygen mole fraction could not quite reach 54 because of the continuous purging of the refueling bay by the containment flows, which were completely devoid of oxygen. Therefore, the sensitivity calculation, Case 9A, was run from a restart file at 19.4 hours (1.2 hours after containment failure) to determine the sensitivity of the releases to the environment to the oxygen ignition limit by reducing the ignition limit to
- 44. The smass releases to the environment were ignificantly enhanced by the resulting additional burn. This is an example of a threshold sensitivity, where a small change in input can make a significant change in the results.
l The effect of the hydrogen ignition limit can be seen by comparing the ,q hace case (see section 5.1.1) and Case 4. The base case was rur with a i 10% litnit and Case 4 with 5%. This comparison 10 shown for the xenon class gases in Figure 5.1 72 Case 4 had more frequent but less severe burns than the base case. But it was difficult to compere the two calculations because Case 4 was not run out as far as the base case. It appears that if Case 4 had been run out to the same termination time as the base case that the environmental releases might have been about the same. The base case and Case 4 results are compared by class in Table 5.1 34. This table provides a i. ans of computing the actual masses for the various sensitivity calculations from the masses L'eported for the base case in section 5.1.1. 1ur example, the base case xenon mass in the environment dt the end of the calculation was 202.4 kg (Table 5.1-5). The corresponding mass for Case 4 is 44.5 kg (202.4 multiplied .22 from Table 5.1 34), and for Case 1, the mass is 296.9 kg (44.5 divided by .15 from Table 5.1 27). 5-;7
Table 5.1-1 High Pressure Short Term Station Blackout Events Event Time Event Description Seconds Hinutes Hours
- 0. 0.0 0. Accident Initir.ted, Reactor Tripped, Isolated 2117. 35.3 0.59 Collapsed Water Level Below Top Active Fuel 4546. 75.8 1.26 Fission Products Released from Rin6 1 Rod Cap (First Fission Products Released) 4875. 81.2 1.35 Fission Products Released from Ring 2 Rod Cap 5325. 88.8 1.48 First Core Material Relocated 5954. 99.2 1.65 Fission Products Released from Ring 3 Rod Gap 6460. 107.7 1,79 Core Channel in Ring 1 Blocked 6989. 116.5 1.94 Core Channel in Ring 2 Blocked 11191. 186.5 3.11 Core Channel in Ring 3 Blocked 11541. 192.3 3.21 Fission Products Released from Ring 4 Rod Gap 13951, 232.5 3.88 Core Plate.in Ring 1 Failed 13975. 232.9 3.88 Lower Plenum Pool Dried out 14107. 235.1 3.92 Lower Head Penetration in Ring 1 Failed (Reactor Vessel Failed) 14674, 244.6 4.08 First Debris Ejected from RV to DW Pedestal 15740. 262.3 4.37 Core Plate in Ring 2 Failed 15869. 264.5 4.41 Drywell Podestal Floor Drain Pipes Failed 15879. 264.7 4.41 Wetwell Pedestal Received Core Debris '
15931. 265.5 4.43 Lower Head Penetration in Ring 2 Failed 16926. 282.1 4.70 Debris Ejected 17224 287.1 4.78 Core Plate in Ring 3 Failrd 17451, 290.9 4.85 Lower Head Penetration in Ring 3 Failed 17862, 297.7 4.96 Debris Ejected 1P<n8 310.1 5.17 Debris Ejected it. /, 310.2 5.17 Debris Ejected 19813. 330.2 5,50 Debris Ejected 21917. 365.3 6.09 Core Channel in Ring 4 Blocked 22064. 367.7 6.13 Containment Wetwell Pressure Exceeded 60 psig 22486. 374.8 6.25 Debris Ejected 22907. 381.8 6.36 Debris Ejected 22927, 382.1 6.37 Debris Ejected 22942. 382.4 6.37 Debris Ejected 22945, 382.4 6.37 Debris Ejected 22953, 382.6 6.38 Debris Ejected - 24617. 410.3 6.84 Core Plate in Ring 4 Failed 24769. 412.8 6.88 Lower Head Penetration in Ring 4 Failed 29295. 488.3 8.14 Containment Wetwell Pressure Exceeded 85 psig 34115. 568.6 9,48 Debris Ejected 37803. 630.1 10.50 Containment Wetwell Pressure Exceeded 145 psig 65463. 1091.0 18.18 Containment Failed from Wetwell Leakage 68101. 1135.0 18.92 Deflagration Burn in Upper Reactor Bld Unit 2 68978, 1149.6 19.16 CCI Ablation Exceeds WW Pedestal Wall Thickness 81060. 1351.0 22.52 Deflagration Burn in Upper Reactor Bld Unit 2 (Burn Propagated into Refueling Bay and Unit 1) 115712. 1928.5 32.14 Calculation Terminated l 5 28
~ , - , - ,w., e - p :p w,v--n , - - - ---- - - --- -
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y . Table 5.1-2. High-Pressure Short-Term Station Blackout Core Meltdown Teorperatures (K) Ring 1 Ring 2 Ring 3 . Ring & FUE1JCIAD CAlt CANB STRU DEB
- FUEL CIAD CAN CAltB STRU CEB
- frE1, CLAD CA!t CAND STRC DES
- TUEI,CIAD CME CANB STRU DEB B.* D 13 0 987 907 888 685 0* O 912 840 824 650 0* O 800 744 73 1 606 0* 0 675 643 635 572 0 588 12 1176 1173 1128 1122 1069 0
- 1058 1955 1009 1002 939 0* 895 893 85C 8 1 773 0* 721 720 687 680 633 0 589 11 1147 1143 1086 1079 998 0
- 1037 1034 975 967 883 0* 882 880 825 8. "a O* 716 714 673 666 616 0 588 10 1042 1037 960 953 '838 0* 958 954 877 870 759 0* SIS 8?S 75? ~*. -
s* 689 687 641 635 589 0 575 9 915 910 806 798 673 0* 853 850 751 745 638 0* 758'.755 674 6Sv w7 0* 655 653 604 600 566 0 574 8 677 673 581 :579 555 0* 656 653 575 574 554 0* 624 622 568 567 553 0* 590 589 381 560 553 0 563 I $66 561 557 557' 556- 0* 565 560 557 557 556 0* 563 500 557 557 556 0* 561 560 557 557 556 0 554 0 559 557 .557 556 0* 0 559 557 557 556 0* 0 559 557 557 556 0* 0 559 557 557 556 0 554
. ... ......... ... 553 0*- 553 ~ 0+ 553 0* 553 0 552 4 . T= 4546 Sec . 553 0* 553 0* 553 0* $53 0 552
- 3. = 1.26 Br . . 353 0* 553 0*. 553 0* 553 0 549 2 . First rission . 553 0* $53 0* 553 0* 553 0 549 1 . Product Release . 553 0* 553 0* 553 0* 553 0 549
+ .... ....... .. .. ..... ... .. ....... .. ... ....... ............................ ............... ... ...... .....................
0 0* 0 1340 1300 1240 1193 0* O 1049 1039 973 664' 0* 0 798 790 750 669 4 650 y' 13 0 0 0 'O. 0 660 12 0 0' O O O O
- 1997 1995 1952 1988 0 0
- 1448 1451 1503 1426 1411 0* 931 931 1000 916 894 11 0 0 1588 1548 1514 0
- 1857 1832 1792 1800 0 0
- 1374 1376 1407 1346 1326 0* 916 916 961 893 866 0 657 10 0 0 2095 2098 0' 0
- 1801 1800 1786 1754 1700 0
- 1234 1234 1302 1191 1172 0* 8e6 866 892 836 801 0 571 9 0 0 2093 2098 0 2098
- 1531 1538 1586 1595 1656 0
- 1084 1083 1121 1043 1000 0* 813 812 805 773 729 0 568 8 2106 2098 2085 2098 0 2098.* 1242 1243 1358 1176 1139 0* 963 961 967 903 833 0* 757 756 729 708 654 0 566 7 1690 1708 1648 1619 1532 1733.* 883 881 906 804 .702 0* 772 770 719 742 626 0* 663 662 :626 619 380 0 564 6 0 847 793 .779 714 892
- O 572 582 561 553 0* O 565 560 558 553 0+ 0 560 556 556- 553 0 557 5 .. ..... ........ $ ~,6 555
- 553 0* $53 0* 553 0 553 4 . T= 6461 Sec - . 353 0* 553 0* 553 0* 553 0 553
- 3.
- 1. 79 Hr = . 553 0* 553 0* 553 0* 553 0 549 0* 553 0* 553 0* 553 0 549 2 . First.Blochese . 553
- 1. (Ring 1) . 553 0* .553 0* 553 3* $53 0 549 13 0 -0 0 0 0 0* 0 0 0 0 C 0* O- 0 1513 1540 1551 0* O 1213 1219 1184 1164 0 920
-12 0- 0 0 0 0 0* 0 0' O O" 0 0* 0 0 131S 0 .O O
- 1219 1219 1205 1208 1203 0 919 11 0 0 0 0 0 0* 0 0 0 0 0 0* 0 0 1316 0 0 S
- 1215 1216 1205 1210 1206 0 918
,10 0 0- 0 0 0 0* 0 0 0 0 0 0* O O 1513 0 0 0
- 1216 1217 1234 1223 1222 0 9 13 9 'O O 2828 2830 0 2822
- 0 0 2673 2669 0 2672
- O- 0 0 0 0 2233
- 1229 1233 1452 1264 1266 0 695 8 0 0 3102 3110 0 3112
- 0 0 2824 2831 0 2835
- 0 0 0 2207 0 2149
- 1107 1111 1476 1129 1133 0 687
'O. 0 2594 2599 2597 2604
- O O 2314 2313 2311 2316
- 1942 1940 1943 1921 1902 1941
- 839 839 1030 804 800 0 650 7
0 1916 1781 1648 1379 1931
- O 1701 1587 1361 1296 1706
- O 1626 1516 1401 1365 1648 * ,0 700 918 692 715 0 647 6
969 1041 *- 922 954 *. 885 962
- 599 0 583 5 ...... 4...........
0* 561 0* 559 0* 553 0 573 4 . T= 12001 Sec . 561 0 550
= 554 0* 554 0* $54 0* 554 3 3.33 Er .
. 554 0* ' 154 0* 554 0* $54 0 550 2.
1, . 554 8'
- 554 0* 554 0* 554 0 550
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1 Table 5.1-3. Reactor / Turbine Building Peak Pre sures and Temperatures for the liigh Pressure Short Term Station Blackout. Burn 1 at 18.9 hours (68100 seconds)- ) i - Location Differential Pressure Temperature i Pa Psi K F Lower Unit 2 48600 7.0 439 331 Upper Unit . 49400 7.2 1293 1868 i Unit 1 49400 7,2 401 53 Refueling Bay 35200 5.1 440 31 Steam Tunnel / Turbine 2 7430 1.1 315 108 Steam Tunnel / Turbine 1 6910 1.0 314 105 Burn 2 at 22.5 hours (81060 seconds) location Differential Pres'qrg ,emnerature Pa Psi -K F Lower Unit 2 128000 18.6 659 -726 Upper Unit 2 133000 -19.3 1308 1894 Unit i 136000 19.7 1224 '1743 Refueling Bay 131300 19.1 1078 1481 Steam Tunnel / Turbine 2 16000 2.3 348 166 Steam Tunnel / Turbine 1 22300 3.2 382 228 l 4 1 l 5-32 l +
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- i Table 5.1 4.
Sununary of Rad' ionuclide Releases and Transport Initial Fraction Ratio of Ratio of Fraction . Reactor Bld. Class Core Relea i.1 In to Ex Total to Retained Decontaminaticn NO. Rep. Inventory
- from Fuel Vessel Radioactive by Primary Factor' (Kg) 1 Xe 463.7 .989 28.3 1. .0045 1.8.
! 2 Cs 268.4 .986 6.51 1.13 .48 4.2
!- 3 Ba 207.5 .564 2.57 1. .76 3.4 4 I 20.93 .975 40.4 1. .24 1.9 i 5 Te 40.79 .967 50.8 1.12. .78 4.7 . 6 Ru 307.0 6.89E-4 3.44E5 6.41 .69 .8.9 7 Mo 350.6 .0172 0409 313, .82 -3.3 l 8 Ce 594.0 . 52E-4 33.9- 55.0 .- 6 8 3.1
- 9 La 571.1 .0306 47.5 1. .70 3.6
- 10 U 132390. 5.49E-5 1.06 1.13 .69 7.9 11 Cd 1.407 .625 145, 1. .15 2 '-
j 12 Sn 8.587 .780 3.92. 1. .77 11, 16 Cs1- 0. .74 4. 3.-
- Radioactive Fissio.. Fragments-(except Uranium) i i
? w l' 5-33
--u v .-& r%-d
Table 5.1-5. Distribution for the Xenon Class Radioactive Mass (Kg) Vessel Failure Containment Failure Before After Before After , Releases from Fuel Primary 4.267E+02 4.271E+02 4.431E+02 4.431E+02 Cav.ity 0 0 1.567E+01 1.567E+01 Overail Distribution Primary 3.125E 01 1.211E 02 8.497E+00 2.005E+00 Containment 4.264E+02 4.271E+02 4.503E+02 8.891E+01 Reactor Bld 0 0 0 1.591E+02 Turbine Bld 0 0 0 6.527E+00 Environment 0 0 0 2.024E+02 (ontainment Distribution Drywell 1.542E+02 6.858E+01 2.185E+02 4.985E+01 Upper Pedestal 2.218E+00 1.583E 01 4.276E+00 1.410E+00 Lower Pedestal 6.936E+09 1.498E+01 5.247E-19 1.233E-01 Wetwell 2.631E+02 3.434E+02 2.276E+02 3.753E+01 Suppression Pool 0 0 0 0 Table 5.1 6. Distribution for the Cesium Class Radioactive Mass (Kc) Vessel Failure Containment Failure Before Aftar Before After Releases from Fuel Primary 2.0;8E+02 2.049E+02 2.083E+02 2.083E+02 Cavity 0 0 3.478E+01 3.478E+01 Overall Distribution Primary 1.130E+02 1.015E+02 1,017E+02 9.984E+01 Containment 9.109E+01 1.025E+02 1.286E+02 1.304E+02 Reactor Bld 0 0 0 7.562E-02 Turbine Bld 0 0 0 5.223E-04 Environment 0 0 0 2.413E-02 Containment Distribution Drywell 4.589E-02 6.670E+00 8.805E+00 9.776E+00 Upper Pedestal 4.534E-01 8.180E-01 1.373E+00 1.891E+00 Lower Pedestal 1.594E-03 7.834E-03 3.904E-02 6.115E-02 Wetwell 9.059E+01 9.45.!+01 1.184E+02 1.187E402 Suppression Pool 9.041E+01 9.386E+01 1.146E+02 1.147E+02 5-34
! Table 5.1-7. . Distribution for the Barium Class l Radioactive Mass (Kg) 4 Veerel Failure Containment Failure Before After Before After i Releases from Fuel . Primary 8.393E+01 8.393E+01 8.424E+01 8.424E+01 i Cavity 0 0 3.271E+01 3.276E+01 3 Overall Distribution , Primary 6.488E+01 6.488E+01' -6.491E+01 6.449E+01 4 Containment 8.730E+00 8.739E+00 3.910E+01 3.951E+01 Reactor Bld 0 0 0 3.314E-02
- Turbine Bld =0 _ _0 0 3.180E-04 l Environment 0 0 0 1.367E 02
! Containment Distribution j Drywell 1.086E-02 1.218E-02 1.058L+00 1.281E+00 Upper Pedestal 4.348E-02: 4.428E 02- 3.596E 01 4.715E 01 l Lover Pedestal 3.101E 04 1.545E-03 1.567E+01 1.565E+01 i' Wetwell 8.675E+00 8-681E+00-
. 2.202F+01 2.211E+01 Suppression Pool 8.629E+00 8.633E+00 2.039E+01 2.046E+01 i-1 Table 5.1-8.
Distribution for the Iodine Class ) Radioactive Mass (Kr) Vessel Failure Containment Failure Before After Before After l Releases from Fuel , Primary 8.774E 05 8.774E-05 -1.150E-04 .1.377E 04
- Cavity 0 0 0 0
- Overall Distribution
- Primary 3.817E-05 3.817E-05 3,998E 3.272E-05 d
Containment- 4.956E 05 4.956E-05 :7.504E-05 1.050E 04 Reactor Bld 0 0 0 1.824E-09 Turbine Bld 0 -0 0 7.251E-11 Environment _ 0 0 0 2.039E 09 , Containment Distribution-4 Drywell 8.550E-11 8.832E-11_ 8.832E-11' 3.195E 08-Upper Pedestal 2.716E-08 2.734E-08 2.550E-05. 5.540E-05 Lower Fedestal 7.465E-12 7.465E 12 2.333E-27_ 5.471E-14 } Wetwell -4.954E-05 4.954E-05 4.954E-05 4.955E-05 Suppression Pool 0 0- 1.374E-13 9.235E _ 5-35.
------.7 Table 5.1-9.
Distribution for the Tellurium C'_ ass Radioactive Mass (Kr) e vessel Failure -Containment Failure Before After Before After-Releases from Fuel Primary 3.738E+01 3.742E+01 3,868E+01 3.868E+01 Cavity 0 0 3.689E 01 ~7.019E 01 Overall Distribution Primary 3.161E+01 3.160E+01 3.167E+01 3.070E+01 Containment 4.539E+00 4.586E+00 5.834E+00 7.060E+00
- Reactor Bld 0 0 0 9.486E 02 Turbine Bld 0 0 0 3.697E 04 Environment 0 0 0 2.560E-02 Containment Distribution Drywell 1.244E 02 3 756E 02 6.389E 01 1.185E+00 Upper Pedestal 1.012E 01 1.035E-01 2.772E-01 -5.239E 01 1.ower Pedestal 5.733E 04 1.024E 03 5.419E-02 2.098E-01 Wetwell 4.425E+00 4.444E+00 4.864E+00 5.141E400 Suppression Pool 4.407E+00 4.421E+00 4.799E+00 4.962E+00
=.,
Table 5.1-10. l Distribution for the Ruthenium Class Radioactive Mass (Ke) Vessel Failure Containment Failure Before- After Before After Releases from Fuel Primary 2.114E-01 2.114E-01 2.114E 01 2.114E-01 Cavity O. 0 5.885E-07 6.152E Overall Distribution Primary- 1.472E-01 1.472E 1.471E-01 1.471E Containment 3.083E-02 3.084E-02 3.074E-02 3.079E 02 Reactor Bld 0 0- 0 4.222E-09 Turbine Bld 0 0 0 3.849E-12 Environment 0 0 0 5.343E Containment Distribution Dryvell 9.607E-06 6.284E 06 1.014E-05 1.014E-05 Upper Pedestal 1.218E -l.223E-04 2.306E-05 -7.581E 05 Lower Pedestal 2.(30E-07 1.406E-06 8.768E-07 8.864E-07 Wetwell 3.070E-02 3.071E-02 3.0iOE-02 3.070E-02 Suppression. Pool 3.066E-02 3.067E-02 3.069E-02 3.069E-02 5 36
. . ~_. - - . . . . - - .. . . . . . -
Table 5.1-11. Distribution ~for the Molybdenum Class Radioactive Mass -(Ke) ' Vessel Failure Containment Failure Before After Before After Releases from Fuel Primary 2.353E-01 2.353E-01 2.374E-01 2.374E 01-Cavity 0 0 5.800E+00 5.802E+00 Overall Distribution Primary 1.673E-01 1.673E-01 -1.965E 01 1.965E 01 Containment 3.169E 02 3.169E-02 5.398E400 5.398E+00 Reactor Bld 0 -0 0 3.252E-04 Turbine Bld 0 0 0- 3.847E-06 Environment 0 0 0 1.409E 04 Containment Distribution Drywell 1.178E 05 7.547E 06 1.786E-01 .1.786E 01 Upper Pedestal '1.360E 04 1.365E-04. ~8.251E-02 8.260E 02 Lcwer Pedestal 3.711E 07 1.581E-06 2.468E+00 -2.468E+00 Wetwell 3.154E 02 3.154E-02 2.669E+00 -2.669E+00 Suppression Pool 3.149E-02 3.150E 02 2.449E+00 2.449E+00 Table 5.1-12. l Distribution for the Cerium Class Radioactive Mass (Kc) Vessel Failure _ Containment Failure Before After Before After Releases from Fuel Primary 2.033E-01 2.033E-01 2.033E-01 2.033E-01 Cavity 0 0 3.957E 03 5.999E-03 Overall Distribution Primary 1 402E-01 1.402E-01 l'. 402 E - 01 1.401E-01 Containment 3.111E-02 3.112E-02 3.479E-02 3.649E-02 Reactor Bld 0 0- O. :2.326E-04 Turbine Bld 0 0 0 2.740E-06 Environment 0 0 0 1.099E Containment Distribution Drywell 9.628E 06 6.310E-06 3.487E 4.437E-04 Upper Pedestal 1.141E-04 1.146E-04 l2.278E-04 3 282E-04 Lower Pedestal- 2. 626E- 07 1.418E 06 1.698E-03 '2.516E Vetwell 3.099E-02 3.099E-02 3.252E-02 3.320E-02 Suppression Pool .3.095E-02 3.095E 02 3.223E-02 3.267E 5-37
Table 5.1-13. Distribution for the Lanthanum Class Radioactive Mass-(Kr) Vessel Failure Containment Failure Before After Before- After Releases from Fuel Primary 1.708E+01 1.708E+01 1.710E+01 1.710E+01 Cavity 0 0 -3.579E 01 3.596E 01-Overall Distribution Primary 1.203E+01 1.203E+01 1.204E+01 1.204E+01 Containment 2.431E+00 2,432E+00 2.750E+00 2.755E+00 Reactor Bld 0 0 0- l'.777E-04 Turbine Bld 0 0 0- 1.481E Environment 0 0 0 6.770E 05 Containment Distribution Drywell _ 9.144E 04 6,001E-04 1.512E-02 1.519E 02 Upper Pedestal 9.606E 03- 9.646E-03 6.533E 03 1.072E 02-Lower Pedestal 2.610E 05 1.313E-04 1.574E 01 1.581E 01' Wetwell 2.421E+00 2.421E+00 2.571E+00' 2.571E+00 Suppression Pool 2.417E+00 2.417E+00 2.554E+00 2.554E+00 Table-5.1-14. Distribution for the Uranium Class Radioactive Mass (Kr)
- Vessel Failure Containment Failure l Before After Before After l
-Releases from Fuel Primary 3.742E+00 3.742E+00 3.742E+00 3.742E+00 Cavity 0 0 1.475E+00 :3.524E+00 ,
Overall Distribution Primary 2;555E+00 2.555E+00 2.565E+00 2.585E+00 Containment 5.814E-01 5.814E-01 1.941E+00 3,603E+00-Reactor Bld -0 0 0 3.127E-01 Turbine Bld 0 0 'O -5.024E 04 Environment 0: 0 0 4.549E Containment Distribution i Drywell . .2.045E-04 1.311E-04' 7.031E 1.386E-01 L . Upper Pedestal 2.042E 03 2.051E-03 ~3.567E-02 6.156E 021 l -Lower Pedestal 5.517E 06. 3.030E-05 6.919E-01 1.458E+00 Wetwell 5.791E-01 5.792E 01 1.143E+00' ,1,945E+00 Suppression Pool 5.782E-01 =5.783E 01 1.074E+00 11.278E+00 5-38
Table 5.1-15. Distribution for the Cadmium Class Radioactive Mass (Kr) Vessel Failure Containment Failure Before After Before- After Releases from Fuel _ Primary 8.657E-01. 8 .660E-01 8.735E-01 8.735E-01 Cavity 0 -0 8.950E 05 6.007E 03-Overall Distribution Primary 2.518E-01 _2.455E-01 2,418E 01 1.349E-01 Containment 6.139E-01 6.204E 01 6.149E 01 7.220E-01 Reactor Bld 0 0 0 .2.584E-03 Turbine Bld 0 0 0 5.084E-05'
' Environment 0 .0 0 1.887E 03-Containment Distribution Drywell- 4.051E-02 2.767E-02' 6.631E-02 9.463E-02 Upper Pedestal 1.681E-04 4.636E 04 2.115E-02 8.303E 02 Lower Pedestal 5.971E 1.201E-02 2.4SSE 05 5.030E-03 Wetwell 5.726E-01 5.803E-01 5.274E-01 5.393E-01 Suppression Pool 1,512E 01- 1.677E 01 4.274E-01 4.357E-01~
Table 5.1 16.- Distribution for the Tin Class Radioactive Mass (Y2) Vessel railure- Containment' Failure Before After Before After Releases from Fuel
-Primary 5.285E+00 5.287E+00 5.333E+00 5.333E+00' Cavity 0 0 1.767E-02 1.362E+00-Overall Distribution Primary 4.086E+00 4.086E+00 4.118E+00 4.133E+00 Containment 4.520E-01 4'543E 4.789E-01 1.438E+00 Reactor Bld 0 0 0- -
3.370E-01 Turbine Bld 0 0 -0 1.379E Environment 'O O O 3.232E-02 Containment Distribution Drywell 3.864E-04 1.809E-03 9.440E-03 6.788E 02 Upper Pedestal 3.216E-03 3.375E-03 3.069E-03 '2.245E-02 Lower Pedestal 1.627E-05 3,884E-05 1.172E-02 1.145E-01 Vetwell 4.484E-01' 4.490E-01 4.546E-01 1.234E+00 Suppression Pool -4.475E-01 4.479E-01 4.535E 5;990E 5-39 w -. - - - - ,
Table 5.1 17. Distribution for the Cesium lodide Class Radioactive Mass (Kg) Vessel Failure Containment Failure Before After Before~ After < Releases from Fuel Primary 3.939E+01 3.942E+01- 4.093E+01 4,093E+01 Cavity 0 0 8.285E 01 8.285E 01 Overall Distribution Primary 3.325E+01 3.322E+01 3.240E+01 2.075E+01 Containment 5.136E+00 2.124E401 -7 919E+00 9.461E+00 Reactor Bld 0 0 0 6.287E 02 Turbine Bld 0 0 0 3.734E-04 Environment 0 0 0 1.936E.02 Containment Distribution Drywell 2.286E 02 6.045E 02 1.212E+00 2.082E+00 Upper Pedestal 1.041E 01 1.615E+01 5.550E 01 9.753E-01 Lower Pedestal 1.025E 03 2.026E-03 5.698E-03 3.500E 02 Wetwell 5.008E+00 5.035E+00 6.147E+00 6.369E+00 Suppression Pool 4.969E+00 4.989E+00. 5.971E+00 6.135E+00 e 5-40
Table 5.1.18. Early and Late Transport of Fission Products from the RCS-Class Early. Total Late Late. - Relative
- No. Rep. Transport Transport Transport Fraction Impact (Kg) (Kg) (Kg) 1 Xe 4.271E+02 4.411E+02 1.405E+01 .0318 1.032 2 Cs 1.035E+02 1.085E+02 5.061E+00 .0466 1.037 3 Ba 1.940E+01 2.010E+01 7.055E<01 .0351 1.014 4 I 4,956E 05 1.050E 04 5.543E-05 .5279 2.118 5 Te 6.190E+00 8.361E+00 2.171E+00 .2597 1.312 6 Ru 6.538E-02 6.551E 1.374E-04 .0021 1.002 7 Mo 6.916E 02 4.211E-02 . 2.705E 02 (.6424)" 0.995 8 Ce -6.418E 02 6.425E-02 6.709E-05 .0010 1.001 9 La 5.135E+00 5.150E+00 1.490E 02 .0029 -1.003 10 U 1.207E+00 1.177E+00 .982E-02 ( 0253)- 0.994 11 Cd 6.205E-01 7.386E 01 1.181E 01 .1599 1.188-12 .Sn 1.225E+00 1.224E+00 -9.472E 04 (.0008) 1.000 16 -Cs1 6.583E+00. 1.056E+01 3.978E+00 .3766 1.537
- Effect of Including Late RCS Releases on Containment Releases Relative to Or.ly-Early Releases
** Ratio of Late In-flow to RCS to the To*al RCS Release I
5-41
Table 5.1-19. Containment Distribution of Fission Products Containment. Location within Containment Class Fraction of (Fraction of Containtnent Mass) No. Reo. Total Release Drywell Vetwell luopression Pool 1 Xe .194 .577 '423
. .0 2 CS .536 .090 .910 .880 3 Ba .338 .044 .956 .518 4 I .762 .528 ,472 .088 5 Te .179 .242 .758 ,703 6 Ru .146 .003 .997 .99?
7 Mo .894 .048 .952 .454 8 Co .174 .021 .979 .895 9 La .158 .009 .991 .927 ..
.355 -!
10 U .496' .056 ,944 11 Cd .821 .246 .754- .604 ! 12 Sn ,215 .053 .937 417 l 16 Cs1 .227 .323 .677 .649 ! 5-42
Table 5.1-20. Airborne, Pool, and Deposition Distribution Fractions-of Fission-Products for the _ Reactor Cooling System at End of Calculation Class Airborne - f_221 Dooosited.
- 1. Xe 1. O. -O.
2 Cs SE 04 .0276 .9719 3 Ba .0001 .0083 .9916 4 I .0139 .0172 .9689 5 Te .0010 .0205 .9785 6 Ru 3E-11 .0117 .9883 7 Mo 1E 08 .0001 .9999 8 Ce 9E-07 .0118 .9882-9 La 8E 09 .0112 .9888 10 U .0001 .0122 .9876
.11 Cd 1E .0072- .9928 12 Sn 8E 05- .0092 .9907 16- Cs1 .0027 .0210 .9763 Table 5.1-21.
Airborne, Pool, and-Deposition Distribution Fractions of Fission Products for the Containment at End of Calculation Class Airborne f_q21: Deposited-1 Xe 1. -0. 0. 2 Cs .0016 .8800 .1184 3 Ba .0009 .5179 .4812.
- 4. I 3E 05 .0880 .9120 5 Te .0223 .7028 .2749 6 Ru 2E-07 .9968- .0032 7- Mo 4E-05 .4537 .5463 8 Ce .0040 .8953 .1007' 9 La 4E-05 .9271 .0729 10 U- .1109- .3547 .5344 11 Cd .0020- .6035 .3945 12 Sn .2831 .4164 .3005 16 Cs1 .0160 .6484 .3356 5-43 s,
Table 5.1 22. Airborne, Pool, and Deposition Distribution Fractions of Fission Products for:the= Reactor / Turbine BuildingL at End of Calculation
.Q,1,ggg Airborne f2gl Denosited 1 Xe 1. O. O.
2 Cs .4861 .1548 .3590 3 ~Ba ,4338 .1674 .3988 4 1 .3364 .2606 .4030 5 Te .6645 .1432- .1923 6 ku .8999 .0745 .0256 7 Mo - .7242 .1798 .0960-8 Ce .7240- .1798 .0963 9 La .7380 .1715 =.0905 10 U .8912 .0796 .0293 11 Cd .5149 .1945 .2907 12 Sn .9153 .0674 ,0172 : 16 Cs1 .4825 .1561 .3614 l Table 5.1-23. Airborne, Pool, and Deposition-Distribution Fractions of Fission Products for the Entire Plant-at End of Calculation Clasg Airborne fool Deposited 1 Xe 1. O. - O'. 2 Cs .0014 .5101 .4885 3 .Ba .0007 .2018- .7975 4 I .0034 .0711' .9255 5-- Te .0073- .1480 .8447 6 Ru. 6E 08 .1822 .8178! 7 Mo 2E 05 .0991 .9009 8 Ce- .0024 .1942 .8034 9 La 2E-05 .1818 '.8182 10 U .1107 .2038 .6855 11 Cd .0055 .5075 ,4870' 12 -Sn .1260 .1111'- .7630 16 CsI .0071 .1685 .8245-44
Table 5.1 24. Fractional System Distribution for Deposited Fission _ Products at the End cf Calculation _ _ Class EqE Containment Reactor Bld. ,, 1-- Xe -- -- -- 2 Cs .8625.. .1371 .0002 3 Ba .7707 .2291 .0002 4 I .2487 .7513 6E-06 5 Te .9388 .0606 .0006 6 Ru- .9993 .0007 7E-10 7 Mo .8677 .1323 1E-06 8 Ce .9740 .0259 .0002 9 La .9834 .0166 1E 10 U .5689- , .d 9 0 - .0020 11 Cd .3192 .6790 .0018 12 Sn .9034 .0954 .0013 16 Cs1 .9038 .0956 .0007 Table 5.1 25. Fission Product Vapor to Aerosol Mass Ratios in the Atmospheres;at End of Calculation Cla s . B.QS . Containment Reactor Bld. 1 Xe 1. 1. 1, 2 Cs 3.99 2.9E-4 0 .~ 3 Ba 3.78 9.3E 0;
-4 I 1.9E6 72.9 .0616 5 Te 3.99 9,8E-4 0.
6 Ru -4.3E-6 .8.4E-9 0. Mo- 1.E-12 6.4E 0. 8 Ce .0205 .0352 0. 9 La 0. O. O. 10 U 0. O ', -0. 11 Cd .272 3.6E-6 0. 12 :-Sn .0445 .0311 .O. 16 CsI 4.95 .00191 0. 5 45
__ _____ = . . . __ .. Table 5.1-26. High-Pressure STSB Sensitivity Depressurization Modified Calculation gentainment Depressurization Qgag Break Area Termination limg* Ccroletion (m2) (hr) (hr) (%) Base .000199 32.14 12.96 60.9 1 .0924 19.73 1.55 Complete 2 .00924 21.21 3.03 90.1 3 .0924 18.89 0.71 Complete , 4 .000799 20.41 2.23 16.4 5 .00924 21.29 3,11 89.0 ; - 6 .0924 18.96 0.77 Complete 7 .000799 22.19 4.01 23.2 8 .000799 21.76 3,58 21.3 9 .0924 20.58 2.39 Complete 9A .0924 19.55 1.36 Complete 10 .00924 21.29 3.11 88.9 11 .924 18.51 0.32 Complete Containment Failed at 18.18 hours at 1,4459E6 Pa (195 psig)
- Calculation Termination Time After Containment Failure 5-46
l Table 5.1-27. High Pressure hid5 Break Area Sensitivity Results: for a Vetwell Break.and 5% Ignition Limits L s Fission Product Releases from Containment Masses Normalized to Case 1 Case Modiffed 1 2 3 4 5 6 7 81 9 10 11 12 16 HIL 6Ita _Xe fa la 1 12 A .32 GE .la U Gd 10 Gal (m2) 4 7.99E-4 .25 .02 .01 0, ,02 .13 .20 .16 ,11 ,11 .01 .05 .01 2 .00924 .91 .52 95 0. .51 1.1 1.1 1.4 1.3 1,3 1,0 1.3 .19 1 .0924 1.0 1,0 1,0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 11 .924 .99 .80 .74 1200 .62 .39 .71- .46 .35 .36 .48 ,30 .84 Fission Product Releases to Environment Masses Normalized to Case 1 Case Modified 1_ _2 3 4 5 6 7_ 8 9 10 11 12 16 Esb Artea ' Xt Ca an 1 Te _.h A ce _la U Gd S.n Gal (m2)
= 7.99E 4 .15 .02 0. . .03 .16 .16 .18 .16 .16 0. ,08 .01 2 .00924 .86 .84 1.4 --
1.0 2.0 1.2 1,9 2.3 2.2 .50 2.5 .34 1 .0924 1,0 1.0 1,0 -- 1.0 1.0 1.0 1,0 1,0' 1.0 1.0. 1.0 1.0 11 .924 .89 .55 .51 -- .55 .67 ,92 .71 .66 .66 .33- ,39 .54 Reactor Building Decontamination Factors Case Modified 1 2 3 4 5 6 7 8 9 10 11 12 '16 Est Arta _Xg Cs _Ea ,_1 _la: .Au .E.g . Ce _La U- Cd _In Eg1 (m2) 4 7,99E-4 2.5 2,9 4.1 -- 3,6 2.8 2.3 2.7 2.8 2.8 7.9 3.0 3.1 2 .00924 1.6 2.6 1.5 - 2.3 2.0 1.8 2.2- 2.3 2.2 3.6 2.4 2.2 1 .0924 1.5 4.1 2.2- -- 4.6 3.5- 1.9 3.0 3,9 3.9 1.8 4.4 3.9 11 .924 1.6 6.1 3.2 -- 5.2 2.1 1.5. 2.0 2.1 2.1 2.6 3.4 6.0 5-47
\
-Table 5.1 28. '
High Pressure STSB Break Area Sensitiity Results:- for a Drywell Break and,5% Ignition Limits Fission Product Releases from-Containment-Masses Normalized to Case 3' 10 11 12 16 <- Case Modified 1 2 3 4 -5 6- 7 8 9 HL h X1 _G1 .la 1 _IR Bu tin . ce La u ._Gd : _En Gal (m2) 8 7.99E 4 .31 .07 . 06 - 0 .- .10~ .22 .14 ' .24 .22 . 22 .07 .18 .05 ' 10 .00924 .93 .25 45 0, .32 1.3 .75- 1.0 1.3 1.3 .59 1.9 .09' 3 .0924 1.0 1.0 1.0 1.0 -1.0 1.0- 1.0_ 1.0 1.0 1.0 1.0 1.0 1.0 Fission Product Releases to Environment-Masses Normalized to Case 3 Case Modified 1 2 =3 4 5 6 7 8 9 10 11_ -12 16 H2 AI.ta Xe _Ga la I 12 _Ru H.2 - Ga La- _11 - Gd S.D- G.11 (m2) 8- 7.99E 4 .18 .07 07 -- .10 .13 .09 .13- .13 .13: .04 .12 .05-10 .00924 .92 .47 .72 -- .53 1.1 .68 ,96 1.1 .58- 1.6 .24 3 .0924 1.0 1.0 1.0 -- 1.0 1.0 1.0- 1,0 1.0 1.0 1.0 1.0 1.0 Reactor Building Decontamination Factors Case Modified 1 2 3 4 5- -6 7 8 9 :- 10 11 12- 16 UL_ AIra X2 G.1 _BA I .It _B.u ._ tin Ce -~ La U _Cd S.D' Gali (m2) 8= -7.99E-4 2.5 3.1 3.3- -
--3.3 3.0- 2.6-.3,0 -3.1-:3,1 -5.2 3.2 3.3 10 .00924 1.4 1.8 2.6 --
2.0 2.0 - 1. 7. 1.7 2.1 2.1' 2.9 2.6' 1.4-3 .0924 1.4 3.4 4.1 -. 3.2 1.7 -1.5 1.6 1.8 --1.8 .2.8 2.2- 3.5 O 48
Table 5.1-29. High Pressure STSB Break Area Sensitivity Results: for a Drywell Head-Break and 54 7gnition-Limits Fission Product Releases from Containment Masses Normalized to Case 9 Case Modified 1 2 3 4 5 6 7 8' 9 10 11 12 11 6 - , E2 Arta Xn Cs aa 1 Ta Al Mn ._01 La H Cd In' . Gal (m2) . . 5 7.99E-4 .90 24 .31 4E4 .27 .63 .59 .67 .59 .59 .62 .56 .09 7 .00924 .33 .07 .04 0 .08 .12 .12 .18 .11- .11 .09 .06 .05 9 .0924 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 t t Fission Product Releases to Environment Masses Normalized to Case 9 Case Modified 1 2 3 4 15 6 7 8- 9 10 .11 12 16 H2 6Iea XE _.G1 J.a 1 la ._Ru lig Ce La - U ._Gd S.n . Gal - (m2) , 5 7.99E-4 1.2- .31 .79 -
.69 1.5 .93 1.3 1,6 1.6 44 2.4. .13 7 .00924 .35 .12 .13 .-- .17 .24 .13 .26 .24 .24 .08 .25 .09 9 .0924 1.0 1.0 1.0 --
1.0 1.0 1.0 1.0 1.0 1.0 1.0' 1.0 1.0 Reactor Building Decontamination Factors Case Modified 1 2 3 4 5 6 7 8- 9 10 - 11 12 16 Eq . Area xe Cs aa Te Ru Mo Ce La U- Cd 1 S.D Gal (m2) 5 7.59E-4 1.1 2.5 2.2 1.1 L1.5 1.5 1,2- 1.3 1.5 1.5 2.2 1.8 2.3 7 .00924 1.4- 1.8 1.9 -- 1.9 1.8 1.8 -1.7 1.8. 1.8 .1.8 -1.8- 1.9 9 .0924 1.5 3.2 5.5 -- 3.8 3.6 L2.0 2.5 4.0 4~0- 1.6 7,5 3.4 5-49
Table 5.1-30. High Pressure STSB Break Location Sensitivity Results: for a Break Area of .0924 mz and 5% Ignition Limits Fission Product Releases from Containment Masses Normalized to Case 1 Case Break 1 2 3 4 5 6 7 8 9 10 11 12 16 UA__ Location le Sg A ,_.1 .le Eu A _Q.g la _U _Q.d 1D CEl 1 Wetwell 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 3 Dryw (1 .99 1.3 .16 0. 1.0 .59 1.1 .57 ,47 .48 2.0 .28 1.3 9 DW b- 1.0 1.5 .23 0, 1.3 1.2 1.4 .88 1.1 1.1 2.0 .94 1.5 Fission Product Releases to Environment Masses Normalized to Case 1 Case Break 1 2 3 4 5- 6 7 8 9 10 11 12 16 J UL _ Location Xe _Q g Ja __1 .le J.g A' Sg la U _Qd A Gal 1 Wetwell 1.0 1.0 1.0 -- 1.0 1.0 1.0 1.0 1.0 1,0 1.0 1.0 1,0 3 Drywell 1.0 1.6 .08 -- 1.5 1.2 1.4 1.1 1.1 1.1 1.3 .57 1.4 9 DW Head .99 1.9 .09 --
- 1. 7
. 1.2 1.4 1.0 1.0 1.0 2.2 .55 1.7 Reactor Building Decontamination Factors case Break 1 2 3 4 5 6 7 8 9 10 11 12 16 UL Location Xe Cs la _ _1 .It _ Ru A ce ._la U __Q.4 ln .Q.El 1 Wetwell 1.5 4.2 2.2 -- 4.6 3.5 2,0 3.0 4.0 3.9 1.8 4.4 3.9 3 Drywell 1,4 3.4 4.1 --
3.2 1.7 1.5 1.6 1.8 1.8 2.8 2.2 3.5 9 DW Head 1.5 3.2 5.5 -- 3.8 3.6- 2.0 2.5 4.0 -4 . 0 1.6 7.5 3.4 5 50 s
. Table 5.1 31.
High-Pressure STSB Break Location Sensitivity Results: for a Break Area of .00924 m2 and 5% Ignition Limits Fission Product Releases from Containment Masses Normalized to Case 2 Case Break 1 2 3 4 5 6 7 8 9 10 11 12 16 L Location lg ,_Gg ,_1g ___1 _,Tg lu ,_tLq ._Q.g _La __U E JD Gil 2 Wetwell 1.0 1.0 1.0' -- 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 10 Dryvell 1.0 .65 .07 -- .65 .66 .73 .42 .46 .48 1.2 .40 64 5 DW Head 1.0 .69 .08 --
.71 .66 .73 .42 .46 .48 1.2 .40 .67 Fission Product Releases to Environment Masses Normalized to Case 2 Case -Break 1 2 3 4 5 6 7 8 9 10 11 -12 16-L Location l g ,,_Q.1 la __1 la du }i2 -.ff _La U Ed S.D .G.El 2 Wetvell 1.0 1.0 1.0 --
1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0- 1.0 10 Drywell 1.1 .93 ,04 --
.78 .66 .79- .53 .52- .53 1.5 .37 1.0 5 DW Head 1.4 .70 .05 -- 1.1 .93 1.1 .71 .73 .74 2.0 .54 .65 Reactor Building Decontamination Factors Case Break 1 2 3 4 5 6 7 8 9 10 11 12 16' L Location lg Cs _}n I -. Te .Ru _tLq Ce _La U Cd _S.D Gal 2 Wetwell 1,6 2.6 1.5 -- 2.3 2.0 1.8 2.2- 2.3 2,3 3.6 2.4 -2,2 10 Drywell 1.4 1.8 2.6 --
2.0 2.0 1.7 1.7 2.1 2.1 2.9 2.6 1.4
'5 DW Head 1.1 2.5 2.2 --
1.5 1.5 1.2 1.3 1.5 1.5 2.2 1.8 2.3 5-51
Table 5.1-32. High Pressure STSB Break Location Sensitivity Results: for a Break Area of .000799 m2 and 5% Ignition Limits Fission Product Releases from Containment Masses Normalized to Case 4 _ Case Ereak 1 2 3 4 5 6 7 8 9 10- 11 12 16 No. Location lg Q.g ,,1g _I 3 .lu ,Jq ._Qg _.La 1' Q.0 .lu fil 4 Wetwell 1.0 1.0 1.0 -- 1.0 1.0 1.0 1,0 1.0 1.0 1.0 1.0 1,0 9 8 Drywell 1.2 5,9 1.6 -- 4.4 1.0 .80 .86 .94 .94 12. .92 10, 7 DW Head 1.3 6.3 1.8 -- 5.0 1.2. .86 .95 1.1 1.1 16. 1.1 10. Fission Product Releases To Environment Masses Normalized to Case 4 Case Break 1 2 3 4 5 6 7 8 9 10 11 12 16 No. Location Xe Qg ,.lg 1 ,, Ig l g . _tig Ce .,La U ..,,.Cd _Sn Sal 4 Wetwell 1.0 1.0 1.0 -- 1,0 1,0 1.0 1.0 1,0 1.0 1.0 1.0 1.0 3 Drywell 1.2 5.4 2.0 -- 4,9 ,96 .73 .80 .86 ,85 19. ,85 9,2 7 DW Head 2.4 10. 3.9 -- 9.7 1.8 1.1 1.5 1.6- 1.6 74. 1.7 17. Reactor Buildir.g Decontamination Fantors Case Break 1 2 3 4 5 6 7 8 9 10 11 12 16 L Location Xe Cs Ba I Te Ru Mo 'Ce _La U- Cd .JD .Q.A.1 4 Wetwell 2.5 2.9 4.1 --- 3.6 2.8 -2.3 2.7 .2,8 2.8 7.9' 3.0 3.1 :' 8 Drywell 2.5 3,1 3,3 -- 3.3 3.0- 2.6- 3.0 3.1 ~ 3,1- 5.2 3.2 3.3 7 DW Head 1.4 1.8 1.9 -- 1.9 1.8 -1.8 1.7= 1.8 1.8 .1,8 1.8 1.9 5-52
Table 5.1-33. High-Pressure STSB Ignition Limit Sensitivity Results: for a Break Area of .0924 m2 and both Wotwell and Drywell Head Breaks Fission Product Releases from Containment Masses Normalized to Case 1 or Case 9 Case Hydrogen 1- 2 3 4 5 -6 7 8 9 10 11 12 16 Nom Ignition _Xg ._Ca _Ba __1 _TI -.Au _lin fa La U _G.4 Sn Cal 1 5% 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1,0 1.0 1.0 6 100% .99 .90 .98 .71 .84 .65 .82 .71 .61 .62 .97 .59 .89 s 9 5% 1.0 1.0 1.0 1,0 1.0 1,0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 9A* 5% .99 .98 .91 1.0 .91 .66- .86 .76 .63 .63 1,0 .51 .99 Fission Product Releases to Environment i Masses Normal' ired to Case 1 or Case 9 Case Hydrogen 1 2 3 4 5 6- 7 8 9 10 11 _12 16 UL Innition _Kg Cs _Dg _1 Te Ru -- Mo Ce La U Cd SD Gal 1 5% 1.0 1.0 1.0 -- 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 6 100% .19 .12 .19 --
.15 .19- .25 .20 .18 .18 .11 .16 .09 9 5% 1.0 1.0 1.0 -- 1,0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 9A* 5% 1.4 1.7 3.4 --
2.3 1.9 1.4 1.6 2.0 2.0 1.4 2.7 1.7 Reactor Building Decontamination Factors case Hydrogen 1 2 3 4 5 6 7 8 9 10 11 12 16 UA._ Innition _Eg Cs _Ha _.1 ' _T.a Au tie Ce- U _Cd la _.Sn Gal 1 5%- 1.5 4.2 2.2 -- 4.6 3.5 2.0 3.0 4.0 3.9 1.8 4.4 3.9 6 100% 7.6 31; 12. --
- 25. 12. 6.5 10. 13. 13. 16, 16. 40, 9 5% 1.5 3.2 5.5 --
3.8 3.6 2.0 2.5 4.0 4.0 1.6 -7.5 3.4 9A* 5% 1.1 1.8 1.5 -- l '. 5 1.3 '1.2 1.2 1.3 1.3 1.1 'l.4 2.0 Cases 1 and 6 vere Wetwell Vall Breaks Cases 9 and 9A were Drywell Head Breaks
- 0xygen Ignition Limit Reduced from 5 to 4% at 19.4 hours 5 53
Table 5.1-34. High Pressure STSB Ignition Limit Sensitivity Results: for a Wetwell Wall Break with an Area of .000799 m2 e i.i ! Fission Product Releases from Containment l a- Masses Normalized _to Base Case a Case Hydrogen 1. 2 3 4 5 6 7 8 9 10 11 12 16 L Ignition A Q. .g .,J.a ._1 _Ig _B.9 112 _GA La U RA .CA1 Base 10% 1.0 1.0 -1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 i 4 5% .31 .21 .28 0. .10 .01 .22 .15 .11. .03 .17 6E 4 .15 i i i Fission Product Releases to Environment Masses Normalized to Base Case 3 Case Hydrogen 1 2 3 4. 5 6 7 8- 9 10 11 12 16 ; j FA Ignition _xg ._g.s _g.a ._1 _Ig. _Eu .E.2 _f.a _La U _ 0 1 S.n G LI .!
)
[- Base 10% -1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1.0 1. 0 l 0 - 4 5% .22 .30 .24 0. .13 .05 .31. ,17 .15 .07 .05 .002 .21 I i Reactor Building Decontamination Factors Case Hydrogen 1 2 3 4 5 6 7 .8 9 10 11 12 16 UL _ Ignition Xg _GE _B.a I _In .lu J Cg _La U _.C.d _S.n : 9.11
- Base 10% 1.8 4.2 3.4 1.9 4.7 8.9 3.3 3.1 3.6 7.9 2.4 -11. 4.3
!- 4 5% 2.5 2.9 4.1 -- 3.6 2.8 2.3 .2.7' 2.8 2.8' 7.9 3.0 3.1 l i i l 5-54 l h
<~.e
90 , , , , ,
, 7 LASAlik MElr0H HICH PRESSURE S110RT TERW STATION BLACKOUT , g ,g 8.0 - -
- 1.1 7.0 -
-~ ' '
Total b.li h ; Hydrogen a.0- - 0.9 l' 9 :' C . li
,k}
- 0.8 7.
e s.0 - ., Tc w o :" :g - 0.7 C ._
- .ss n E .
- .:l o 3 :' ..' c.
;l 4 0-e
-- 0.8 6.
ll: A :: : :
- 3. 0 -
- 0.s
. U
- ': - 0.4 e.0 - ;:: :
- -- 0.3 e... ,
- ': - 0.4 e.0 - ;:: :
r ,:'.:::: : ,
- 02 t.0 - ;;:: .
- f!j
- (. ,,,,,-
, y f . .. .... ,. --
~ - 0.1 0.0 ,-
, , -f' ,
aa 0.0 12.0 24.0 08.0 48.0 80.0 72.0 84.0 98.0 108.0 120 0 TIME (103) s Figure 5.1-1 Core Channel Pressures 5-55
- - _ _ . .. . - . . ~ . _. .. . .
16 0 3 i i r-- i i i i : LASALLE MElr0R 111011 PRESSURE S110RT TERM STATON DLACKOUT b30.0 1 13.5- g -
- 4) --- Upper Plenum - 500.0
- - - Downcomer iss- 6, ~
Core Channel
-X- Core B ass - 450.0
\ .
Lower enum los- -
\/N - 400.0 A._._.._.._._.__._._._.._._._.._j TAF 3 00-
~
< l -
350 o I . - . Jet Purno ... y ~l l Nozzle j ' I I -' 300 0 *~ p 7.5 - t_~ " 't g -- i o
- ij j:
- s 3, ,
' - 200.0 $ en- ; i ((j i,j '*;'"g,,1,.,; ,, ,
Ev i v i
^
- ly u
^
UAI y
.....;^' .
; 6
- 300.0 45- " -
I 150 0 l Downcomer 3.0 - f l - 200 0 1 l 1.5 - - l -- 60.0 l
' Iser licad . aa 00 , , . , , , ; . ,
4.0 2.0 40 8.0 8.0 10.0 12.0 14.0 16.0 18.0 20.0 TIME (103) s Figure 5.1-2 Reactor Vessel Water Levels 5-56
,,e..-, . .
,w 1
260.0 i i a i i e i i i LASA1.LE MELCOR l'ICl! PRESSURE 2262 Sl! ORT TERM STATION BLACKOUT r - ouu u I ,, 200.0-d- *
-- 450 0
- 400.0 175.0 - f' '
,,__ -3 .
j- - 350 0
? " ~ , t'M " ~
O_ 125.0 -
- S - 250.0 -A O Upper Plenum to Dome E
$ 100.0- '
-O- Dome to Downcomer -
S y A Separator Drain ~ 800* o l Jet Pump y, C***~ ~ o
--X-- Channel Inlet - 130 0 O ? -O--- Bypass Inlet ,
p 50 0 --V- Channel Outlet - AM- " w -+-- Bypass Outlet -- 200 0 3 - - - Safety / Relief Valve JX 5 a s.0 - .
- Vessel Breach -- --- -- 60.0
~
3 ,,-- ax .- , h ^ O.0 { m' 'A. v'^=r,'-, n'w % ' ^
-g ;, 'v -- 0.0 v 1 m v v
-2 5.0 - M - -son
-50.0 - ---100.0 n
fv v
-75,0
--150 0 i i ,
i e i i i i 0.0 2.0 4.0 8.0 B .0 10.0 12.0 14.0 18.0 18.0 20.0 TIME (103) s Figure 5.1-3 Reactor Vessel Integrateu Steam Flow Rates 5-57
14 i -i i i i i i T e LASALII MELLOR HIGH PRESSURE SHORT TERM STATION DLACKOUT 1.2 -
-G- Upper Plenum to Dome 26
-e- Dome to Downcomer i .0 - -tr- Separator Drain
! Jet Pump
--X- Channel inlet -20 Dypass Inlet , . - - _ _ _ _ _ _ - .
g 1e- Channel Outlet / , M -+-- Bypass Outlet ' n - - - Safety / Relief Valve -r 4 - 1.5 S. 0.6- - --- Vessel Dreach _ s T
- 1.0
> 0.4 - ,- _ n O
O ,_,
~
Ls.
))
0.2 - -- .- 0.5 0 - ** y z a .......- 0.0@@ Q@f d j
-- 0.0 vmN
~ 0'2 -
g o
.r O a
- - 0. 5
- 0.4- .
- - 1.0
- 0.6 , , , , , , , , ,
0,0 2.0 4.0 6.0 B.0 10.0 12.0 '4.0 16.0 18 0 20.0 TIME (10 3 ,) s Figure 5.1 4 Reactor Vessel Integrated Hydrogen Flow Rates 5-58
t80 , , , , , , , , .
-- - Dome Upp2r Plenum ,
i.ee - - - - Core Channels is
~
2 60
---- - Core Bypass , i Saturation X , i 1 A '- , .
-_ 22s g s'I
# I il I
1.38- I gg I
--- 2 00 lt t I N # i t-I I t 2 II \ /
h g 1.24- il .I .. i/g -- 1.75 e i i i i C d i i i I it i g n 2 i 1 / o 1I 1 g . 1.10 - , ; ., f ,\
\
j ,, ,j- l'.50 E. ., i- [d,h p'..-
, i 3
b g-- ,,,,. , ,:.... a ..
. ,, ~ u,
, j \ ,t 1 b,c, i . s; .
0 e2- i !1 i -_ 3_on
, i .d y...u._ ...
. ~ ~ ~ ,,,_ _ _ _._._
0.ea - f., ,7' 4 i -_ o,73 J t I .1 f..- 1 0.64- C ^ " M - X-~ ~ N -_ o 3o LASAllE MELCOR ~11GH PRESSURE SHORT T"RM STATION BLACKOUT x_ 0.40 , , , , , , ,
- - x, ,
0.0 2.0 4.0 8.0 8.0 - 10.0 12.0 14.0 16.0 18.0 20.0 TIME (103) s l Figure 5.1-5 Reactor Vessel Vapor Temperatures 5-59
__m ._ _ . . _ _ _ . . __ _ _ _ _ _ , _ . . _ _ . _ _ . _ _ . .__- . _.. - _ _ _ _ - . _ _ . _ s r, , .-- - .. . ,_ m _ . 7 7__ , - _ , . . -
---.7. ,
1.ASA1,1 E MElf0H lilCli PRES 3UHl: Sil0RT TERM STATION ll!.ACKOUT 14-2.0 1.3 - j ,ggjp3.xd, ggir" j-V-{~y -W
~
38~
~
x i . -+- N
.-- g,, p O .
38 a Q 11 - . c C to- ; ,;;: V/ - l e E ni
' f- e f se O \
Ei 0 o-I
$ i j
?> ; y'- c 10 gg 08 - _
S
-B- Separators d[/ -A- Upper Shroud 07- ' ' -+- Upper Shroud licad -
'03 J/ -X- Mid Core Shroud '
c --4) - Dottom IP Shrou . oeJ -V- Vessel-Inner cylinoer -
-+- Vessel-Outer Cylinder l
*~ -o4 04t 1 , , i- , , , , ~r , ,
OD 10 0 20 0 00 0 40 0 60.0 60 0 10.0 80 0 90 0 100.0 110 0 120 0 TIME (103 ,) Figure 5.1-6 Reactor Vessel Structure Temperatures l l 3.co
i j i
)
s.n -- , , r , , , . . . LASALLE Welt 0R 111G11 PRESEUlit' ! Ring 1 S!! ORT TERW STATION lit.ACKOUT ! 3~76 3 Ring 2 - j Ring 3 Ring 4 3 60 X Clad Fhliure - an. o
- 3 26 '
- . 9 Cladding 'hinperatures / l: # 1 n.o - At Top of Active Fbel l. lsi
-' 3 00' !
l 'l,I - k .' '; i f l 2.76 ' l l i g i i 33 . I .I . n a s I O l $, ! ./ '* - E b0 O l 1: I I I # ! g f a l 3.s_ , l 11 ( l l- j
/ I.
l
-- e en . n O
E "* -,; I
~
i ~f ' K l ,{ g' f i ,/ .1 2.00 E o 13- l j l g
/ I
-/- -{
F l
/
~s l
- 176 g g,3 g l- / l
.Y. y -- - -y-y---- L, t1-
/ , _
/
- t/ l -
i so I / /) / l 00-f' /'
/ j 't "
. /-
/ I I [ - tes-l 'l- . .- ,
,/ \. #
l /
/ /
; 1 g ',e ji
- t ' 1 00
// /' ! 1 ON ~'
l',/ { g ~. nys '
/ .
. If , Y '
~
I - 0.60 0.S ,- , , , , y , , r 0.0 3 fa 3.0 4.6 6.0 7.6 9.0 10 8 12.0 13 6 M0
. TIME (100) s I
l. l~ Figure 5.1-7 Cladding Failure for Cap Fission Produrt Releases i 5 61
.2,,.,.. . . . - . -. . . ,-,-- .- . . - . - . . . . - . ._w. ;_ ,,
2 6 ~- - T t i T-- - 7 r -- i i , LASALLE MElf0R 111011 PHESSURE Sil0HT TEHW STATION liLACKOUT 3 93 2.3 - - 3 60 Zircaloy Melting Temperature (2090 K) f, 32s i,_
-~-
Cell 106 -) _ a oo t 44- Cell 107
- O -- Cell 108 I -
[] ~* 2 ~6- Cell 100 1 / n 17- l Cell 110 ) ' o --X- Cell 111 - 2.60 0 8
-O- Cell 112 Cell 113 c
3 1.5 - _ ) ,,
- - a co n R J l t)/ i #
c li, / t 00 h E t.3 - - l' [ i
#2 l< i i ,' i .,,
' / i ,
.< '5 $
t.t - : j i,) 4 , j I i -- s t,0 4: / i I
- i j [3
.- i I /
es - 1.26 De- ..
?
' si
\' -
j i /, , , it
/ ,
J gi 1 00
?
# p ON~ IO 'q ~ g _.,s f< .
g , 7-) 594- -_= - -y;y g,: . 9'.< - 0.60 45 i T i e i i 1 i , t-- 0.0 1.0 2.0 30 4.0 6.0 e,0 7.0 8.0 W.0 10.0 TIME (103) s Figure 5,1-8 Inner Ring Fuel Cladding Temperatures 5-62 o
.m
s6 r , ,
-r- , ,- 1 1--- ,
LASALLE WE1 LOR 111011 PRESSURE S110RT TERW STATION DLACKOUT b6
""~ ~
. . R9?. f?l19.10. . . . . . . .. .. . .. . . . . .... ...
. . Z. .R.02. . .(29. 90. .K.). . ... . .... .. .... .. . . . 60 g.
t.o - ~ l i. j 46 e.8 - -
- 4.0 2 g a ..
i j? - S ZR (?O98 K)
- at e.0 - C y - s.0 n
$ SS (1*100 K)
E .7 - . .. . .
- j. . . .. ..... .. . . .. . .It..._
4 y e6
~
.--O- [hc} EO
-O- Cladding
] -6~ Canister tt-
) Q'
-+- Canister (Blade) $' - . i.6 X -X- Steel -
r -O- Debris l '/ s o3, -E- Local Fluid 8
/ -
Melting # i,, 3o
, a m.
o 1 4K 0 5t g ., 0.0 1.6
"[30 i , , , , ,
-. ./ s 05 4 .6 6.0 7.6 0.0 10.5 12 0 .b. 15 0 TIME (103) 3 Figure 5.1-9 Core Cell 109 Temperatures 5-63
35 0- -
, , , - i- ,- i- ,- t i ww LASALIE n:P!IOR }{lCil l'Hf:SSUHl:
SHORT TEHW STATION DLACKOUT - 3.26 - 31.6 -
*.. ... .. - 8.00
,.
- j en o- ,
! - e35 g ,..' ;
. e.no c4 s- ,->
i - e ts
- 52 e s .o - .
9 ' f - e 00 0
* - - - SS Mass : - i n -- - Zr Mass E o
e st o- U02 Mass : -" 1#6 3 a r. ;- SS 0xide Mass i 2 E i --- ZrO2 Mass : - 2.so 8. Canister Temp
- 34 o. j ./\
.i E
- g l l\
l
- 1.25
. I 10.6- ,- ,l, *
-- 1 00
.1 :
,! ots 7.o - 1,1\ - -
. .) *
\ ,
\ /./.y --- - - - .' .' _ / a ,,.----~~----...,,--.,'f-
+ ' - s -
- o.so
~~ .
l: 33- t M f fI\ I J p. - O.t $
- ,\ ___,<.-- "
, tus , a aa '
oo- ! , h , l ; , , , , i ; 5.0 8. 0 70 80 90 10.0 11.0 12 0 13 0 14.0 16.0 TIME (103) s Figure 5.1-10 Core Cell 109 Masses 5 -
. . _ _ . . -. . . . . . - - ~ . . - - . _ ~ . - ~ . . . . _ _ . _ - _ - - . _ . - - . . . . .
I l l 4 l l l l 36- , , ,- , , , , , , LASALLE Elroll!!!Cil PaESSURE S110Rf TERW STATION ULACKOUT 6.6 3.2 - -
. P.07. f?.110.' ' . . .. ... ... .. . . . . ... ;
3!i.92,(299hy,, , , ,, ,,,,,, ,
,,,,,,0,,j ._. _ _.. _ g . .
,,, ,, , , 3, 64 ;
i I ri- -tr- CR Guide 7bbe ' ' '
---X- Core Plate
--+- Cell 100 Steel -
e' H I 3 _ 4 ~3 t ae- -+- Cell 100 Debris '
' f ._
-G--- Cell 107 Steel 12 Cell 107 Debris
-G- Cell 108 Steel
,' U 40 n 88- - O - Cell 108 Debris ,0 .
o - - - - Melting i b IU' - 35
. . ' J_~ E^ tb'
, U(
g_ ZH (2008 K) C y so r~ ; ,4 s a n o, A
~
SS (1700 K) f> , .I
/ !
E 1.7 - - - - - l
# /g
- 26
( h / r
/il- '
!.4 ~ 3 '
gg
, /
f I (i) i i i c
; / I /
['.] # 1.1 - ', I i
-- 1.6
, I l 1 , [. )
}' ! -f- 4/ [] }
ny. 0)/'y~
' Fl Il 10
) d If 3 1
AL grn 0.6 '
, , *N t
", E 5 0'O 0.0 1.5 30 45 6.0 7.6 9.0 1OA 12.0 13 6 16 0 TIME (103 ,)
i i l l Figure 5.1 11 Lower Core Teinperatures of Inner Ring 5 65
, .. . = , - , . - .. -. . . - . - . . . - ..:
A 'l a t i 1 16 0 , i t i i . i e i _. - SS Mass usAux WEtron Hicli Puissunt
- . . _ . _ _ y, Mass SHORT TEHW STATION DUCK 0UT
-~ 8 0' , t s.6 - UO2 Mass
- - SS 0xide Mass 7
- 02 Mass - av o k 12.0 -
e------ g - 26 0 i i i 10 6- r-*----*----"** - , g I - 22.6 I
- b. ; . . _
! 00- I g
-" 20 0 1 9 i k
< M , I g i 17 6 i n0 ,
- N
- 7.6 - 1 .
o 1 w I e m I w t- i 15,0 e a n , I I 3 i ; 3 - -J 6.0 -
, ,, s I
I 4.6 - - - 10.0 I I I - 7.6 j , ;
- I 3.0 -
I ! 1 - 6.0
-..-.. . . . e
- g - i t .6 -
2.6 .' r.-.-.-.~.-.-.-..-., j O.0 - l , l, ~ ~ ~ - 1 , , 3 i *^--(=*~*-- i aa 0.0 20- 4.0 8.0 8.0 10 0 12.0 . 14.0 18.0 16.0 20.0 l TIME (100) m a l i 1 1
-l 1
11 l_ Figure 5,1 12 Core. Cell-106 Hasses 1 5-66
.-..- .u_. .
u_._. = - - :-. = u- . ,. -.- - a : . - .- . . =_- -=_-_. .; _ --
2 00 , , , , , , , , i 1.ASALII MELColt illCH I'RESSURE Ring 1 - 1 00 SiloitT TI:RM STATION DLACKOUT 1.e6- - Ring 2 - Ring 3 Ring 4 - a 16 1.to-X bilures -
- 2 60
\
t 1 66- \ -
\ 2 26 1%
2 [. I \ 140- : s - s e 00 c - i s _l'a 3 K .
. .)(' -:
< s R
1.26-
~ ~ .; - i. 7. y . g . C 3 ;, s ,, ; i .-
2 p -
- /
fs 1/
-s,N - _ 3 ,,, n S
/. % s
- n. A l,; .
6 h110- ,f,/. / { / ~- 1.60
.f * '.. . -!
- ./
0 96- ,1 ,/ l
.e *c -- 126
's :/
/:
0.00- i
'; .- 1.00 t ,I ;
'/ :
/ l 0.86- '
; -- 076 o .
_ - J ' ,. -
~
0.50 , , , , , , , 30 0.0 6.0 9.0 12.0 16.0 18 0 21 0 24.0 27.0 30.0 TIME (103) s
' Figure 5.1 13 Core Plate Temperatures 5 67
_ _ . . . . _ . . _ . . _ . _ _ _ _ . _ _ . _ . . _ _ . _ _ _ - - - -.. _ _._ _ _ _ . _ - _ . _ .. .. .._ ~ _.. _ a6 , , , , - , , , , , LASALLE WElf0R 111G11 PHESSURE , ,,,, Cc!) 101 Debris ~ Sii0RT TERM STATION DLACKOUT
- - - Cell 102 Debris ~
88' -+- Cell 103 Debris
- - - lower liend Penetration. - as 2., - i- Vapor .
- 4.b 26- t
- g
?
,# -- 4.0 l ' t 2 3
,- g-p 33" s r o - ' - 5.6 C '\j , _s' '
/.+*,..l- .
{
...- I Q *
- s 2.0-s g . .* g-
**..- 3.0 O I B O
~ ~ - ~ ~} , ,
N y,k.-s . . , , p*U ,' ' '
,- _'l s/ \s g 3,.y . -..
l i g- ,"
,# .. 26
_ 'l 1- l
'g ;
1*4 - " .
~
/ l -
80 , p' l 1 e
, e s
/ ' l !
t ,1 - -/ ! *- 1.6 +
'{
l < l
.\ l o.e - 1 gg - -" i0 o.6 , , , , , , ,
1- -! - c.6 : i i 12.0 12.8 .13.8 14,4 16.2 16.0 15,8 17.0 18.4 19.2 20.0 TIME (100) a L T Figure 5.1-14 Inner Lower Head Temperatures 5-68 ' t 1
, _ - . '.-..U.,hi,,me.-_,'-,
, s m i c ,. .h.., --o,.J.. -..d~r' ,
..1,. r, ,, . - . - . . , . + 4.,..,-...5 , .,..,,.,,,...b.-m.m
. . . . _ . . - - . . . . . . - . _ . _ . - ~ - . . - - . _ . . - . . .- ~ ..- -.. - - -- - -. .-
t i l
'f0 0 - , i i , , , , , , i LARALLE Wl?lEOR !!!Gil PRESSURE !
05 0- Sil0RT 1ERW STATION BLACKOUT - i'
- 140.0 I
00 0- - SS Mass os o. Zr Mass .- see.o UO2 Mass
~+- SS 0xide Mass i so o- ---
Zr02 Mass -
-i 1
460- -- 100 0
) 40.o- .
1 a
% ,s o.
- e. / -
- 60.o n
- 5 l f 30.o-60 0 ;
26.0- . t P 20.0- -
- 40.0 16 0- .
so o- I
\. -- 20.0 I
go. - i - __
, _ , s_ ,________,1,
\ t o0 l , l, , !, 7 4 , , l , * ,l aa-10,0 11.0 12.0 13 0 14.0 16.0 16.0 !?,0 18.0 19.0 - 20 0 i 3
TIME (10 a) r I l l l. Figure 5.1 15 Lower Pletuun Cell 101 Masses l l 5 69 i. _ - .. -.a,- . , . _ . . . - . . - . . _ .. . .
_ ._ _ . - . . ___.__m . - . . - . _ _ . . _ _ _ _ _ _ _ 2 6 -- , t , t , , , , . Ring 1 t.AsAt.tx uttron lucil rnessuar sliont tr.av !TTATION DLACKOUT Ring 2 - 836 23- - -- Ring 3
- - Ring 4 - 8 '"
X thilures et- '
- 8tB
~
, . - * ...., 3.00 go. .
! ' ..,* . . - 2.75 I 2
t, Melting (1700 K) l' _,_._n_,._m--.u--- .'"*". . I 1 m ,. -
- 2.60
. o .. , _~ ~~_' 1 O l s g 8 : a> 1.5 - * / .- E tB n
. g l 's, O
-
- v E .s'~~.---
g ; I l . g on g 3,3_ lhilure (1273 K) : 1- _ x - - _. 3 __ g __ __ _ _ _ _ _ __ _ __ ;k' g _ _ _ _ _ . - l
- 1.76
, l c
. t l
1.1 - : , -. 1 60 f I l
- i l - i.es
- o. : , ,. - .
I a s'p'/* - 1.00 03 .. ! i ./'
/ _- 016
,_.a.,---.. .
l
/
~ '
0.5 , , , , , , , , , 10.0 12.0 14.0 16.0 18 0- 20.0 22.0 24.0 28.0 28.0 30.0 TIME (103) s Figure 5.1 16 Lower _ Head Penetration Teinperatures t F 5 70. 4
70.0 7 ' ' i_ s i i LASALLE WELCOR 111 0 11 PRESSURE 65.o- S110RT TERW STATION DLACKOUT -. 140.0 a, jg co.o- - ! A Axial Cells 1 l 33 o- X Axlal Cells 2 3,g , O Axial Cells 3 Radial Cells 1 60 0- . Hadial Cells 2 - A j' -
- Radial Cells 3 45.o- -- Radial Cells 4 -- 200 0 ) 40.0-a 7
n o - 80.0 d os.0 - ).\ - n E
)N j 30.o- -
X - co O 25.0-20.0- # \
- 40 0 i s.0 - -h-6} ). s .
,h -6 10 O' i
-gi co o
,,_ I,.
Fp owf.l * " i , z px===6,==r g-' O.0 m r 10.0 12.6
- 49t>
16.0 17.6 dhv 20.0
... .4 fa->d6M)>r-/frOM>O>-6E>%XO 22.6 25,0 27.6 30.0 32.5 h35.0aa TIME (103) s l
l Figure 5,1 17 Lower Plenum Fuel Masses 5-71
1-l 00 o- - 3 ; , -, , , , 7 ,
,._,___,__ 1.ASAL1.E WE120H litCil PHE$SURE :
- SilORT TERM STATION RLACKOUT a -,
ec o- 's - s i n
- 1.2 70 0- -
' ~. g - i ... . ... . .
l Q ' x so o - ..g** - -
-- i.o }
o % i n ;l o
~-
3 . l t
- . E
$" 60 0- ,' I
=
- o.a j y ,. - _ _ - _ _ _ . _ _ _
_ -g w : u o 40.o - ,.
. Zr
~
8-I5 I [ - - Zr02 - ce R
.. - - - Steel . . ..
a / --- Steel Oxide
# 30.0-
- - - - llydrogen g ...
** - 0.4 l 200- l .
./
, - " *{ .l
.i *- .'.- ,
W 10 0 = .
,#. ## % \.m. -
~. 0.2
. ,/
s' \L.y
),__.._.._...-.._..-.c l
oo i j i j i i e
~ t i
i
.. .-.~.
i
- 00 3.5 7.0 10.6 14.0 17.6 21.0 - 24.6 - 28.0 31.5 35 0
- TIME (103) s l
Figure 5.1-18 Total Core Masses 5-72
1 I
-3.00 , , , , , , , , ; =,=-4 -
/ '
- LASALLE WElf0R 111011 PRESSURE f l SHORT TERM STATION DLACKOUT / 8- l' t.*rs . . ,' , _ , .4 -
4s Volume - Channel j /
--- 1,ocal - Cell 113 -/
N- -
--- local - Cell 213 Local - Cell 313 ' // .
i I 4o i - --- local - Cell 413 l /- ! ers- I l- - !+ - sS li.l' 2 1 / " 2 00-1'l lj 1
+ n s l 1 o , ,
- 3.o ~
f - I, t 4 g A \ r'gg ' ~.'/ \,f li I
'i I g IM- it . e.
- g. s Q
j 8 lI 's -
- e.s no .)
M1l-
- o- si il s / s .
s, j i Eiso-
# 'f I / /3
). 1. / ,i
/3 i j
- l. s ,.
- 20 i
r i.es . s I Ni i,, .. i - - q l: I }. , k l ,[ ' ' j
/ .
1.6 - l - s.co- - ,'/ -
,/ , ,,... ,
; 'f - , j-
,/ ,1. I...',..' V
- 1.0
! 0.75-ji ,,..... ! ' a ! c oo , , , , , , , , , i 0.0 1.6 3.0 4.6 6.0 7A 9.0 - 10.6- 12.0: . t34 16 0 ! TIME (103) s T t 1 i- , i i 1 . l ; Figure 5.1-19 Core Exit Vapor. Temperatures l t I-I~ !. 5 73 i i (- n..w,,s.,-+ . ,b, ...,v .r., wn.~,a-- --- .,,w, .n.,r-n_,. a n . ---nm._,n,,,,,..n.nn.- cw,e.-,, , ,-w++~,-- + . e .-
~ .- . . . - . . . . . _ - - - - - . - . - . . . . . . - . _ .. . . . . . . . . . . _ - . --
i i 1 1 J i
$$ 1 i s e i , q I I 1.4 -
Drywell i
! DW Steam _- 800 0 1 -
Wetwell l 1.3 - . .+.- WW Steam ~ 1.2 - - 176 0 4 3 1.1 - -
\
- - 160.0 '
. 1.0 - - l l n i k 09- -
- 126.0 i to n l O 0.6 - - .N a
v g E - , :s 0.7- -
-. 100.0 -
rn j -
~ 0.e -
a Ob-f ~
~
0^0 1. ! 0.4 - % -
!, 60.0 J 0.3 - .,
! + , ; . 0,.
,_ f.4:'.4.',...+.. ........ + . ..........+ ....
0.1 - ) l+ LASAllE WELCOR HICll PRESSURE
.. S110RT TERM STATION BLACKOUT 0.0 "l M12.0 0.0 24.0 38.0 48.0 60.0 72.0 aa B4.0 96.0 108 0 120 0 i-TIME (100) s 4
?- .I i i I Figure 5.1-20 Containment Pressures 5 74
.. . _ . . _ . _ _ . _ _ _ . . _ _ _ _ . . _ . - _ _ ___ _ _ _ _ _ ._ _ _ ~ . _ ._ - _ _ . . _ . _ . _ . . . . . _ _ . _ . . . . - ~. . - _ . _ _ ,
1 l 600 0 , , 7 -- - , , , r , , ! l LASALLE1AELCOR 111011 l'RESSURE S}l0RT TEHW !TIATION 11LACKOUT 6'r0.0 - - 650 0 iT 640.0 - l
- 600.0 bl0?J - ~
,g 4500 l
2 4800- - 400 0 b $* - i
/
3
/ s *'. ,*' e < .g-
- c.
- 460.0~ f -
Osop > 8
/ - - i R.
E *
' # p~ r g 420 0- -- soo o N& l..-
Atmosphere i 3 00 J --- Liner 250 0
/ - - - - -
Floor
' +--
Reactor Shield 1 " - - Misc. Steel - - co0.0 s oo--o- ,; - e r
..- t oo.o ;
330.0-pJ f 100 0 300.0 , , , , , , , , , 00 12.0 24 0 36 0 48.0 -60.0 72.0 64.0 - 98.0 105.0 $20 0 0 TIME (10 s) F Figure 5.1 21 Drywell Temperatures 5-75
. ~ - . . . . - . . . . . - . . - . . - , - - . .- - .
. . ~ - - . - . . ~ . .- - . . ~ - - . - - - . . . . . _ - . . _ . . _ - _ ~ . . . - . - - - . - - - . . - - ~ . . -
800 0 m- 1 i m i ,. , i i LA$A1.LE WEtrOlt 111011 PitESEUllE S!! ORT TEltW STATION llLACKOUT 670.0 - - 660.0 Atmosphere Suppression Pool 640 0- . - - Wall 600.0-
+-
Misc. Steel 610 0-
-- 460 0 2 4B0 0- -% 400.0 }
v -6 8 3 450.0 - *. -- 260.0 C- I E ,/ - o , s. b. g l
-- a00.0
\
g 420 0- < 4 f I 390.0-
.a f- *\ [ ' N , * * * ~ . , ~ . - . . , - + - * " * ~ '
250.0 f f
./*
-- 200 0 300.0 ~ q ,!' l*
**~~__. + ~~~-"*~~~ -
1 .t.
- 4. / .
, # j ' *j' - 160.0 - l 330.0 ~ / ,1 - -
f L. ' */
- IM O 300.0 ,
t-- , , , - - - , , ,
- " = - -
0,0 18.0 24 0 '36.0 48.0 CD 0 78.0 '04 0 96.0- 108.0 ' 120.0 TIME (100m ). , i f-4 i Figure S.1 22 Metwell Temperatures l L 5 . '76' . l l ll l h
. . _ . . . . _ _ _ - . . . . - - . . - , . . . .2......-. ..._.,,..._2.,:_., ._. _. ;.-,,_.,,,,,. . . _ . . .
7 I I T I E I I D'
-X- DW Pedestal Atmosphere 80-
--4"- WW Pedestal Atmospherc ~
-X - DW Ped Wall Surface
-+- WW Pod Wall Surface a00 WW Ped Wall Interior (.11 M) te- --- WW Ped Wall Interior (.30 M) .
j -' W A % g.% ~ y.-_ f r- e so 1o- . . - - e ns R f, 14- ~' no . e co
=' ~~,
C 0 {.
- s 1.2 - / -- 1.76. mo l / O o ' /
1 50 a : < g t.o- : -,' - p j j - 1,26
- o 0.e - F - -
- 200 n e.
+ .
s
, s M ...M o vo
) . . . ye . .-,
o go
- $)A![
04- ,
'~ , / --- o gh m'fr' 4 Q 1 *g ~_ _, _- LASALLE Metr 0H 111C11 I'Rf'SSURE Sil0RT TERW BTATION DLACK0t'T 0.00 02 ,
i . , , , , , , 0.0 . 12.0 24 0 36 0 46 0 60.0 72.0 84.0 90.0- 108 0 120 0 TIME (103) s 3 Figure 5.1-23 Podestal Temperatures 5 77-
. = _ _ .._.__..__._.__._._-...._..-_.-__....___.~..,,______.-..__o.._. ._.__m . . . _ _
7 g o ~_ 4.)_.,p , ;- y .- ,., _ ,- ; -- , 7 .- , , , i! ll LB- d f ~~~~~ ~ ~"~"" ~ "' ~ ~* ~ '" ~ *' - - 35 0 jg e
'( /
- 1. <
16- It / -
/
p if / - 30.0
/
i.4 - 16 / - f L.ASALLE MEIColt HIGsf Pf0MSURE k Ir! l; n; i
/
S110RT TEltW BTATION til,ACKout
- 26 0 sa- .. --
li; i 9 li; 8 x n ji: -o
'o 1..e - ) l, 20.0
$. l li:l'/ $
5 ji C 0 n. j ill - - - Drywell Liner Flux A g ; i .s Drywell Flcor Flux is o e WW Wall Flux A II N/ --- Total DW Condensate . 0.6 - 1 ,
'l i .
j' i i) ** - 10.0 . 04- I)', ;;4 - S it ', . l.I I su s
., - s.o 0.2 -
s{' g 4 ,a A l" ' 0.0 '"f1 i (, , 3 ., aa 0.0 12.0 24.0 36.0 46,0 6 010 72.0 84.0 96.0 400.0 120.0 i TIME (100) n 1 4 Figure 5.1 24 Containment Condensation l 5 78
- . . - . . . - - _ - - . . . . . . - , - . - . . . - . - _ . - . - . . . . - .:- - ~ . -., .-. .
=. . ~ . - - . - . - _ - - . - _ _ . --- .-. -- - _- - - . _ - . . . -.
l e to o- , , , , , , , , ,
; 1.ASALI.E WE!EDH ll!Gli P!WSSURE
;;, SH0d7 TERW STATION llLACKOUT - 32 6
- 1 se o- l l, -
I f. 30.0
- l ',
- er.n 12.0- l ; -
I - es e i i Drywell Pedestal l : - Wetwell Pedestal so.a . i- :- - - - Drywell c sa a j-l 3 i - ro.o 5
- :3 n .
9 a o-
- i
. St.n "o c
R < j j 16.o 60- - - l 12.6 l
; - 10.0 40- -
, ; - 7.6 e t : - s.o 20- * - I . J t a . ; - en 1 l l 0o . . , ,--. - r- d' C .I,W ,
,.j..____ , , .
0.0 8.o 16.0 24 0 38.o 40 o 48 0 66.0 64.0 72 0 '80.o 0 l TIME (10 s) l (~ l Figure 5.125 Drywell and Pedestal Water Masses 5 79 _ .. .. = . . _ _ ... - _ . - - _ . . __ . ~ . , _ ,
- - - - - - . ~ . . - . - - . . . - - - - . ~ _ . - - . ~ - ~ . . . . -. . . ... .x -. - - .. - .. - .. . - . - - -
400 0 , , , , i--T i i i www w LASALLt Wl "OH 111 0 11 l'HESSURE bil0RT TEk
- FAT 10N DLAEDUT S60.0- -
390 0 320.0- - 300.0 280 0- --- 3*t0.0 240 0-
.... *****" .... ,,....- -- 360 0 2 9 e -
, M ...
.- K n ,-
3 2000-350 0 3 v 6 E b. c a b $
' 160.0- -- 340.0 120.0- -- 330.0-80.0- : Mass Increase -- 320.0 t
- Temperature - ,
40.0- -- 310.0 Initial Water Mass of 3.062.400 Kg 00 , , , , , , , , ,
-aa a 0.0 12.0 24.0 36 0 48.0 60.0 72.0 84.0 96.0 108.0 120.0 3
TIME (10 s) Figure 5.1 26' Suppression Pool Mass and. Temperature 5 80
. . - . _ _ - ._._ . _ . _ . _ . . _ _ .. ._ . _a .. a _ _ - - - _ . _._._.._.-a_
I
-16 0 , , , , , , , , i LASALLE WELCOR 111011 PRES 3URE '
Sil0RT TERW STATION BLACKOUT
-i s.7 - __-__--__----------------- - - - - - - -
- 706 0
- t e.4 - -
g4
,- , ___..wm -- - m 702 6
-17.t - ,. .
\kj -
\ - 700 0 V !
\
-t7 8- - 3 R cs7.6 .@ '
v ::2 g -18.5- g ;
> Downcomers o
.3 - - Wetwell cas o M a
-tem- @ .
s:
- 692.$
- 19.9 - -
- 090.0
-na - -
--u -----.-- --- --
Downcomer Exit
-ei.a _- e a7 6
-22.0 , , , , , , , , , , M60 0_0 12.0 24.0 36.0 48.0 60.0 72.0 64.0 96 0 108.0 120.0 TIME (103) s f-4 1 .
Figure 5.14,7 Suppression Pool Water Levels 5-81 _ _.- - --.- - -.-- -- _ ----- --..- ...-.-- _.- -..--.-.._- ..-_;--.. - .=..-.-
a0 , r--- ,
-- t . . 1 i
- vu LASALLE MELCOR lilC11 PHESSURE _t -- 1.ight Oxide Mass sitout TERM STAT 10N 11LACKOUT -X- Mel.nl Mass t.e - -+- Light Oxide '!bmperature -
Metal Temperature ' 76
- -X-j 6.4 - ,'.'.
N.,'. X X '
.- 1.60 i
.+..
- 1 68- ',, X
- 1 25 . Q' 48-Q -
. O x X n L M '[ . M f g 4.o- -- 1 00 y j e . . .
E
'. e y
4. 3.2 - .'. f i )', .. - 0.75 g , t 4;. X, 2.4 - . .
..,. b' .' X. . , . . . . . ' ' x . X
\
'" +- '; , +-X..-.
+
., , . . . .X . - .... ,
~ 0 50 j to- -
t--** -
' <[.
s* -O I - 0.25 08- i,,'., 0.0 --tY(4 X1 1 d T i i e i i 14 50 14 85 14 60 14.95 15.10 15.25 15 40 15.55 1570 16.85- 15.00 TIME (103) s A i 9 i i ! Figure 5.1-28 Drywell Pedestal Debris Masses and Temperatures i 5 82 l l..
l I l 1.0 , , -, , , , , , , 1.ASALLE WI:120R lilCli PRESSURE
'Iblal Sf! ORT TERM STATION 11 LACK 0UT i 0.0 - --- !! cavy Oxide - 2.00 l IJght Oxide .
Metal O.B - *- 75 \ 1 o.t -
/[. .. -
- t so q O s .- .. ...**.... --
x ,.... *,,. ses 7 e *.... - g 9 c.o- , e c) a
... . 3.00 C
, x 0.4 - l ,
- - o.vo c.3 -
b 0.e - - r' I i 1 o.so r j j i ot- -
--j -- o.es t
.L--------------------
4 .I 0.0 r-- ,- , , , , , , , n an i ' O0 12.0 24.0 Se o. 48.0 60.0 72.0 84.0 . 98.0 108.0 120.0 [ TIME (103) s l I-1 Figure 5.1-29 Wetwell Pedestal Debris Masses 1 5 83 l _ _ , _ ._. ..
. . . . - . ..--.---,-- .-.. .~.~. - -
-.- - _ . _ . - _ . = . . - _ . _ - . . . .
2 26- , , r i i i i i LASALt.E WElro". li!GH PHLSSURE SilORT TERM STATION DI.ACKOUT 2.16- 34 (' l 1 t os- g 't 3, ; I. \, i I \ 1.98h - : l I,
- I t no 2 ! Yf i ;
n ins- : 4 o - 4 L to e : ti: pt ! i:. 'li 4 E,
}gis.....--.....,,,,,,<......."--...
,,,3 ;,;.. _
- \ ', C U
- .' n , ,,u ee no
*s ,
D. ., I \ ,--~~= p -- e 1.85- . :: , , / n.h-,..---. - i@. : 3, e I'ef 'm /
- , I
-e. '
- t t so- ;;;
l {I ll s
- ' 'i
- , : :' I .h jla i
es
' !! )
i.< s - g g, j If k
\
il Heavy oxide i ji a.o 3.ss - :.; ,ll g l ,l
--- Metal Light Oxide -
gi ql l ,j Iower Pedestal Atmosphere es ! , 'E,' 4, "
-1a 0.0 12.0 24 0 36.0 48.0 80.0 72.0 64.0 96.0 108.0 120.0 TIME (103,) ,
Figure 5,130 Wetwell Pedestal Debris Ternperaturea
- 84
. . _ . , _ _ _ _ _ . . _ . . _. . , ~ . . _ _ . _ . . _ _ _ _ . ._ _ . _ . _ . ..
. . . . . _ .-....... ~ . ~ . - - . _ - . - . - . . . . _ . ~ . - . . . - . . . . - . ~ - _ - _ . ~ - _ _ _ , . . . . ~ . - _ . . . .
l 2 i I
-i
*160 , , , , , , , , ,
- LASALLE MElroR HIGH PRESSURE
-16 4 - S110RT TERM STATION tiLACKOUT '
.- 706'0
- - - . .- __ _ . - _ _ __ =- Port _13ot_ tom Elev_at_i_o. n_ _ . _ _ .
-teo-
/\ -
-]. .
-t 8.6 - ,
[ -
- I
- j. .... . .......................................................... ..
- 702.0
- 17.0 -
m ... ..."l 1 f._ 0 c _ .lnitial kwerNestal Boor - - - - Eb% - - - - - - ~ _ : 100 o a
.8 -17 4} - -- w : _ -. \
O
- m g g ti B
\
Melt. Surface
~ 8 8' - t ~-
Ibol Surface .. esa o ~U
\ ---
Mcit. Bottom -e
- s. ""*" Suppression Pool s . 5: .
-t e.s _
v - s - epe.o s '
.s -
- 19.0 -
,'~~,~~~_' - - - .te4 0 '
- 19.6 - '
Upper Last Lower Pedestal Pedestat Ejection Rupture Thllure - co2.0
-20.0 - T i i i i e i i OO 12.0 24.0 15.0 - .48.0 60.0 72.0 84.0 96.0 10'8 0- - 120 0 Time (103 ' gec) .
l'
-Figure 5.1 d1 Wetwell-Pedestal Mel't Elevations..
, 5 85 L m/'P-m%-.wtwemtr mm'we w!* wTt**W'B'F3'" f' '# CMT 'W F ^'Y#9' '-Y 8A' "NDT" *""'T '*'*'Y **8"' '9 # I' E YY5 'I T
- . _ _ . . . _ - . .- ~ . . . . . . . . . . . ~ . . - ~ ~ - . . . _ - - . . - . ~ - . _ . ~ - ~ . . ~ . - ~ - - . . . - _ ~ ~ . . -
20 , , , , , , i r* 1 LASALLE MElf0R HICil PRESSURE SHORT TERM STATION llLACKOUT 1.8 - - ---60 e 0
'"0
,- - 6.6 1.0 - ,
Pedestrl bll .Thickness - 6.0 - !
. . .. . . . . . . . . . . . . . . . . . . . . , .. ... ......... . .... ... l
- i n- ' l
- 4.6 S
w i
/ -- 4.0 8 12- i a
,% I u
.!G / - 3.6 ,
O l.0 - / c
.2 - 3.0
~' I 2 I
- 0.6 - -
I - t.6 N S 8 Radial i - - - Vertical . ... e.0 o s. ' o i J
- 1.5 0.4 - y.
I - to . I 02- I -
, - 0.5 - .
s 00 , , , , , aa 0.0 12.0 24.0 38 0 ' 48.0 60.0 72.0 84.0 96.0 108.0 120.0 TIME (10 3 s)-
}
Figure 5.1-32 Maxirnwn Concrete Penetration Distances l l l
-5 86
_. - - - _ _ . = .: = .~u.., ..- ._ - - -- ... - . . . . . . _ _ . . , _ . .
4.0 - 3.$ - 30-2.5 - y 2.0-g , ., . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . g - _ _ _ - \ ,. L 1.0 - 8 '?,
- C, 0.6 -
' /,. .b - 0 .0 1 * ,
a -0.s - / / /-/' uy
-.............-.........................s- ,
l . i
'f ,
~1.0- *** /
0 ...................................................***' e i's . _/ , e _ _ _- . -- _/_
", -2 0-o c:
e -2.$- Y! IIGEND A -3 0-Initial 3 s. E..c..o...n..t..a...i.n...m.. .e...n...t..Er..e. .s..s...u...r..e.....
.. 1.0. 0....P.s..i
..160. .P.s..E i .c.o. .n.t.a.i.n.m.
. . e.n.t. P.r.e.s. .s.u.r.e. . .
-4.0 - Containment fai, lure,(195 ns,ig)_
Pedestal wall failure
~4 8 ~
End of- calculat_ ion _
- 5.0 , , , , , , , , , ,
0.0 0.5 1.0 1.5 2.0 2.5 3.0 3.5 4.0 4.5 6.0 Radial Distance from Pe-'estal Midline (meters) a Figure 5.1-33 Axisymmetric Cavity Shape 5-87
. - . . . . .. ~ . . . . . - _ , -. . . . .
100 0 - ~ , , , , , , ., , i-LASALLE WEl.COR 1t*011 PRESSURE SilORT TERW STATION Bl.ACKOUT 00.0- -- 200.0 lbtal
-- Carbon Monoxide e0 0 --- Carbon Dioxide -, 37y,
- - - liydrogen Steam 70.0- -
150.0-80 0- - g
, - 125 0 y
% p &
v 3 60.0- e M F) #. O M e - 100.0 0 C / 40.0- ,/ -
/
/
/ 75.0
/
30.0 ~ s -
* /
/
e s ... ** - 50.0 20.0- p
/
z ...... 10.0-
""~ - 25 0
/- ,
,s
~<
="",s*
,f 4
10 i
," ,~~. , , , , , ,
8 aa 0.0 12.0 24.0 36.0 48.0 60.0 72.0 84.0 96.0 108.0 120.0
' TIMF: (103) s l-l l' Fir're 5.1-34 Core / Concrete Interaction Gas Production i
5-88
. , _ . . ._. . . . . . m . _.. . _ . - _ . . . _ . _. -
12 i e i i , i i . i
- 2 60 1.1 - [ -
~ ' ~
1.0 - ....-
- O_g - ,.. **..... , -- 2 GQ .
0.8 -
,, ** ..*,.. *.. ~ - 3,75
*Q o.7 -
M / .1.50 m$ to l 0' 30.6- ! - n
!g- In-Vessel - t25 2-l --
Ex-Vessel li 0.5 - , l . .
;. - 1.00 0.4 - ,'
.' - 0.75 02- ; -
; - 016 0.
02- , O1_ - 0.25-
, LASALLE MELCOR HIGli PRESSURE
, S!! ORT TERM STATION BLACKOUT 0.0 ) ,
, , , , , , , , , a a" 0,0 12.0 24.0 - 38.0 48.0 80.0 72.0. 84.0 98.0 108.0 120e TIME (103) s Figure 5.1-35 Hydrogen Production .;.
5 . - . . . . ,
60 0*- , , , , , , , , , . vu
; 'Iotal Mass -
; - - - Steam Mass '
64 0 - ! - Carbon Dioxide Mars "6 j ..... Total Rate-a e 48.0-f :
* - 1.60 13; 42 0- ; - ^ .Q g ... .
x 'g :
. .... . _ i .2 3 - ee oose o- ,'a g
,s
,h
.*+.*. .,s-. k.
m m
;,,s;' ;. ., ,
,- a o
;,s.. !, ; . ', ---_ t.00 - T p: .
X5 -30.0- , .:t., .
.i< . , . . . . .
m ..- -
< -m 3 .: .
.~ ... . ..
.- m-e 8 :t : i!*
<- - X no 24.0- ;- -
O ll; l;g ;
.. . - - 0.75 - o 'd
- l, ll
' e 18.0--
l $ ,-
} :.
, r y,,;
I *.
- Q-
, ;; .: . e. ; , .-
l ';.g ,,t; . / - ; I....,.s...,' ,,.. - 0.50
. . :.. ll ' f - ,*...
,c t 2.0 - l ..
- s. ",
/ .., ,.
. / ....!
e
/ . ..=**+
; .,/- *
.,.,...." ', ,' O.26
~
8.0- -/ l / LASA!.l.E MELCOR 111011 PRESSURE :
. .r Sil0RT TERM STATION BLACKOUT
, a aa 0.0 , , , , , , , , ,
0.0 12.0 24 0 36.0 . 48.0 . 60.0 72.0 - 84.0 96.0. 108 0 120.0 TIME (103) 3 e Fi Bure 5.1-36 Concrete Degassing Masses
-5 90
140 , , i , , , , i s. LASALLE MEIr0R HICil pre.SSURE SHORT TERM S:TATION DLACKOUT
-X- Drywell Mass l Wetwell Mass ~ 3 t es- ---
WW Pedestal Mass - DW Pedestal Mass - 3.2 Downcomer Mass 1.12- - - -X - DW Mole
. + . . w y M ole - 1.1 0.98- -- 1.0
- 0.9 004- -
9M - 0.8 U O n s ..y o-3 0.70- 4' .
-- 0.7 N
' C y .
.R y ,',
- 0.8 j 0.5e - ,- ,.
*+
,s.
- 0.5 o.4a - l ~~ '4: '
-f ,*. -
04
- ' */
0 28- 4 \ :
.X"*..,l..,'M......
4 .
-+-- - ..., ~+..- -_ o'3
-- [ l
.... ,,M ..
- 0.a -
o.t 4 - .
,y ........- x..' --....+.: .:: y ; -
,,. -. .., - c.t n_oo h ; M~;(5$ ,p- - - -_- ~ _-- - - -f_ -} '= -7:- :k . .: = ,,
0.0 12.0 24.0 36.0 48.0 60.0 72.0 84.0 96.0 108 0 120.0 TIME (103 ,)- Figure 5.1-37 Containment liydrogen Distribution 5-91
110 0 , , , , i i i i i LASAILE MElc0R HIGH PRESSURE SilORT TERM STATION BLACKOUT - 226.0 100.0-
,- 200 0 90.0 60.0- --- 175.0 70.0- -
.g - 160.0 -
U > g 60.0- - {J v - 125.0 { j - - - - Steam . ;3 - e: so.0 - } .. ...
. Hydrogen
~
- Nitrogen . - 300 0 0 -
i E 40.0- -
- 75.0 30.0- --
l - 50.0 20.0- ' -
~ '
10.0- -- I j% . 0.0 ,
- i. .. ',- - - , -
i
- , _ _ _ h, ,
__._~
~ I, _ 7 - _ ~,
! 14.0 14.1 14 2 14.3 14.4 -14.5 . 14.8 14.7 - 14.8 14.9 15.0 TlME (100) s l l i:- 1 l-Figure-5.1-38 Containment Vent Exit Gas Flow Rates. i .. 92 - l - - r - + ____l
i t s.0 , , , , , . , . . LASALLE MELCOR 111011 PRESSURE S110RT TERM STATION BLACKOUT 14.0 - - 30.0-12.0- *
- 25.0 10.0- -
- 20.0 6 .0- - ^
.d {~ ~.
M - 15.0 - 6.0 - -
~
> - 10.0 0
C _ 4,o 2.0 - -
.0
\ *l I
4 l A a.o ip .. ......
...... .............. ... . . . . . . . . . . . . _, oo l ,.
-a.O .
- 1
-4.0 , , . , , , , , ,
0.0 :2.0 24 0 - 36.0 48.0 60.0 . 72.0 84.0 96.0 '108 0 320.0
.' TIME (103,)
Figure 5.1-39 Containment Vacuum Breaker-Flow Rate- . 5-93
1.00 , i , i i 1 , i i LASA!JE MEICOR 111Gli PRESSURE S110RT TERM STATION BLACKOUT 0.7S -
- 1.6 I l' ItUj\-
040- ( g(} 3, g
- 1.0 I \ l Il \\$ ,\ {
it f I' f! pt',11 g ; f t4 0,00
,P 4
b {
++- - --
i
** ** - - 0.0 v
/ \ \
i e-M -0.25 - -~~ ' v C
-- 1.0
-0.50 - -
- - 1.6
- 0.75 - -
- - 2,0
- t.00 -
Steam
--- Carbon Dioxide -
- - 2.5
-1.25 , , , , , , i . ,
0.0 12.0 24.0 36.0 48,0 60.0 72.0 - 84.0 96.0 108.0 120.0 TIME (100) s Figure 5.1-40 Vetwell Pedestal to Drywell Pedestal Flow Rates 5-94 l
2.50 , , , , , , , , . DW Pedestal to Drywell e.26 -
-- WW Pedestal to Wetwell .- n.u WW Pedestal to DW Pedestal
- - - Vent Exit Vacuum Breakers 2.00- l Drywell Drain ~- 4.s
- 4.0 1.75 - -
9M - 3.6 n 1.50-O
- 3.0 3c 125- O W-t: )o - 2.5 0
o N 1.00- , j - 3 eo r N v a i ,,.~', - 2.0. s *,.- e ,I y 0.75- _._.._~...::.y. . j p s - 1.0 r i
\
0.50- #
-/ ..
s -
/ ./ '\ - 1.0 s a
/ \
0.25- [ '
'g .
- 0.5
),. ....
.............................2 %. q ............
e . '~. 0.00 l
. A - .a
~,'~~ - 0.0 LASALLE MEICOR HIGH PRESSURE.
SHORT TERM STATION BLACKOUT
-0.25 i , , , , , , , .
--0.5 0.0 12.0 -24.0 36.0 48.0 60.0 72.0 64,0 .96.0 106.0 320.0 TlME (103 ,) .
Figure 5.1 41 Containment Integrated Hydrogen Flow Rates 5-95
4.0 , , , , , , , , i 4.w LASA11E WELCOR HICH PRESSURE S110RT TERM STATION DLACKOUT 3.6 - -- 0,9 lbtal Mass Flow s.e - ....-+ 11 drogen Mass Flow "oe.
--- 11 drogen Mole FYaction l
2.6 - -- 0.7 l 24- -- 0.8 ng a O i v c y 2.0 - -- 0.6 g > E 2 c 1.5 - -- 0.4 M 12- -- 0.3 0.8 - ,
-- 0.2 .
0.4 - -- 0.1 65.0 70 5 75 0 0 L,6 87.0 - 92.5 98 0 103 6 . 109.0 114.5 120.0 TlME (103) s Figure 5.1-42 Wetwell Leakage Flow Rates l 5-96 .
t 350 0 i , , j; i , , , , i M LASALLE MELCOR HIGH PRESSURE
): 1 SHORT TERW STATION Bl ACKOUT 346.0- t I 1e0 0 i
i ---- - . Lower-Unit 2 i Upper-Unit 2
' t
-- Unit 1 340.0- __
' n -
Refueling Steam 1bnnel Day /lbrbine Room 2 25 -
): - X-- Steam 'Ibnnel/Ibrbine Room 1 I *~
335.0- - I
- 140 0 1
2 330.0- g e v o l I ; - 130 0 .N
$ t )\
- 325.0- g ,
g o g
~~
\
g .. g - 12 0.0
@ 320.0- 1 I, i -
' \ ;
g s . s Al \ - 11 0.0 i 315.0- ;. l,
'g q
);\ \ --, -
s,
',- 'Ns, ~s
~}'\. ,
\ - 100.0.
310.0- ; \ s
, '-.1-~,___,, -
( 4--.. ,- 305.0-@ _- 90.0 300 0 , r , , , , , , , 55.0 70.5 76 0 81.5 B7.0 92.5 90.0 103.5 '109.0 114.5 120.0 I TlME (10 0 s). t i l :. l I: [ :. Figure 5.1-43 Reactor Building Vapor Temperatures 5 97
. .- .. .- . .. . . - . . . . . = . _ . . . - -. -
J I b E 1 4 0 20 i
, r , , , , , , ,
s n' \ t H2O 0.t a - s H2 -
, - - - og
;,~s ---
CO gI s -- - CO2 'i 0.16- s * , y s s s
. I ' , -
g %l -/ *
,/
0.14 - / *
,/ -
l i < l
,--./..~~#
j / '~- 4 0.12- i e - 1 c t <
/
.9 s
/' I a J 0.10- - -
s
/
.v. \
i o /
/ . y '
3 0.00- s/ - t, .
/
i i /. ,.
- j.'N
0.06- / f ...
+ **.. .... , ... -*
*... *... ,.'. / ./.
> ), . / ,' *-
i ,. ./ 0.04-s's.
' ll , * ' ./ ..
, . ' ,*/ ,. *
,s*
s * . *
/
*/ *..-s'./
, 0.02- , ;I- i ,
- l. LASALLE MELCOR HICH PRESSURE
! / lN'.'" SHORT TERM STATION BLACKOUT - l 1' l 0.00 , , , , , , , , , 650 70.6 78 0 81.6 87.0 92.5 98.0 103.5 100.0 114.5 120.0 TIME (103) s j l i 4 i Figure 5.1-44 Upper Unit 2 Mole Fractions i 5 98
i : 1 i l l 4 0.0e0 , , , , , , , ,
- -- Lower Unit 2 Upper Unit 2 0.o7a- ----
Unit 1 - l - Steam Refueling 'Ibnnel Bay /lbrbine Room 2
,_,,,, , X- Stearn 'Ibnnel/lbrbine Roorn 1 _
'p ;-
/
- 5 O.056- f ./ ' /
../' -
m I,/
,'~p' l h I ,/ e/
/
{ 0 048- I*
/ f ~ '/ -
@ //
! ,l
% //
.9 e
'/
o 0.040- ' ./ //
/ . -
C I *l /j Q t/ /.' 2 0.032- tj' / ./ - p, // m j/,
/ ,*
s , 0.034- 'l 1/ i j LASALLE MElCOR li!CH PRESSURE i ,f , 3110RT TERW STATION BLACKOUT f Il
- i /
0.016 *
.,'l... .. . . ....
I
,/
,l 1,.}..........
... ........ ~~....... - ..-.-.. ~.--
0.008- if i/ - C q ij
,, ,3, .._.7___t..,__ _
Wv
,,- n i s,
,s
! s,
,s - l p,-
55.0 70.5 76.0 81.5 87.0 92,5. 98.0 103.6 109.0 - 114.6 IP0.0 TIME (103) s t Figure 5,1-45 Reactor / Turbine Building Hydrogen Mole Fractions l l 5-99
50.0 i , , i i i : i i LASALLE MELCOR HIGII PRESSURE SHORT TERM STATION BLACKOUT 42.0-At 115712 Seconds ENveRoN u ENT .[ 000'0 [Kg/Sec] nEruruMG BAT - 875.0 34.0- - s.oos -
*st r
on .oooze C .oe4s - r e .4, t.e s s t . osse - 850 0 M J
- 2e.O - l1l714 31.24l l -
V FUEL
. StoRACE W PO '
- 825.0 h
m 10.0-O La. C v
- c d - -
e
- 800.0 o M $ C ! C o n o
.9 10.0- -
C o - - M
"$ N y Q) um:71 33,s4 u*PER - 775.0 W g
y uMIT a C a w 2.0 - - - M 4 E E - 750.0 09 "O - 6.0 - ETEAM TuMhEL AND .sae STEAM TUMMEL AND - Tuna:Ns moom uM:7a .s e. TumerNE moow uM 7m L ,. , , J - 725.0 h g
-14.0-
' * * * ** *ll eem s
- o. o. - 700.0 ho,7,Eg
-2 2.0 - -
- 675.0
-30.0 , , , , , , , . ,
0.0 1.0 2.0 3.0 4.0 5.0 6.0 7.0 0.0 9.0 10.0 3 Volume Area (10 M2) i
, . . .m . . . _ . . - . - . _ . -. _ - . = _ - _ _ . - . . . _ . . . - . _ _ _ . . . . _ - . - - _ .. __ .
50.0 T i a i i . i LASALLE MELCOR HIGH PRESSURE SHORT TERM STATION BLACKOUT At 115712 Seconds ~ 80 0 42.0- EwvanoNwEm? _ m r [Kg]
$" 34.0-mEruEunona?
- 1.8E8 -
- e75.0 a
v see sees - i.eze e.oE4 - - s e7 e W , e,oE4 s.7E e , , _j - 650 0 p 26.O- i i : - N FUEL STORADE [ ~ ~
- 825.0
$ 18.0- -
n R w kv ~
- 600.0 C
o
- u. g C g---
- c .9 - 10.0 - e o r s@ - -
& y tu h
g Uwr? 1 1.7Es UPPER - 775.0- 5 M w' Unst a a C
. 2.0 - -
m D n - 750.0 e
" -0.0 - STEAM TUNNEL AND STEAM TUMMEL AND
" TumanwE aoom uwst 1 e .4 E+ i .eu. Tumaint accu up 7a n
a L J - 725.0
- 1. O E 5 C.
**~'
m ~14.0 - _ s {*galt _ I' k msse g,E R sete - 700.0 W m -22.0 - - - n
- e75.0
~30.0 , , , , , , , , , -
0.0 1.0 2.0 3.0 4.0 5.0 0.0 7.0 0.0 9.0 10.0 3 Volurne Area (10 M2)
16 : i . . . g . . , , - LASAllE ME100R HIG11 PRESSURE :
- S110RT TERM STATION BLACKOUT .
1d =: = 16 = 1 i
................. -.- - -- -~~ - i .
ac 10~'
- =
x : # / :
, /
m . s . - C a
*10-' : : I .-
/ =
, f ____ - -- --
t , t : / 10~ ' = 1 ! ' i e i / i
- , : e :
/- -
l 10~': i j
/ In-Vessel Total . 1 i j --- In-Vessel Radioactive :
- , i , ------ Ex-Vessel Total . :
,_ jj ---
Ex-Vessel Radioactive
' I' 10 , , , , , , , , ,
0.0 12.0 24.0 35.0 -48.0 80.0 72.0 - 84.0 96.0 108.0 120.0 TIME (103) s Figure 5.1-48 In and Ex-Vessel-Releases of Cerium from Fuel 5-102
4.0 - , , , , , , , , LASAILE MElf0R HIGH PRESSURE SHORT TERM STATION BLACK 0UT 3.5 - - 3.0 - -- F 88- ~ Concrete - q ---- Steel M M v 3 2.0- - In
- as .'
X l 1.5 - l l l 1.0 - l - - l l l l 0.5 - l - l l 4 4 0.0 .
)*'. , , , .
O.0 12.0 24 0 36.0 48.0 60.0 72.0 84.0. 96.0 . 108.0 120.0 TIME (103) 3 I e E-t I ( 1 I l. _i-l
-i l
l
~ Figure 5.1-49 Uon-Radioactive Ex-Vessel Releases I 1
5-103' 1 4 , . - . . . ._ --.
10 - 44 y i , , , , , , , _* + LASALLE WELCOR lilGil PRESSURE f O9- i ; f. SHORT TURM STATION DLACKOUT irf > p L 3 30a-e f ,, 4 l
.h l
% .yV .__ . ._. , __. . _
; __ I 3o ,,
- x. X _.
x '- 7m o X X X X X-m
. . c'
*s
>@ ,o I 06- E -
c ~o' t -
-- Xe
; ^
v --- Cs 3 . . _ . . - 9a 0.5 - _.-- ._ _______ ,------ _ _ __ E _ - - . _ _ -'j-x -X- Ru
" 4-s.t Cd E o4 '\ -V-
-O ' Csl )
Sm
,);, -
' }
N 0. 3- ! i / i t , e t o \ g')e (2 e , s ~ _ g --_______ F 0.2 - 0 h :4 o - v c ( v o-
-s 0.1 - -
[' 0.0 - - {h ,' -^' , i- , ,
' ,84'. 0
~, -' P .- -
00 12.0 24.0 36,0 48,0 60.0 72.0 96 0 108.0 120 0 TIME (100) s Figure 5.1-50 Primary. System Retention Fractions 5-104
500 0 , , ,- , , _ __ , , , ,
. LASA1.LE MElf0R HICli PRESSURE Sil0RT TERW STATION DLACKOUT 460.0- ~. j"~' '
1 - i
\
400.0- I I 1 \ t \ r \ 350.0- j - g 1 \
'l \
300,0- #
\ -
I \ S x 1
\ $ 250.0- l g -
M
- i \
3 \ coa.o. ) Total Released s. -
-- --- Primary System
- l '
Containment ' l
- + - - Reactor /lbrtine Bld '
1500-ll'- X Environment '4 . --
.t
(
...e-s
.s
, s 100.0-- '.O, g .' .f' 's- s t *:.,
** + /
,.' X -.
50.0- I
/ -
t I
,' ' ~.60.0 TT 0.0- 4X 0.0 12.0
'NI*
24.0 i' 36.0 tX; - 48.0 ~,'.0 72
'" " ,' 84 .0 06.0
'y 100.0 120 0 TIME (10 3 )'
3 . Figure 5.1-51 Xenon Mass Distribution 5-105
.. , s a .. ~ - . ..- . . _ . - . ~ ,, .- ,,_ . . - . - . - _ . . . ._ - ~ .
1 4-4 4 a 2 2 N 7 4 060 0 , , , , , , , , 4 LASALLE MELC0h tilGil PRESSURE 225.0- SHORT TERM STATION Bt.ACKOUT . ? 200.0- . j t 'Ibtal Released . j- ----- Primary System - ino - .,: Containment -
- .: Reactor /lbrt3ne Bld
-: -X- Environment ..
4 + 150 0- .f . } I
, To ; i M .
---- - ----- l v ; ,,_ _,__.,. _.
125.0 - __s
- p , ,
e
. /
1 4 X '
/
' ~'" * * * " ' ' ' " ' " * - - " * ----- .....- ... . ... .,
100.0 - ,, I I J l i 75.0- - t i 1 I l !" 60 0- l l I
- 25.0- t _
;. I I ~
0.0 - % , ltX , 4X, , lX , , -;X ,- ,;x 0.0 12.0 24.0 36.0 48.0 80.0 72.0 - 14 4 : 0 96 0 108.0 120.0 - i ' TIME (10 s3)- 4 i Figure 5.1 52 Cesium Mass' Distribution 4 1 5-106
~ . _ . , . ,._,s_ , .._ - , .._.- _ ... . m ~. . . . . _ , . . . . ~ . . _,
, . ., ~ - . . . .
+
k' Y i a { 2 .. u. 4 e l l , 3. I 400.0 , , , f 5 Ex-Containment. Tbtal
---- Environmental Release >
i. t 350.0- - 4 Xenon Class 1. 1 300.0- - - 1- F i 1 1 260.0- - e m I
- M 7, 200.0-e *....... 4 j
I llll! .,...*.. - i, j 150.0-
...+**..... ,,, '. -
t. l 100.0 - ' - l l 60.0-
- l .i LASALLE MElf0R 1110l! PRESSURE 1 S110RT TERM STATION BLACKOUT
, 0.0 - 65.0 70.6 76.0 81.6 87.0 02.5 ..98.0_ 103.5. - 109.0 .114.5 120.0 - ' 0 TIME (10 s)-
-i
- l
, i s j , 1
.i
-l i.
;l l Figure 5,1 53 Xenon Source: Term to Environment l ;-
i ! 5-107 ' I,
- i
= . _ . ..
j l 1 4 0 13 - - , -y , , , , m i i LASALLE MElf0R lilCH PRESSURE SilORT TERM STATION Bl.ACKOUT 0.12- /
+ Cesium
"~ X Barium ~
A Tellurium O Cesium Iodide 0,10 Ex-Containment Total $ 3 -
- Environmente.1 Release /
/
0.00-l 0.0 s - / - I -
/
. E 0 0,- -
$ 0 Or - -
x A l 0,05J _ l s i 004- - 7 0.03- X -
* . por
*~ .I
/./ 4_. .. .. -+-
~
_ _ . . . + ;. :. : -
. :bO.
'^
,. . .O -
~
0 01- v%%:e =:' "'- -k?'^' '- ~~ ' ~ ~X* * ~
. 'I 0.Do - M$ a&
SS 0 70 5
"- 78.0 81.6 87.0 92.5 98 0 103.5 109.0 t14.5 120 0 TlME (100) s i
Figure 5.1-54 Cesium, Barium, Tellurium, and Cesium Iodide Source Terms-5-108 9
t-s W 4 .i Ao i > i > r i i i - , LASALLE MELCOR HIGli PRESSURE $ j SIIORT TERM STATION 13LACKOUT 45~ +- lodine - X Ruthenium " Ex-Containment Total 4.0 - Environrnental Release , , N
~
3.5 - - a i E 30- - Ch i o 2.5 - - 4 O _ rn N M 2.0 - ..-+-*'~ -
.....+ * ,,,
~ '
_.....+***.. ts_ .---F . 1.0 - l s 4.- O.5 -
'M i -
.' #s ..-
0.0 -i(W, "
, , , , r-
' SfLO 70.S- 78.6 81.6 87.0 92.5 98.0 103.5 100.0' 114.5 120.0 TIME s (103) e f
9 Figure _5.1'-55-Iodine and Ruthenium-Source Terms - L 5-109 . t
l d 0 50 i i i , , , , i i LASALLE MELCOR 111011 PRESSURE SIIORT TERM STATION BLACKOUT DA5~ l + Molybdenum
~
, X Cenum
; A Lanthanum 0.40- Ex-Containment Total -
---- - Environmental Release 0.35- -
j 4 g 0,30- - M i n i O 0.25- - O E 2 ~ d 0.20- .- . 0- . .. -4*' .X' -
..X-i s . .. . X' . -d -
4 0.05- *
.. .-8-* -
. , .. . g... ...-- "d'
fk 0.00-M j - i ' . . Q'.'. '
- eb.0 70.5 78 0 81.6 87.0 92,5 98.0 103.5 100.0 1 4.5 120.0 TIME (103) s i'
Figure 5.1-56 Molybdenum, Cerium, and Lanthanum Source Terms 5-110
0 40 i , , , , , , , , LASALLE MElr0R HIGH PRESSURE SHORT TERM STATION I1LACKJUT - a y ,. + Uranium . . X Tin Ex-Containment Total
---- - Environmental Release 0.30- -
0.25- - - 'E3 x 7 0.20- , a x 0.15- - 010- - 0.05- - p[y
+X 0.codtMc 65.0 70.5 76.0 01.5 5[
87.0 92.5 98.0
; -X~ '
103.5 109.0 114.5 120.0 TIME (103) s Figure-5.1-57 -Uranium and Tin Source Terms 5-111
.. . .. . -.. ... . . - - . . _ . . . . . ..- . . . ..- . .- ~. . . - - - . _ _ - . . . -- .
so . , , , , . . i s Ex-Containment Total; Environmental Release - - 4.6 - Cadmium i 4.0 - 3.5 - ) i_ g 30-M n I-o E6- --
=
m. m e
" 2.0 -
1.s - ... - ,, .. - 4 1.0 - - l l } o.6 - [* . LASALLE MEICOR 111011 PRESSURE '
~
,. - . S110RT TERM STATION BLACKOUT 00- , '. . ., , .. . . . . i 65 0 70.5 76 0 814 . 87.D - 92.S 90.0 . 103.6 - 109.0 114 6 120.0-TIME (103) .
s i !~ l' - Figure 5,1-58: Cadmium Source Term: ,
.-5 112
, p- ,
15 0 i , , , ., , , , i r \, Ml j 1 ..... yo in.s - *
--- Cs -
l I,
---- 1 (
j \ - - - Te ~~ s K , Sn g 12.0-~ _ _l
'1 X Csl
- ~
E D r, c l , \. 2
/
\
; 10.s- l- s \. -
c \, w 3- -; 4 s l 6 o.o-- ' e :t 'l s !
.5 i s s T
e 7.6-
' /'i \_~ \ l-6 j' s
i s s, s s E. i s - ys 6.0 --' ,,, ,, y - s .. y - , I ., e l 'l 8 3o 4 s- ,: p
~
u x '~,-~ ~ ~ ,_ ,, _ _ _ ^ - - n
. , ~ . '
o 3.0 - '~
- F l j
LASALLE MELCOR HICH PRESSURE I SHORT TERM STATION DLACKOUT v , I ' o.0 f , , , , , , , , 65.0 70.5 76.0 81.5 87.0 92.5 ' 98.0 103.5 109.0 114.5 120.0 TIME (103) s 1 l I l l l Figure 5.1-59 Reactor / Turbine Building Decontamination Factors-5-113 1 1
-l
id ; . , , i . , , , 3
,g 2 < os>
10~ ' : LASAllE MELCOR HICil PRESSURE : SHORT TERM STATION BLACKOUT 10~ * - 1 E 2 0-* s x : i
} [+ !
$10" _ 1 ;
- -o .5 - 2 MICRONS E i
-O- 2 - 8 MICRONS :
o 8 - 32 MICRONS 16 - 32 - 128 MICRONS I X 128 - 512 MICRONS i
~
16 * , = i i X~ ' : 10 = # ; 10"- Y e , , , , , , , i 65.0 '
. O.S 76 0 61 5 87.0 92.5 98.0 103.5 109.0 114.6 1200l TIME (103)s l
l Figure 5.1-60 Reactor Building Aerosol Size Distribution 5-114
200.0 , . . . . .i i i , g, stem Total LASALLE MElf0R HIGH PRESSURE - 650.0 M Rd SHORT TERM STATION BLACKOUT g7g,n _ e d
.- 600.0 u - 550.0 150.0- - Htamem.
- u. ..
,,, y a -
_.n,j
,, na
- 450.0 125.0 - -
u-
^
k me
- 400.0 T g1 O
g,
"' s 100.0- ~
"' E - -
.h D u. S k o
& u- ,
,u
- 300.0 . C 75.0-1 na.
i
.v{i\
.u -- 250.0 u, ;
'h*..*
J
'* , ,,~~~-........ ... ,, l
.....s,,
60.0- g. * " " " u as as ea esa .
.s e
. " "we" .
see ma eas seu ma 8
- 150.0 TIME 00 4 25.0- ', .
- 100.0 p -
- 50.0 t,
0.0 , 3 aa 0.0 12.0 24.0 36.0 48.0 60.0 72.0 84.0 96.0 108.0 120 0 TIME (100) s Figure 5.1 61 Radioactive Decay Power Distribution 5 115
16 , , , , , , , , , LASA11E nlEILOR lilCll PRESSURE t4- Sil0RT TERhl FIATION BLACKOUT
-- roo o
?. ':
1.a - - 1.3 .- sts.o l 11- , i;0NTAINMENT BREAK ARF 4 SENSITIVITY STUDY - d, - 160 0 to- - 2 Oo-o - seso I o,s- s, 3 o' Case 4 .On00 m2 & 07-
>-+- Case 2 .0002 m2 -
X- Case 1 .0924 m2 ~- 100.0 ~ c., -O-- Case 11 .924 m2 oe-u6- - -' 70 0 O o4- -
- 60.0 0.3 - '
Wetwell Wall hilure
, 6% Hydrogen Ignition Limit o& . .
- 26.0 0.1 , J( , M , , , ,.. ,
85 0 68.2 67.4 e8 8 89.8 71.0 72.2 73.4 74 0 75 8 77.0 3 TIME (10 s) Figure 5.1-62 Containment Depressurizations 5 116
--y.-- , .-
m -
400 o . , , m -, , , , LASALLE MELCOR HIGli PRESSURE f SliORT TERM STATION BLACKOUT A Case 4 .0000 m2 4oo o. . O
+ Case 2 .0092 rn2 N X Case 1 .0924 m2 0 Case 11 .924 m2 35o o. Ex-Containment .
Divironment , ano.a. ........y... ~
..M. .. . .
9'i' ~ s esoo_ i :, .-+-~~ . M ' : 4... ...** w . g .' : ..+..- m 200.0- 4; 4 : .... - 0 :
- x /
f CONTAINMENT BREAK AREA t oo.o - l SENSITIVITY STUDY _
.+'
loo o- '
,i : ~
': Wetwell Wall Ibilure i
. 6% Hydrogen Ignition Limit bo.o-
.' .V j
/y ,, ,,
- p. ....t " " 'O '......... 6 oo ._ . . . . . . . . .
55 o 662 6'
- 68.6 69.8 73.o 72.2 73.4 74.8 76.8 77.o TIME (103)s l
l Figure 5.1 63 Xenon Source Terms for Break Area Sttidy 5-117
2 50 , , , i LASALLE MEISOR 111C}! I'RESSURE v_ SFIORt TERM FTATION BLACK 0UT ets- < e.co- -
- l 1.75-CONTAINMENT BREAK AREA SENSITIVITY STUDY 33o. g........... .
g ......f..' g ....
" 1ts- . *. . .{t . ' -
C ..:
; o, ,+ :
.: M -f - X- .
X, ,. . :
- :+
0.75 - : A Case 4 .0000 m2 - [
- .' + Case 2 .0092 m2-
+: X Case 1 .0924 m2 e '3, . : O Case 11 .924 m2 9.; ' ,
- Ex-Conta:nment Environment 025- -
f .'* + Vetwell Wall Ibiture (),.A ,' ; 6% liydrogeo Ignition Limit
%* n a <
' esi" es a ev.4 ee e se a 71 o 72 e 73.4 74.s 7s a 77.o TIME (103) s Figure 5.1 64 Barium Source Terms for Break Area Study 5 118
500 0 , i , , , , , , LASALLE WELEOR 111G11 PRESSURT Si!0RT TERM STATION DLACKOUT X X~ 460.0-
. .& v n --a :=F_ -
A Case 1 - Wetwell Break
'*~~ + Case 3 - Drywell Dreak ~
X Case 9 - Dh llead Break Ex-Containment Environment 350.0- - E 300.0- , ,-Q:.'.:.:h:::: ::g::::h::::::::g: .. . ..
, ...y ... . . . ... -
g.q. T 250.0- - a ;!j ;: x . 200.0-s, 8...:: ' l l CONTAINMENT BREAK IDCATION
,! a. . .
+. SENSITIVITY STUDY 150.0- .
.,: 4 100.0-b lr. ,: .
'l .0024 m2 Dreak Area 60.0- ll 54 Ilydror,en Ignition tirnit _
bk 0 0--MI 65.0 66.0 67 0 68.0 69.0 70.0 71.0 72.0 73.0 74.0 75 0 TIME (100)s Figure 5.1 65 Xenon Source Terms for Large Break Location Study 5 119 o
__. = . - - 20 1 , , . , , , ,
,o,
-- x -.
-X-Le- / -
A Case 1 - Wetwell Dreak
, + Case 3 - Drywell Break ie-X Case 9 - Dh IIcad Break --
Ex-Containment Environment 1.4 -
, -a .. + -W ,
g ie-4 [ CONTAINMENT BREAK LOCATION j I
, t.0 -
SENSITIVITY STUDY - 0 M ,
.t 0,8 - ,1
/ 0924 m2 Break Area
/ $% liydrogen Ign1Llon L.imit 0.8 - ,
, .y..,,,.....y,,,,,,,,,,,,,,y,,,,..........X.-
. . _p , ..
~
04- 1 K'l 3 . ......6... De- l
.b+ ,
l LASALLE MELCOR 1110!! PRESSURE
' g. : :. ' '.' '
SH0HT TERM STATION DLACKOUT o.0 , , , , , r , . . 65.0 66.0 67.0 68.0 69.0 70.0 71.0 72.0 73 0 74 0 75 0 TIME (103) n Figure 5.1 66 Cesium Source Terms for Large Break Location Study i l 5-120
1 l l i' 2 50 - - LASAllE MELCOlt it!Gli Im r.SSURE Siloft? TEltW STATION DLACK0UT e ro-6- - r.oo -
/ .
2
** ~
A Case 1 - Wetwell Dreak
+ Case 3 - Drywell Break X Case 9 - Dh Head Break 2.60- Ex-Containtncnt -
U Environment M Y 126- - a w
, 3 ..6.. .-. . . 6- ,
f CONTAINMENT BREAK IDCATION a.73 SENSITIVITY STUDY 4 ! - 0.60- ; X ^ - l A .0924 m2 Break Area
, 6% llydrogen Ignition Limit 0.25- *
, , _ p.x. . . . . . . . .x . . . . . . . . . . . . . X - - - - X - - " -
0.00-- 65.0 P 66.0 87,0 88 0 m , , , , , 09.0 70.0 71.0 72.0 73 0 *te ,0 75 0 l TIME (100) s 1 1 1 l Figure 5.1-67 Barium Source Terms for Large Break Location Study 5 121 1 i
0 70 , , , i r i i i i 1.ASAL1.E ME1 Con !!!Gil PRESSURE Sl! ORT TERM STATION D1KKOUT
' 80 ~ A Case 2 - Wetwell Break .
~
+ Case 10 - Drywell Dreak 0 os. X . Case 5 - DW ~licad Break -
Ex-Cont.ninment-Environment - e.co- - 0 46- - _ o 40- CONTAINMI'NT BREAK LOCATION - o _ x SENSITIVITY STUDY 7 0,35- - a 0.30-
~
0.25- {* '
--. +
A..,.,,.j,,:"+" 020- ,j.. -p c is- :i:,,,p'.'..x""-",......X#
~
, ,y . . . . * , ,&$$
0.10- .
.-.~[~~' -
. ' ' -- .00924 rn213reak Area
. h *. * , . ' - ' 5% liydrogen Ignition Litnit 0.0S - *-l-> -
.. ,.g f 0.00 --
65 0 4"' 66.2 67.4 60 6 69 8 71.0 72.2 .. 73.4 74.8 75.8 77 0 TIME (103) s 4 Figure 5.1-68 Cesturn Source Terms for Small Break Location Study 5 122
2 26 i 1 i ~7 i t i i i LASALLE WELCON 111G!! PRESSURE S110RT TERW STATION DLACKOUT f 6 2.00- , CONTAINMENT BREAK thCATION SENSITIVITY STt!DY 1.76- . 1,60- _
.........6..
,.....6...'
....h ew tas- : b' .
; .00924 m2 Dre6k Area R b% Hydrogen Ignition Lirnit
$ 1.00-o.7s - ,'O
- .: A Case 2 - Wetwell Dreak o.oo- A + Case 10 - well Dreak X Case 5 - DW Icad Dreak -
- Ex-Containtnent Environment
?
0 26-A ooo- - g, & ...n su n W ;.n .,,....l U,........... h l... r >a -:55$:~ . ::':::A=X 660 66 2 67 4 68.6 69.8 7't .o 15.2 7b.4 Tk 8 76.8 77.0 TIME (103 ,) Figure 5.1-69 Barium Source Terms for Small Break Location Sr.udy 5 123
otto- , i i i , i 1 i i LASALLE WELCOR ll!Gil l'RES$Uf1E Sil0HT TERM S1AT10N DLACKOUT C018- -
)
1 j A Case 4 - Wetwell Break
- 0018- + Case 8 - Drywell Urcak .
~
X Case 7 - DW IIcad Break ' Ex-Containment n oi ~ - Environment - 0 012- - G bc
- 0 010- '
E CONTAINMENT DitEAK LOCATION n A SENSITIVITY STUDY 0 00e- :'**"- 0.006-b l l ,. . . . . .p
. . . . . . . . . .X " " " ' ;
0.004- - g, ..;
. . . .{-
~
0 002-
/p 7 , . . . . . . .
l
.,.u .f'!.
. . 9 , '. ..' ' ' ' ' .000799 m2 Dreak Ares p4X" " *
- 5% liydrogen Ign1 don Limit 0.000- ' " My--MM ,
65.0 66 T. 67,4 68 8 69.8 71.0 72 2 73 4 74 6 76.6 77,0 3 TIME (10 m) l Figure 5.1 70 Barium Source Terms for Small Leak Location Study 5-124
_ _ ._ .._ .. . . __ . ~ . . _ _ . - - ~ . . . . . _ _ _. . _ . _ - ._ = _ _ _ . ~ . , , . _ _ . _ . _ _ _. 2 60 , , , , , , , , i LASAtl.E Melton 111 011 PRESSURE S110RT TERM STATION DLACKOUT 2 25- y -- 5 . m.co- -
"*' ~
+ Case 1 - 5% 112 X Case 0 - 100% 112 l 4 Ex-Containment l- 1.60- ......
Environment - 6 x t ,26- - R e 2 - 1.00- .. +. .... ....... +.. ...... ... + .. . ......... +.... .- - j llYD110 GEN IGNITION LIMIT o.75- j SENSITIVITY STUDY - 4
+
0.60- {
~
Wetwell Vall FMlure j .0924 m2 Break Area I- 0.25- 1
+, , -
. . x..............x.............x o.00-ry - > > , , , , , , .
66.0 66.5 $6,0 68.5 87.0 07.6 68,0 68,6 69.0 09.6 70.0 i TIME (103) s I i l'igure 5.171 Barium Source Terms for Burn Effect Study 5 125
- p. wy ----r* ye$
e mw=-+ - ym B - ,,-._w -
'-r-----
_ . . . - _ _ _ _ _ _ _ _ . _ _ . _ _ _ , _ _ - - . . ..m__ - _ __.-_..__-.m.____ 250 0 , , , , , , , , i LASALLE MElf0R 111011 PRES 5URE S!! ORT TERM STATION BLACXOUT teso-
+ Base Case - 10% 112 X Case 4 - 5% 112 -
Ex-Containment ro0.o- .~~ Environinent 175 0- *'
~
if,0 0-llYDROGEN IGNITION IJMIT
,d SENSITIVITY STUDY ,-"-+""'
125 0-E l j e +. 1 100.0-l l 75.0- {
+ .
60.0- -
.. ..: . + * ... -
es O- .... d .t **,,, ... -
,}.. . . . //l '/ * -Y......"'.. Wetwell Well hilc re
.000799 in2 Break Area
"".. . . x 0.0- , , , , , , , ,
D M 1".0 65.0 67 69 0 71.0 73.0 76 0 77.0 79.0 81.0 83.0 86.0 0 TlME (10 s) Figure 5.1 72 . anon Source Terms for Ilydrogen Ignition Limit Study 5 126 -- -. _ _ _ _ . - _ . _ . , . _ . - ~ - _ . _ _ _.
i 5.2 Low Pressure Short Term Station Blackout All onsite AC power fails, in the low pressure short term station blackout I sequence, resulting in the failure of all ECC systems and containment heat ' removal capability. The ADS continues to function normally and the core : meltdown occurs at low pressure. This sequence is described in more detail in Section 4.2.1. 5.2.1 Base case 5.2.1.1 Overvies The progression of the accident is illustrated with the list of key events and the timing of those events found in Table 5.2 1. After-the accident was initiated and the reactor tripped, the decay heat began to boil' water within the core and the ADS was actuated to reduce primary system pressure. l Initially, the ADS was _ operated gradually, such that, the pressure was- l reduced at a rate corresponding to a reduction in the _ saturation temperature of 100 *F per hour. The water level in the core dropped below the top of the active fuel at 26.6 minutes and the exposed fuel began to heat, j According to the ADS operating procedures, all ADS valves were opened at l about 52.5 minutes when the core water level dropped _below the 275 inch l 1e, vel and the vessel pressure was less than 700 psig at the time. The primary system rapidly depressurized af ter _ the valves. opened. The ADS
- valve flows were piped into the suppression pool where. virtually.all of-the steam was condensed, heating the pool.
Core degradation began in tne upper central portion of the core, . The first fission products were released from the fuel rod gaps at 1,26 hours, the first core material relocated at 1.46 hours and the first core channel-I blockage occurred at 1.80 hours. As the core meltdown progressed, molten and solid debris accumulated on the core plate, heating the plate resulting in core plate failure at 2,54 hours, The core debris fell into the nearly full lower plenum after the core plate failed. Large quantities of steam were generated as the! debris partially quenched, The water in the lower plenum ' completely boiled away ;42,6 l minutes after che core plate failed. The _ partially quenched -hot debris resided on the lower head, heating-the lower head and penetrations. The inner penetrations reached their failure - ' temperature = at 4.16 hours. Then, the core debris continued to heat and, when the required molten conditions - for pouring . were . reached, the first debris _ pour ~ from the vessel occurred - at 4.52 hours . This : molten . core debris began attacking the pedestal . concrete producing -gases _ which contributed to' the pressurization of the containment. Intermittent debris' pours continued until 7,62 hours. 5-127
- a. . .z -. . - - , - ,_ -. --- -. . : - --
The containment exceeded 60 psig (the venting pressure criteria) at 8.40 hours and continued to pressurize reaching pressures of 85 (ADS failure pressure), 145 (lower limit of containment failure pressure), and 195 psig (median containment failure pressure) at 9.22, 14.23, and 21.25 hours, respectively. When the wetwell reached 195 psig, the containment failed with a rupture in the wetvell vall, followed by rapid containment-depressurization. The suppression pool remained subcooled throughout the calculation. Flow from the containment depressurization enteted the reactor building in the upper Unit 2 volume and several deflagrations occurred in the reactor building during the calculation. At the end of the calculation, there was insufficient hyJrogen or carbon monoxide to meet the ignition criteria and further deflagrations were unlikely. l l 5.2.1.2 Primary System liydrodynamics ) l The ADS was actuated upon the initiation of the accident to begin a slow ) The primary ~ system pressure is-depressurization of the primary system. shown in Figure 5.2 1. The safety / relief valves cycled open and c.losed once before the ADS interceded. Initially, the ADS was operated gradually, such that, the pressure was reduced at a rate corresponding to a reduction in the saturation temperature of 100 oF per hour. According to the ADS operuting procedures, all ADS val.ves were opened at about $2.5 atuuter. When . the core water level dropped below the -275 inch . level and the vessel pressure waa less than 700 psig at the time. The primary system rapidly depressurized af ter the valves opened. A_ pressure spike of 3.8 MPa (551 psia) occurred after core plate failure (2.54 hours) caused - by the rapid heat transfer and steam generation in the lower plenum when the molten core debris contacted the water. The reactor vessel water levels are shewn in Figure 5.2-2. The water level in the core (collapsed) dropped below the top of_the active fuel at 26.6 minutes and the exposed fuel began to heat. The rapid depressurization following the opening of all ADS valves caused the water in the vessel to flash and the levels in Figure 5.2-2 reflect this flashing at $2.5 minutes, The core volumes emptied of water at about 1.1 hours; The primary system integrated flow rates - are shown in Figures 5.2-3 and 5.2-4 for steam and hydrogen flows, respectively. 'The general direction of flow was, from the lower plenum, up through the core and downcomer, through the upper plenum, separators and dome, to the ADS valves. 5.2.1.3 Cora Meltdown I The physical - destruction of' the - reactor core' began in- the upper central l portion of the core at 1.26 hours with the failure of.the f uel rod cladding to contain trapped fission product gases located in the fuel rod gaps. The 5-120 a= a - = ~ _ . . = . - - .- = .. _: = --
first core material relocated at i,46 hours and first core channel blockage occurred at 1.80 hours. ..s the core meltam ,rogressed, molten and solid debris accumulated on the core plate, heatir.g too niate resulting in core plate failure at 2.54 hours. The metal oxidation and debris ejection processes are illustrated in Figure 5.2 5 which shows the total masses of zirconium, zirconium oxide, steel, end steel oxide, as well as the 9 ass of hydrogen produced in the prirary system. A hydrogen mass of 373.1 kg was produced in vessel by the time of reactor vessel failure and 577.5 kg by the end of the calculation. The percentage of hydrogen, produced in vessel after vessel failure occurred, was 35.4%. For comparison, the low pressure short-term station blackout calculation produced roughly half as much in vessel hydrogen as the hi&h pressure calculation (refer to Section 5.1.1.3). The lower steam densities associated with the low pressure calculation, resulted in the in core oxidation process being much more limited by steam depletion, than occurred in the higher pressure calculation. The low pressure core meltdown occurred at lower temperatures-than the-high pressure meltdown. In the high pressure calculation, some fuel in the center of the core debria. had begun to melt (3113 K) but in tho' low pressure calculation, the debris temperatures peaked 280 K below the fuel-melting temperature. This difference in peak temperatures in related to-the differences in their exothermic energies associated with the metal oxidation processes - and to pressure t nsitivity heat transfer parameters (e.g., saturation temperatures and latent heats of vaporization). 5.2.1.4 Containment Hydrodynamics The containment contained the released fission products until it. failed due to overpressurization caused by the addition of steem, non condensible gases, and heat. The containment pressures are shown in Figure 5.2 6. + Initially, the pressure increased primarily due to the steam and hydrogen gases released.from the primary system through the ADS into the suppression pool. The centainment pressurization continued after primary system depressurization due to the boiling of water :in the raactor pedestal-volumes and the production of steam and non condensable gases from . the core / concrete interaction and-concrete degassing. Containment-failure was predicted at 21.3 hours with a rupture through the wetwell wall.- Fission products were transported from the' containment into the . reactor building, following failure, first by the depressurization flows and then by the flow of gases produced continuously by' oxidation and concrete Jecomposition. The atmosphere and - surface temperatures of the drYwell and wetwell are shown in Figures 5.2 7 and 5.2-8, respectively, The pedestal volume tamperatures are shown in Figure 5.2 9. The highest containment tempe ra ture s , peaking- at about " 1900 K, were i calculated in the wetwell pedestal volume where the concrete was ablated by b 9
,_. _:_-__--_-_=__
. . ~ . - . - - - - -- . . - -- . _ - - . . . -- .-
molten core debris. The debris in the wetwell pedestal was initially cooled by over laying water, which lasted about 30 minutes after the debris passed through the cavity floor. The wetwil wall surface temperatures exceeded the melting temperature of steel and the temperature needed for ablating concrete. The vetwell pedestal temperatures remained high during the remainder of the calculation and the surrounding concrete was heated, releasing large quantities of steam and carbon dioxide gases into the contain'nent. (The fluctuations seen in this figure were due to additions of water from the drywell pedestal.) Steam condensation within the containment is illustrated in Figure 5.2 10. The condensation mass fluxes for two d;jvell surfaces and one vetwell ; surface are shown, along with, the total drywell condensate mass (as measured by the integrated flow through the drywell drains). The most j rapid condensation occurred in response to the boiling of the water in l pedestal by the core debris. The condensate in the drywell acetunulated on ' the drywell floor and then drained into the pedestal where it was revaporized. The wetwell condensate entered the suppression pool. The water masses for pools that accumulated in the drywell, drywell pedestal, and wetwell pedestal are shown in Figure 5.2-11. Condensate and pump seal leakage that collected on the drywell floor flowed into the pedestal through the floor drains. The drywell floor drainage and leakage from the control rod drives accumulated in the drywell pedestal until the pedestal floor was failed by core debris, af ter which, drywell pedestal water flowed down into the wetwell pedestal. A countercurrent flow situation existed between the drywell an* wetwell pedestal regions with water flowing downwards and steam and non ,,adensible gases flowing upwards through two four-inch diameter holes. The fluctuations. seen in Figure 5.2-11, were the result of instabilities in the countercurrent flow and resources did not permit a detailed enough analysis to determine whether these instabilities are physically realistic or an artifact of the ntunerical solution. The suppression pool mass. increase and temperature are shown in Figure 5.2-12. The pool maas increased due to the condensation of steam from the ADS, to condensation on the heat structures and the pool surface, and to condensing steam flows through the vent downcomers as the containment depressurized after it ruptured. The largest increases occurred following the rapid depressurization of.the primary after all ADS valves opened and following core plate failure when molten debris interacted with the water in the reactor vesst i lower plenum. The-suppression pool continued to heat due to the addition of warmer condensate and the decay heat - from - fission products deposited within the pool. The downcomer and wetvell suppression pool water levels are shown in Figure 5.2 13. The downcomers do not clear at vessel breach as they did in the high pressure calculation because the breach occurred at low pressure, liowever, the downcomers did clear at containment rupture, unlike the high-pressure case in-which the containment failed with'a leak. 5-130
. _- _ _ _ _ _ _ - . __ ;_ _u_ . a _ _
_ _ _ _ . _ _ . _ ~ . . . _ _ _ - _ . _ . _ _ _ - _ . _ _ .. . _ _ _ . _ _ _ _ The first molten core debris poured from the reactor vessel lovar head into the drywell pedestal at 4.16 hours and consisted of liquid steel along with some steel oxide. Twenty minutes later the drywell pedestal floot failed, i as modeled, and the debris transferred completely to the wetwell pedestal (refer to the model. description in Section 3.1.2 and to the discussion of S results in Section 5.1.1.4). The masses and temperatures of this steel and steel oxide are shown in Figure 5.2-14. The debris masses for the wetwell pedestal core / concrete interaction calculation are shown in Figure 5.2 15. The first debris arrived at 4.49 hours consisting of steel and steel oxide and the first fuel arrived a short time later. The debris temperatures are shown in Figure _ 5.2 16. During the period of rapid zirconium oxidation, the temperature of the ; heavy or. ides peaked at 2486 K. The maximum radial and axial molten pool penetration distances into the concrete structure are shown in Figure 5.2-17. The end - of calculation penetration distances were 1.71 m axially and 1.83 m radially. The radial penetration distance exceeded the pedestal wall thickness of 1.47 m at 17.1 hours (4.2 hours before the _ containment ruptured). The - possibilities of-the structural failure of the pedestal wall and interaction between the molten debris and the suppression pool are shown. Steam generated by quenching heac transfer between molten core debris and the pool could increase the containment pressurization rate causing an - earlier failure, The base case did not model this interaction and. continued beyond pedestal wall failure as if it did not happen. However,_a sensitivity calculation was run (rafer to Section 5.2.2) which involved such an interaction. Carbon dioxide, carbon _ monoxide, hydrogen, and steam were produced as by-products of the concrete ablation and the resulting chemistry of the molten-pool, and were major contributors to the pressurization of the containment. The integrated masses of these gases are shown in Figure 5.2-18. The end of calculation totals were 43760, - 20740, -9660, and 1045 kg for_ carbon dioxide, carbon monoxide, steam, and hydrogen, respectively. The in-vessel and ex vessel hydrogen production are compared in : Figure 'I 5.2-19. A total of 1623 kg was produced (578 kg in vessel and 1045 kg ex-vessel). The low and high pressure-calculations produced almost the same l amount of ex-vessel hydrogen. The combined integrated mass and' release _ rate of steam and carbon . dioxide gases released from heated concrete by dehydratio' n are shown in Figure i 5.2 20. The end of calculation total' mass of 44360 kg is comparable to the corresponding mass in the hi6h pressure calculation. The mass and mole fraction distribution of hydrogen within the containment-
- is shown in Figure .5.2 21 and the integrated hydrogen flow rate s illustratr/ in Figure 5.2 22.
l 3-131
_. . _ - - . - . - - - - - - .._- - . - _ _ - .~ - . - -- .- . - - - 5.2.1.5 Reactor 11uilding Hydrodynamics The containment friled with a rupture in the wetwell vall at 21.2 hours (3 hours later than the high pressure calculation). The c ont ai ntne nt depressurization total and hydrogen _ flow rates are shown in Figure 5.2 23. 4 The hydrogen mole fraction of the depressurization flow is also shown. A series of deflagration burns occurred in the upper Unit 2 at.d refueling bay volumes as the containment depressurized. One burn propagated into Unit 1 but sufficient hydrogen did not accumulate in the s te arn tunnel / turbine building volumes to burn. The reactor building combustion dynarnic s are illustrated by: (1) the atinospheric temperatures shown in Figure 5.2-24; (2) the peak pressures and temperatures listed in Table 5.2-2: (3) the mole fractions for the gases in upper Unit 2 shown in Figure 5.2-25; and (4) the hydrogen mole fractions for each volume shown in Figure 5.2-26. The hydrogen and carbon monoxide mole fractions increased prior to each burn and then sharply decrease - as they were consumed, being replaced by combustion products. Oxygen depletion was seen but the oxygen was replenished by natural circulation flow within- the reactor building and with the environment. 5.2.1.6 Radionuclide Transport The transport of fission products throughout the plant and the resulting source terms to the environment are illustrated by a summary of the end of calculation results presented in Tables 5.2 3 through 5.2 9 and by examples of time dependent esults shown 'in Figure 5.2-27 through 5.2 39. The information displayed in similar to that of the high pressure calculation and the uncertainties discussed in the high pressure calculation apply to the low pressure calculation as well. The first release of fission products from the cladding gaps in the low' pressure calculation occurred at almost the same time (1.26 hours) as in the high pressure calculation. However, differences are present in the calculated releases of fission products between the low and high pressure calculations: for example, the cerium releases shown in Figures 5.2-27 and 5.1-48 for the low and high pressure calculations, respectively. . i'he initial releases of ceriu.n in the low pressure calculation were smaller and occurred later than those in the high pressure calculation, due its lower-core temperatures (discussed above). To further illustrate,: tho ' cerium rate of release at.2700 K, as calculated from the CORSOR M' correlation, is less than 1% of the rate at the fuel melting _ temperature. The ex-vessel releases of cerium, and non radioactive structural aerosols for the low pressure calculation, __ shown in Figure 5.2 28, are similar'to those of the high pressure calculation. The primary _ system retention factors are _, listed in Table 5.2 3 and_shown for selected classes in Figure 5.2-29. Since vapors more readily-transport-with gas flows _than acrosoir, and the high temperatures . vaporize more of 5-132
, a . . z . . - ..a . ~~ . -
the lower vapor pressure fission products classes, more of these classes transported from the primary system in the high pressure calculation than the low pressure one. The retentions of the aerosol classes are similar for the two calculations. An abrupt decrease occurred in the retention factors for the iodine and cadmium cle.sses when the at containment ruptured and rapidly depressurized. The final retention factor for the molybdenum c' is is listed in Table 5.2 3 as 1.13. This implies that more fissior products of thir. class entered the primary system than wore released within the primary system (the ratio of in vessel to ex vessel releases was only .00652) and may not be realistic because of the reduced in vessel release rate of this class (refer to Section 5.1.1.6). The retention factor'for uranium could not be calculated because uraatum was not released in vessel. The fractions of the total fission products released from the fuel that l remained in the containment (excluding the primary system) are listed in Table. 5.2-8. The distribution of these fission products within the containment is also given. For example, 43.14 of all cesium released from the fuel (and not - chemically co.nbined with iodino) was located in - the - containmont at the end of the calculation. Of this quantity of cesium, the majority, 83.6%, was deposited in the suppression pool leaving 8.3 and 8.1% in the atmospheres or on surfaces in the drywell and wetwell, respectively. The time dependent mass distributions for the xenon and cesium classes are shown in Figures 5.2-30 and 5.2 31, respectively. The xenon class gases readily transport from the primary system and containment with the majority residing in either the teactor building or environment at the end of the calculation. Virtually all of the xenon class would eventually escape to the environment, where as, the majority of the cesium class deposited either on surfaces or-in the suppression pool resulting in only a small 7 amount being released to the environment. The fission product tource terms-to the environment are shown in a series of seven figures, Figura 5.2 32 through 5.2 38. The masses escaping the containment are shown along with the masses entering the' environment. Some retention by the reactor and turbine buildings occurred for all classes. The reactor building decontamination factors, which are shown for selected classes in Figure 5.2-39, demonstrate the transient nature of-the release and transport of fission products. l l 5.2.2 Pedestal Wall Failure Sensitivity Calculation A sensitivity calculation -was run to examine the sensitivity of the containment pressurization to the possible ' interaction -of the molten core debris and the suppression pool. The sensitivity calculation was initiated at the time (17.1 hours) when the lower pedestal wall was predicted to fail (i.e., when the radial concrete ablation exceeded - the initial wall l thickness). In this calculation, suppression pool water was allowed to 5-133
flow into the pedestc.1 and forrn a boiling pool over the molten debris. The extra steam generated in the sensitivity calculation caused the contairunent pressure to reach its failure pressure 3.8 hours earlier than in the base case, as shown in Figure 5.2 40. The sensitivity calculation resulted in an earlier release of fission products to the envirotunent. The ratios of the fission product masses in the environment (the sensitivity calculation masses divided by the base case calculation masses at the end of calculations) are listed in Table 5.2 9. The releases were enhanced for - some - classes and retarded for others. The enhanced cesium source term, for-example, was related to-some revaporization of deposited cesium near the time of pedestal wall failure and to the enhanced containment depressurization flows. The boiling pool in the sensitivity calculation cooled the debris sufficiently to essentially stop further releases of certain classes,- such as telluritun and-tin, thereby, retarding their environmental source terms. d 5-134 __-___- _ - _ _ _ _ - _ - _ - _ = - _ _ - - _ _ _ -. _ - -
i i Table 5.2+1. 4 Low Pressure Short-Tetw Station Mackout Events ! i Fvent fame Fvent Descritt.on 3etende Minviej Hws s
- 0. O. D. Accident Inittsted, Reactor Tripped, Isolated t 1$98, 26.6 0.44 Co11ersed Water Level Below 20p Active Fuel 4$32. 7$.$ 1.26 Fission Products Released frce Ring 1 Rod Gap (First Fission Products Released) 4786 70.8 1.33 Fission Products Released from Rino 2 Rod Gap :
$2$0 87.$ 1.46 First Core Material Relocated [
$464 91.1 1.$2 Fission Products Released from Rin,a 3 Rod Gap '
6482. 108.0 1.80 Core Channel in Rios 1 Blocked , 6964, 116.1 1.93 Fission Products Released frcro Ring 4 Rod fi p i 7351. 122.$ 2.04 Cere Channel in Rins 2 Blocked , 9134 152.2 2.$4 Core Plate in Ribs 1 Failed Core Channel in Ring 3 blocked I 9139. 1$2.3 2.$4 11690. 194.0 3.2$ Lower Plenum Dried out 14786. 246.4 4.11 Core Plate in Rins 2 Fetted , 14977. 249,6 4.16 Lower Head Penetration in Rins 2 Failed (Reactor Vessel Fa!!ed) Fir 6t Debrie Ejected from RV to DW Pedestal ^ 1$711, 261,8 4.36 1.ower llead Penetration in Ring i Failed 15091. 268.2 4.47 Core Plate in Ring 3 Failed i 16176. 269,6 4.49 Drywell Pedestal Floor Droin Pipes Failed 16181. 269.7 4.50 Wetwell Fede6 at Received Core Debria 7 l 16253, 270.9 4.$2 Debris Ejected l Lower Head Penetration in Ring 3 Teiled , { 16684, 278.1 4.64 Detris Ejected 17907. 298.4 4.97 Debris Ejected 197(1. 329.0 $.48 Debris Ejected 20089. 334.8 S.58 Debris Ejected I , 20303. 336.4 $ 64 Debris Ejected 20408. 340.1 $.67 Debris Ejected i l 20411. 340.2 $.67 Debris Ejected i 20470. 341.3 $ 69 Debris Ejected 2074$. 345.7 $.76 Core Plate in Ring 4 Failed l 21179. 3$3.0 $.88 Debris Ejected . 21373. 3$6.2 $ 94 Lower Head Penetration in Ring 4 Failed !' 21418c $$7.0 $.9$ Debris Ejected [ 21554, 359.2 $.99 Debrie Ejected 21719. 362.0 6.03 Debris Ejected 27439. 4$7.3 7.62 Debris Ejected 27443. 4$7.4 7.62 Debris Ejected 30231. 503.8 8.40 contairnent Wetwell Pressure Exceeded to pais ? 33208. $$3.$ 9.22 containment Wetwell Pressure Exceeded 85 pols
$1212. B$3.$ 14.23- Containment Wetwell Pressure Exceeded 14$ psi 8 61438. 1024.0.17.07 Wetwell Pedestal Well Penetrated 7649$. 1274.9. 21.2$ Containment Failed with Wetwell Break 765$4 127$.9 21.27 - Deflagration Earn in Upper Reacter Bid. Unit 2 76586. 1276.4 21.27 Deflagration Burn in Upper Reactor Bid. Unit 2 76622. 1277.0 .1.28 Deflagration Burn in Upper Reactor. BidJ Unit 2
( 76861. 1281.0' 21.3$ Deflestation Burn in Refuelins Bay 1 76863, 1281.1 21.3$ Deflagration Burn in Lower Reactor Bld. Unit 2-7703$. 1283,9 21.40 Deflagration Burn in Refueling Bay - i 77309, 1288 $ 21.48 Deflagration Burn in Refueling Bay 77321, 1288.7. 21.48 Deflagration Burn in Lower Reactor Bld. Unit 2 77322. 12f8.7 21 48 Burn Propagated into Refueling Bay and Unit 1 788$4, 1314.2 21.90 Deflagration Burn in Lower Reactor Bid. W.it 2
.86003, 1433.4 23.89 Calculation Terminated
+
S 135 1 l.
.,,._.._4._.#.,_~,_,_ ,, , , ,o _ m. _ , _ , . . , _ _ __ , _ , . ,,.,_..,-.m. ...%._.. . . _ . - -,,
Table 5.2-2. Reactor / Turbine Building Peak Pressurts and Temperatures for Low Pressure Short Term Station Blackout Location Differential Eressure Temocrature Pa Psi K F Lower Unit 2 38900 5.6 1131 1576 Upper Unit 2 39000 5.7 1363 1993 Unit 1 38600 5.6 818 1013 Refueling Bay 34600 5.0 1333 1940 Steam Tunnel / Turbine 2 8110 1.2 330. 134 Steam Tunnel / Turbine 1 7170 1.0 326 128 Table 5.2 3. Summary of Radionuclide Releases and Transport for Low Pressure Short-Term Station Blackout > Initial Fraction Ratio of Ratio of Fraction Reactor Bid. Class Core Released In to Ex Total to Retained Decontamination NO. Rep. Inventory
- from Fuel Vessel Radioactive by Primary Factor (KS )
1 Xe 463.7 1.00 34.6 1. 1.4E-5 1.7 2 Cs 268.4 1.00 3.48 1.13 .70 3,7 3 Ba 207.5 .480 .233 1. .82 2.8 4 1 20.93 1.00 34.1 1. .74 3.7 l 5 Te 40.79 .975 41.1 1.12 .70 6.0 6 Ru 307.0 2.79E 5 829, 317. .61 5.2 7 Mo 350.6 .0166 6.52E 3 318, 1.13 6.3 , 8 Ce 594.0 1.20E 5 .340 94.7 73 4.7 ! 9 La 571.1 8.41E 3 .257 1. .63 -6.6 ! 10 0 132390, 3.21E 5 0. 1.13 - 6.3 , 11 Cd 1.407 .336 1390. 1, .38 4.5 l 12 Sn 8.587 .344 43.9 1. .62 8.7 16 Cs1 0, .76 6.4
- l i
- Radioactive Fission Products (except Uranium) i k
l v I 5-136-L L
- -._ - . -._ .. _ .- - .~ .- - i I f Table 5.2 4. j Fission Product Distribution (kg), Classes 1 4 ! locati2D .Radionuclide Class ! Xe(1) Cs(2) Ba(3) 1(4) Releases from Fuel Primary 4.507E402 1.872E402 1.879E+01 1.088E 04 i Cavity 1.302E+01 5.926E401- 8.079E+01 0. ! Overall Distribution . Primary 6.291E 05 1.303E402 1. 539 E401 8.088E 05. ! i Containment 9.737E+00- 1.063E402 7 955E+01 2.793E 05 i Reactor Bld 1.652E+02 7.066E 01 1.054E400 3.353E 14 , Turbine'Bld 2.086E+01 2.771E 02 5.300E 02 8.277E 16 i Environment 2.679E+02- 2.708E 01 6.310E 01 1.255E-14 Containment Distribution . Drywell 8.528E+00 6.332E+00 4.520E+00- 1.7162-10 Upper Pedestal 1.369E 01 .2.497E400 1 149E400 2.626E 05 + Lower Pedestal 1.420E 02 3.360E 04 3.431E+01 8.104E 28 Wetwell 1.058E+00 9. 745E401 3.958E+01 1.665E 06 Suppression Pool 0, 8.888E401 3.389E+01 1.570E 06
. Table 5.2-5. .
1 Fission Product Distribution (kg). Classes 5 8 ! i Location Radionuclide Class Te(5)- Ru(6) -Mo(7) -Ce(8) Releases fr,m Fuel Primary 3.881E401' 8.563E-03 3.763E-02 1.810E 03 Cavity '9.432E 01 1.033E 05 5.773E+00 S.327E 03 Overall Distribution l Primary. 2.712E401 5.218E 03 4.268E 02 .1.328E 03-Containment 1.103E+01 3.005E 03 5.387E400 5.192E-03 i Reactor Bld 2.074E 01 5.543E 10 2.709E 04 2.407E 04 L -Turbine Bld 3.090E-03 9.089E-12 3.404E 06 4.540E 06 L Environment 4.176E 02 1.338E 10 S.189E 05 6.597E-05 ' Containment Distribution Drywell 5 14SE 01 9.427E 04; 2.384E 01 5.201E 04' .: Upper Pedestal 2.247E 01 4.997E 04 5.811E 02 '2.683E 041 ' Lower Pedestal 6.744E 04 1.566E 041 1~.866E+00 ~2.433E 03- - Wetwell '1.029E+01. 1.406E 03- 3.225E+00 1.971E 03 3 Suppression Pool -1.006E+01' 1.300E 03 >2.904E400 1.662E _ b L 5-137
-;..m._,.a._.-..,. _. .~ . a. - _~ _ _ . - -- - - . . . _ - ..;__-- ._ .- .-.._ _ _ ..._;....a--...
Table 5.2 6. Fission Product Distribution (kg), Classes 9 12 Location Radionuclide flagg 1.a(9) U(10) Cd(11) Sn(12) Releases from Fuel Priraary 9.819E 01 0. 4.729E 01 2.887E400 Cavity 3. 820E400 4.252E400 . 3.399E 04 6.583E 02 Overall Distribution Prinary 6.171E 01 1.278E 02 1.778E 01 1.801E400 Containment 3.917E400 3.869E+00 2.412E 01 1.067E+00 Reactor Bld 3.051E 04 9.775E 02 3.667E-02 9.088E 03 Turbine Bld 3.520E-06 1.202E 03 5.736E 04 . 7.411E 05 Envirnwnent 5.519E 05 1.863E 02 1.063E 02 1.189E 03 Contaimnent Distribution Drywell 3,214E 01 2.278E 01 2.467E 02 2.536E 01 Upper Pedestal 8.3$5E 02 4.860E 02 1.017E 01 -1.222E 01 Lower Pedestal 1.702E400 1.766E400 1.613E 05 8.546E 02 Wetwell 1.810E+00 1.826E400 1.148E 01 6.060E 01 Suppression Pool 1.522E+00 1.548E+00 6.690E 02 5.676E 01 Table 5.2 7. Fission Product Distribution (kg), Class 16 , I Location Radionuclide Class
- CsI(16) .
Releases from Fuel Primary 4.163E+91 Cavity 1.220E+00 Overall Distribution Primary 3.176E+01 Containment 9. 732 E400 Reactor Bld 5.296E 01 Turbine Bld 6.871E-03 Environment 9,978E 02 Containment Distribution
- Drywell_ -3.902E 01 Upper Pedestal -1.856E 01 Lower Pedestal - 6.168E 04 Wetwell 9.155E+00 Suppression Pool 8.835E+00 l _5-138:
I 1
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l I Table 5,2 8. Containment Distribution of Fission Products
.J Containment Location within Containment Class Fraction of (Fraction of containment Mass)
No. Ren. Total Release Drvvell Wetvel1~ Eyporession Pool 1 Xe .021 .890 .110 .0 2 Cs .431 .083 .917 .836 3 Ba .799 .071 .929 .426 , 4 1 257 .940 .060 .056 i 5 Te .278 .067 .933 .912-6 Ru .350 .480 .520 433 7 Mo .927- .055 .945' .539 . 8 Ce .727 .152. .848- .320 i l 9 La .816 .103' ,897 .389 i l 10 U .910 .071 .929 .400 ~ ! 11 Cd .510 .524 476 .277 ~ 12 Sn .361 .352 .648 .532 i 16 Cs1 .227 .059 ,941 .908 . Table 5.2 9. Environmental Release Comparison Class Ratio of Pedestal Wall Failure No. Rep. Sensitivity to Base i'ase , 1 Xe 1.13-2 Cs 4.01 3 Ba l'.01 I 4 SE 11 5 Te 0.82 " 6 Ru 2.13-7 ~Mo 3.50 4 8 Ce 1.57 ' 9' La . 1.39 ' 10 U l'.29 ' 11 Cd 2.52 l 12 Sn 0.85 16 Cs1 0.85 I 5-139' _ - _ . ._. _ . __ _ .-.~. .. _ _.... . _ ._ _ . . - . _ . - - ... _ ., _ _ _ - . _ - .
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0.0 3.5 7.0 10.5 14.0. 21.0 24.A 28.0 31,6 35.0 TIME (103) s Fig,ure 5.2-5 Total Core Masses
~
5-144-
.. - - - . ... . - - ~..... .~ - .. -- - - ..-. .. . . . - , -
d a 1 e 7 J B + 1 a i i 15-- , , , r- i i , , .- a
- 1.4 Drywell .
,- coo.o
^ DW Steam Wetwell 1.3 -
.+. .WW Steam -
1 I L2- -- 175 0 1 t 1.1 - . 4
~
1.0 - , i n g 00- .
- 125.0 -
e-gO oa-i - - ?.- a y $
~
. :s 0.7 - '- 1001 m m i o-- b 0.6 - , 0.5 -- _- 75.0. 0 4 -- - - i r
- 60.0 l- 0.3 - ,n,
, ; 4, 0.2 - -
- - + ,l; . '+... -;
. ., _ - 25 0 ;
I ' 0.1 - - i.-
. LASALLE MELCOR LOW PRESSURE
, __ SHORT TERM STATION BLACKOUT
- 0. 0 -04
- i- ,- , , , , , ,
aa
- 0.0 st 0.0 20.0 30.0 40.0 . 60.0 60.0
- 70 0 40.0 90 0
.; .' T1ME (10 3)=s l: f i , Figure 5.2-6 Containment Pressures-i- 5-145 i~
. . . - . , . . .. . .. . .n . - . - . .- .. - _ . . . . . - . ., - - - - - - . n. . . _ ~ . - ,
t i i i~ 1. i ! 500 0 i , , , , , , , I LASALLE WELCOR LOW PRESSURE - 425 o
- . Sit 0RT TERM STATION BLACKOUT 400.0-
, -- 400.0 e ' - 375.0 480 0- - ---- j # <
- 350.0 l' 440.0- - -
i' '
- 3260 b .
/^ -
s ,- 300 0 s ,' <-N y =
~
,. 275 0 __
j
-[g ,
- C.
w , , ,. - 250.0 t
,' -- 226.0
- l. . ' /4 fg -200.0 a
. j, i 360.0- ,
_ g}' -
..-* $ - ns.o 340.0 -
j l- - Atmosphere =
- - - Liner :
. -_ . 3 3o.o 4
i S V. -
--- Floor -
Reactor Shield = - 885 0 : sao o- - Misc. Steel - - 200 0-2 t- 1
- 300.0 , , , , , , , .,
0.0 10.0 20.0 -30.0 '40.0 50.0 .60.0 . 70 0 : 80.0 '. 90.0 - l TIMP. (103) s r I ~ 5 i 4 3 J Figure 5.2-7 Drywell Temperatures 2. 5-146-v ev -
-N
-w ww re".- y-m+--+
-~
$25 0 , , , , , , , ,
LASALLE MELCOH LOW PRESSURE Sil0RT TERW STATION BLACKOUT
- 450.0 600 0- --
476.0 - -
. **i
- .{., ...
450.0 - .' -- 350.0 4' '. g : a
~
v .-' g 425.0-l -- 3000 4, l '... '. 8 e l .. C A 400.0- x v
.s l *g 0 l ,g -
's.)
y ;
't' ' ' % . ~. . , ,l % . s - ' 260.0
(.
; . i 376.0 - ,I , . l -
*/
- . -- 200.0
,. / _____ -
350.0-
,. --:7
'.I -
'/-;
I
. - 1s0.0 325 0- [ . ,- ' ./. '
.- Atmosphere -
--- Sunnression Pool -
yff~
- Misc. Steel - - sco o 300.0--. , , , , , . - r- i 0.0 - 10.0 - 20.0 30.0 - 40.0 50.0 60.0 70.0 80.0 90.0 TIME (100) s Figure 5.2-8 Wetwell Temperatures 5-147
22-- - , r , , , , , ,
-X- DW Pedestal Atmosphere WW Pedestal Atmosphere us 2 0 -- X- DW Ped Wall Surface -
-+ WW Ped Wall Surface _ 3'no
/p 1.B ' ,
WWPed WW Ped WallWall InteriorInterior (.39 M (.11 - M))
~
2.76
.' k 1l)
.G
, ~'~
f ("1" f'~L b 2 s0 2.26 2 ' O -
" 2.00 g .
,_ g
"" ~
0 = h - (' I s-C 8 i j * -- 1.75 m g 1.2 - v i
/
o
} J e
*^
U l' _/,(' 1.50
- a. t -
E i.0- -
@ ,' s
- 1.2 b
- 60 04- jd '
-- 0.25 gg.pyy,73 - ,- LASALLE MELCOR LOW PRESSURE SilORT TERM STATION ULACKOUT
- 0.00 02 , , -y i , , , ,
0.0 10 0 20.0 30.0 40.0 60 0 80.0 70.0 80.0 90.0 TIME (103) s Figure 5.2-9 Pedestal Temperatures 5-148
_ _ _ _ . . .__m. . . - - - . . . . . . _ . _ . - . . . . _ , ._ _ . . _ . . . . . . , _ _ . L
~
20 , , , , , , , , wu
.- LASALLE WELCOR LOW PRESSURE
- - - Drywell Liner Flux Sil0RT TERW STATION BLACKOUT
-- - Drywell Floor Flux
.8 - ~
WW Wall Flux -**"
--- 'Ibtal DW Condensate 1.6 - -
- 30.0 1.4 - -
R
- as'o ;
n e .. , f. ~ ~ ~ - -
- x. / Q
- w
/ .-
n, i n o 1,0- / -- 20.0 O' C / Im n !
~1
/ 40 0 0~8- -
( n / m / - 15.0
& /
/
06- , p - \ i
- ic.o f:
0.4 - (
*j .
,- . r::
f. l1 , ,
- p O
- - 5.0 0.2 - -i - ' . . -
s ..,, ... ,, s g,' ,g_ ;) '
. / '
- i. -~ ; _ [~ - -
0.0 i . aa 01 10.0 20.0 30.0 40.0 50.0 80.0 70.0 . 800- 90.0
' TIME (10 3 s)-
Figure 5.2-10 containment Condensation .l 1 5-149
~
3 0 ... _7___,...__.,___y__.__7 ____-.._.~._m _- . . _ . . .
, , , r --
l
-- Drewell Pedestal 1 AsAm mron t.cw PanssuRE Wot.well Pedestal Sil0RT TERM STATION Dl/CKOUT 46- - - - Drywe}} 1u.0 4o. 9o
-R0 35-70 30-
/ ^
tL
/ -
g* Z
- 60 h
0 M
- 2. 5 -
f
$ 50 f
$ 20- \ e 4.0 1.6 --
I 30 1 0 -- -
, j C5-
- e .
-
- 10 1 l 4- ,.
O 0 , -, - - t - -- [ S p[ ~ _ ,-- p>_____ & 7.. , an 0.0 80 16.0 24.0 32 0 40 0 48 0 66.0 64 0 72.0 80.0 TIME (103) s 4 s Figure 5.2-11 Drywell and Pedestal Water Masses 5-150
400.0 , , , , , ,uu.u LASALLE MELCOR LOW PRESSURE SilORT TERM STATION DI.ACXOUT 360.0-
- 390.0 320.0-
-- 380.0 e
280.0-
-- 370.0 f ,,......
240 0- f. ..******' , 'a x- -
-- 360 0 - 2 m
m " 200.0- [ ,
-- 350.0 M e u m .. .' O a
$- 160.0- 6
-- 340.0 - #.
120.0-
-- 330.0 a 0.0 - ,
Mass Increase -
- 3c0.0 Temperature -
40.0-
._ 330.0 Initial Water Mass
.of 3,682.400 Kg 0.0 . . , , aaa a 0.0 10.0 20.0 30.0- 40.0 60 0 60.0 .70,0 80.0 90.0 TIME (100) s t
a Figure 5.2-12 Suppression Pool Mass and Temperature 5-151=
5 ?
-15 0 , , , , , ,- , ,
LASALL.E MEILOR LOW PitESSURE SHORT TERW STATION DLACXOUT Port Bottom Elevetion
-16.7 - - - - - _ . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
- 705 0
~
- 16.4 - li
. j_ _ {_ = ., ,-rrrA ' ' .- - 102.6 4
- 17.1 - ,'
,. [ ,if f ~
~ 700.0
- tv.e - -
v b 3 - 697.6 h s g -le 6- - g
> Downcomers e 3 f ------ Wctwell _ eg5 o C a
^
l -19 2- - h 5
- 692.5
-19.9 - -
e 5
- 690.0
-2 0.6 - -
Downcomer Exit
-21.3 - --6R5
-22.0 , , , , , , , ,
- ARA D
- 0. 0 10.0 20.0 30.0 40.0 60.0 60.0 70.0 80.0 90.0
- TIME (100) s Figure 5.2-13 Suppression Pool Water Levels 5-152
00 i , , , , , , , , .. u u LASALLE MELCOR LOW PRESSURE : Light Oxide Mass SHORT TERW STATI'JN BLACKOUT -X- Metal Mass 7.2 - _{. . Light Oxide Temperattire
)
-X- - Metal Temocrature ~ "5 3 M
84- f' e,'.
- 1.60
.'X h.6 - .' -
- +
- 1.25 2 a
4 .a -- N
+ , '.,.
n x ' S
$ 4.o-- ; __ .co [;
E l 4 3e e - 2: X' '
. a i
3.2 -- + . E l l ., - 0.75 @ l : X.*
- 2. 4 - . .
4
. 010 j X '
te- *-
-X- - + ' - - X -t- -
-X 4
o . .. s--X x x- x f 26
. s ,
oo- +X-i% rXh~ , -A-- , I r ', + ,X - F a aa 14 5 14.7 14.9 15.1 15 3 16 6 15.7 15 9 16.1 16.3 16 5 TIME (103) s Figure 5.2-14 Drywell Pedestal Debris Masses and Temperatures 5-153
10 , , i , i , i i 1.ASAI.LE MElf0R LOW PRESSURE Total SHORT TERM STATION BLACKOUT 0.o - --- lleavy Oxide -- "O Light Oxide
- -- Metal 3
08- -- 1,7s i
~
- 1.50 m 06-y ,,,-.* - 1.25
}
w . ~. S o a g 0.5 - e O m m ..
** - 1.00 O
, d .. - 7, 0.4 - -
- 0.75 a
03- - l
! ! - 0.50 02-
~'
A i l o t-l~'-} ~
- C.25
..T'-------------------
.l ,',i ft }
00 a aa = i i - , , , , , , O.0 10.0 P0 0 300 40.0 50.0 60.0 70.0 80.0 .90.0 TIME (103) s 4 Jigure 5.2-15 Wetwell Pedestal Debris Masses 5-154
2.50 , , , , , , , ,
. LASA1.1.E MEICOR LOW PRESSURE SHORT TERM STATION BLACKOUT 2.35- -
- 3.75 Heavy oxide Light Oxide 8 20-
--- Metal f --- Lower Pedestbl Atmosphere ~- 85 f3 i
t 1g e e3_ r -- 325 l i %, 2o "a 1. t '3 i
\,
.- 8 00 i\
s
= i
\ C l 1.rs. -! Il,' g lIe j'\ - -
"o ll - )
,, _\,-
~~ -
E g, r jlt
',l ,' '
e - p/-- p -t s uo. i 11 I r 'I ,/ l/ i -i,
'1 ,r-v
- zs0 fIl g ii ie ll llii-11 'l ;li i, '
_ zes us- i1 1 is li ji ll 1,/ i i I; jl il ,l i, if J s ii in gi ) le _ zo, 1.30-P i 11 'l ,1 ) il
; g ,i lj # }' -
f 'I I ll 'l I - 3.75 t I I
% ti ll I us- l l ll ,l -
ll 1 I 11 ll
- 1.50 lI' I 11 'l 1.00 , ,I d f ,' i , ,
0.0 10.0 20.0 30.0 40.0 60.0 60.0 70 C 80.0 90.0 3's TIME (10 Figure 5,2-16 Wetwell Pedestal Debris Temperatures 5-155
. . - ~ . .. .-. . . . . . . - . .. . .-.
4 f 2.0 , , , , , , , , LASALLE WELCoR LOW PRESSURE SHORT TERM STATION BLACK 0UT
- 1.8 - -- e.0
,# 6.5
, 1.6 - - 6/a Pedestal Wall Thickness
... ... .......... .... ..... .. ..,..........s.
1.4 - e' -- 4.5
^ /
% s
'; W / j
> 8 12-
# -- 4.0 g /
/ - 3.6 Q 1,0 / _ -
U
.j ,
l
- 3.0 i e
b 0 6- 5
$ i - 2.5 e , >
- a. Radial
~
' ' --- Vertical ' O.. I _
.- 2.0 I
- I - 1.6 04- 1 i
, - s.o 0.2 - I .
1 - 0.5 l- /
/
0.0 , ," , , , , , , aa-O.0 10.0 20.0 30.0- 40.0 50.0 60.0 70.0 - 80.0 - 90 0
- TIME (103) a 2
Figure 5.2-17 Maximwn Concrete Penetration Distances 5 156 1 r
80.0 , , , , , , ,_ , LASALLE WELCOR LOW PRESSURE SHORT TERM STATION DLACKOUT - 72.0 - .- 160.0
*1btal Carbon Monoxide e4.0 - - - Carbon Dioxide ._ 3,o c Hydrogen Steam 56.0- -
- 120.0 m 48 0- -
ce b:: - 100.0 . y n 3 40.0- '
'd- ,
- .i n m i . -. o g -. - a0.0 c, x ,'
32.o - , . s
' - 60.0
/
24.0 - ' , i f
,/ * , ..-
40.0 16.0 - ,'- ,,, -
/
s o_ / .-
/
3 , d_ '
- s - - - - *- 20.0 0.0 . , , -4+ , , , , l--- e.
OO 10.0- 20.0 30.0 40.0 60.0 60.0 '/0.0 . 80.0 '90.0 TIME (103) s e Figure 5,2-18 Core / Concrete Interaction Gas Production 5-157
1.1 , , , , , , , ,
,. *,,... - 2.25 1.0 -
3g_ ,- _- 2 00 08- l - s.75 a l 0.7 - / - j - 1.60 n l 2 0.6 - l 1
- 1.25 3
c l -D. O E 0.5 -- I - O e - M [ - 1.00 l o.4 - - l - 0.75 o.3 - . In-Vbssel I' ~ l - - - - " Ex-Vessel
- - 0.60 0.2 - ; -
l l g3_ [ _-025
, LASALLE MELCOR LOW PRESSURE
/ S110RT TERM STAT 10*: DLACKOUT 0.0 , i a aa 00 '10.0 20.0 30.0 40.0 50.0 60.0- 70.0 '80.0 90.0 TIME (103) s s.
Figure 5,2-19 Hydrogen Production
]
9 5-158
I I 60 0 , , , , , , , , s,ww
-- 'Ibtal Mass 54.o - - -- Total Rate - -
- 1.75
.'I I'
48 0- lI 0 e ll
- 1.60
. i 42.0- : I. -
m e ,
- I .
- l - 1.25 W s ' ,
O 35 0 ' $ $ . * -
>4 o ; ; -
O ;, :; :;tl i, ..!!. , $l
. . v.
s.., .a ,. .... e
. ~ ,
j 30.0- y; j j ,{,j !. .,.l,. -- 1.00 ? 'O s, . . , tu
'ce ,l
. ;' '. m 'i
;f, .: .: .. . in
- s. . .
s : . e g 24.0- D l .* -
.e. .t : .
- ! - 0,7s '8 c :..
s.
- s. .
. o t e.o- .! .t. :* j; *
.---"..., g
.l
.,'. ,.~~,... .....:. , ,
j j l
*(6
- 0.50 12.0- l
, e .
- J ; .. :
j ll
,..l : .*
- i. ,'l
'). - 0.25 '
- s.o- .
+;. ' -
LASALI.E MElrOR LOW PRESSURE. l'.
;' . SHORT TERM til'ATION BLACKOUT o.O , , .
a aa 0.0 10.0 - 00.0 30.0 40,0 50.0 60.0. _ 70.0 80.0 90.0 TIME (10 3 s)- 1 i l 1 i l 1 Figure 5.2-20 Concrete Degassing Masses t i l 5 159 i 1 l 1 l I. - . .. .
1 40 .- , , , , , , , t. LASALLE MELCOR LOW PRESSURE --X- Drywell Mass SHORT TERW STATION BLACKOUT Wetwell Mass - 13 1.26- - - - WW Pedestal Mass -
--- DW Pedestcl Mass _ i.2
- - - - Downcomer Mass 1 12- - - -X- DW Mole
.+. WW Mole .- 11 1
1 0.ne - -' ' ' " l
. 0.9 m 0.04- ' -
M -ng C x - S n s g 0,?o- -- 0.7 J
= m j , - 0.e j 0.56- -
. se
..Y,
, - 0 <5 0.42- ',
-- 04 l *
- 1.
-- 0~8 ~
0.28-
+. ,.X' ff.','
~~-x.. .., v,s -
\
l ,' ' . . .4
, , ': r yh.:: .' ~; ... -- + O2
. . . . .y .
0.14 - j , - l ,.. , - 0.1 n oo_gp a , . _q -q TI-' _
. . .-__------y-__-~----i A ,, o
- 0. 0 10.0 20.0 30.0 40.0 60,0 60.0 70.0 80 0 90.0 TIME (103) s Figure 5.2-21 Containment Hydrogen Distribution
.5 160
1B i 1 i , , _i _i i DW Pedestal to Drywell'- WW Pedestal to Wetwell-16- ---
' WW Pedestal to DW Pedestal -- a.s --
Vent Exit Vacuum Breakers 3_ , _
. - Drywell Drain _
3.0 12-3 - X. - 2 5. co ) 1.0 - - 3 f~j m - g ..,,..J
,,~,--
' 6-y o.a-O
. *, - 3 g ./*-
n a y .d N., j - 1.5 O 3 ce. ,U b y , Q l' 's>%, \ i o
-y jel % ,%
~
- 10 0.4 -
/ ' -
J ' 2
.04-
- l- - -). ' ; , ;'
eo._t -
' __. ! . _ ; . . - -: . ' ~. I.E ..~. . , . . . . . . , . . . .....
-- c.o LASALLE MELCOR LOW "TIESSURE Sil0RT TERM STATION 'o. ACKOUT
-024 , , i- i- . . ,
0.0 10.0 2C 0 30.0 40 0 50.0 60.0 70 0 80.0 -- 90.0 TIME (103) s Figure 5 2-22 Containment-Integrated liydrogen Flow Rates 5-161~ _ . - - =
l 250 0 i -i m m T- i i i i 4w LASA!.LE MELCOR LOW PRESS '. SHORT TERM STATION DLACKvUT 225 0- -- 09 Total Mass Flow
-- 08 0- Hydrogen Mass Flow 1
Hydrogen Mole & action I 175.0 - } - 0.7 eCJ 150 0 - r- 06 M C v rd y 125.0 - - 0.5 g b $ m >: j 100 01 t 04 q e 75 0- - 0.3 50.0 4 ^ -- 02
\
\
25 0-- s
+ 0.1
--~__
O O t- ---'
\ ' -
T--T
--__m
~
r-T~-- -- ^^
, 70 0 i -' ^
- 7- ~ "e 0-19 -C - E =B0=7~I=
77 0 0 ~=~i==# 81.0 *20
. 83 0 04 0 PS 0 86 0 TIM E (103) s Figure 5.2-23 Wetwell Rupture Flow 5-162
t i I I l I I I Y I LASALLE WELCOR LOW PRESSURE
; SHORT TERM STATION BLACK 0UT l - -- - Lower-Unit. 2 3e U er-Unit 2 32_ %
-- - U t1 _
4' --- Refueling Bay
'8 - 2e
,i l l Steam 'lbnnel/lbrbine Room 2 3.i . li 8 ii -X- Steam 1bnnel/Ibrbine Room 1 _
8 ' li i II 1 's - 3.4 b giI , O lg li
- O e I 1.2 0 0.9 - 'llII i ; - C 2 h lt i n 3 nn ,
.'. S b
o 0.8 - l i ll I g ll -- 10
- a llII i E Ipt b 0.7 - f/I i 08
!ri i .
I!' s 06- {l1I i - O6 l'. l , I li , 05- j- , , - ij
- 04 g
L- j- x : h
'~ ~
\, ' R_ ,,
J,,, ~_ ? ---~ M O.L M + s-w.e m_,v, 0 3 -NX , ,
'i P --- r i- --
76 0 77.0 78 0 79 0 60 0 al 0 02.0 83.0 84 0 85 0 86 0 TIME (103) s i l Figure 5,2-24 Reactor / Turbine Building Atmosphere Temperatures 5-163
. . , . . - = . .. . . -. -_ - ..
N 0 325 , , , , , , , , , LASALLE MELCOR LOW PRESSURE a 0 300., S110RT TERM STATION BLACEDUT _ k H2O Ozis - ~ -- U2 - I ! --- 02 i i -- CO 0.250- \' ( --- CO2 -
\
\
0225- i \ -
\
i \ 0.200--- \ - en
\
O \ Z 0.175 - 4r i T O I \ b \a \ g 0.150 - s - 3 \\ ' y ,; + ~
%,--- ~~__
e l 0.100- , )--
/
) l /
0 075 - - ii / 0.050 - .?, ft!f , ,!' #,
]fr i/ \ Y c.02s-dl ,1 ~. * :: = . = =: : = = : - ~ ~ ~- -- - - ~ = = =
~ -
Y hl ' l 0 000- , , , ,- , , , , , , i 76 0 77.0 78,0 79 0 00.0 81.0 B2 0 83.0 B4.0 B5.0 86 0 TIME (100) s a i. i
- Figure 5.2-25 Upper Unit 2 Mole Fractions i 5-164
. _ . - - - - . ~_
0 25 r , , , i . . . . LASALLE MELCOR LOW PRESSURE 0 14 - SHORT TERM STATION BLACKOUT - o_ 3_ - - - - - Lower Unit 2 _ Upper ' Unit 2
---- Unit.1 -
2 -
--t-- - Refueling Steam hnnel Bay /lbrbine Room 2 0.11- -X- Steam Ennel/hrbine Room 1 -
g 0.10- l ', - r: '
,0 ,'
y 0.00-h , b o 0.08-g . X . g 0.07-l O II i
- 8 0.00- f ;; ,
I N lI i 0 05- - [ (I I, , Ij l l' I , 004- i
!6'l .
g Il I/j 0.03-Il - - - - - - 1 . . .
; ,ji(
.h , 1 --
- ' ' p ,... w er - - - -
e ,s - 0 02" V ~
- y. I p-
/
J g 'I' 0.01- - I / - - _Jg X X X- X X-- 0.00 - 4X , . , . , , r , . 76 0 77 0 78.0 79 0 80.0 81 0 - 82.0 830 84 0 85 0 86.0 TIME (103s)- Figure 5.2-26 Reanor/ Turbine Bt.ilding Hydrogen Mole Fractions 5-165
~ . - . . . . . - . -., , , . . .- ~.- , - . - -- 2 jd i T-- -
i . i ' : LASALLE ME1.COR LOW PRESSURE :
, Sif0RT TERM STATION DLACKOUT :
1d -: = 4 1 1-a Id ~ : - y 5 .. ~ - --
.. 5 AeoO^ . ,. .
M
,e :
4 m . c . 10": - [ :
- g .,..-_s____..__ ,
4 10 ,, '
/
/ ?
. i -
/ :
i . 10" ~ I
, In-Vessel Total- !
' : / -- - In-Vessel Radioactive - :
, Ex-Vessel Total :
f ---' Ex-Vessel Radioactive - 10' , ; -'
, , , i- ,
0.3 10.0 20 0 30 0 40 0 50.0 - 00.0 70.0 80.0 90.0 TIME (103) s i Figure 5.2-27 In and Ex-Vessel Releases of Cerium from Fuel
- )- 16 ti
=.. , :_
I 45 4 , i i . , 1 i
' LAsALLE WELCOR I4W PRESSURE /
SHORT TERM STATION DLACKOUT 36- - 3.0 - -
- Concretc 9 -- --
' Steel x z.s - -
n
-o O
m y 2.0 - -
! .5 -
l' - l 1.0 - ; 05- / - o.o , ,- , , , , , - -- - , 0.0 10.0 20.0 30.0 40 0 50.0 Sito 70.0 00.0 90.3 TIME (103) s Figure 5.2-28 Non-Radioactive Ex-Vessel Releases 5-167
~ >
10 ,. , pi ,- i i----,---a s%_ ,, i
'/
>. _7
.\ LASAC.E MELCOR LOW PRESSURE o e- -
's SIIORT TERM STATION DLACKOUT ! -
\
} [( l
\
I I l y
.,' ,i 't.__.___.__---.-_..-.-.-----------. ,
s
,o 8-
\!. % -v -V V M'.---.
~
$O f.'},1l,M ' Q t o o~ o 4 j 0.7 - s ~ w -d * - - - - - ! - - - - - - - -
,- -l- -
m c .l 'i-06- hh e - X '
'O -
d ' l' [ - f/l! I X 3 l' O.5 -- , m Yl' to Y ~f g 04- g 5 O -
, ^
Y j , d? O3-i, X - Xe l --- C h 02- t g
--- 1(
4 -t- Te
'j y' '
--X- Ru o t -- -&- Cd - -
! -V- Csl )
g _g .{Q .- Ape x=y ,- 3- g -- ., -,- un , 00 10 0 20 0 30.0 40 0 50.0 60 0 70 0 80 0 90,0 TIME (103) s Figure 5.2-29 Primary System Retention Fractions 5-168
600 0 , , , , , , , , LASALLE MELCOR IM PRESSURE
------ Sif 0RT TERM STATION BL ACKOUT
~__ --
450.0 - 1 .
# l I
400.0- - I I I i 3 1 350.0- ; - I i l 1 300.0-1 ; .
^ I
' g 1 Z l ' I v l g 250.0- i g i E 8 M x : 7 200.0- Fi Total Released _ . l l]
- Primary System !.-~.
- r, - - - - Containment . !.f l r; Reactor /Turtine Bld 250 0*
l,; X Environment S
~l ll l
100.0 - g ** m i
-{1
;I *
,g I '. Ig l
50.0- l l _ l .'
.)
, y->( - , -
l 0.0--fX ,
- - - lX - , j X - ,- - ,
0.0 10.0 f%'0 20 30,0 40.0 - 50.0 60.0 70 0 80.0 90.0 TIME (103) 3 l l l Figure 5.2-30 Xenon Mass Distribution l ! 5-169-i
4~ a 4 260 o , , i , , i , i i IJ.SALLE MEICOR LOW PRESSURE . 2250 SHORT TERM STATION BLACKOUT . 200.0- - Total Released Primary System 175 0- .--- Containment - Reactor /lbrtine Bld X Environment 150.0- -
^
t.0 A M ! .... . . . .. ..... .. ... . .. . ... ......... ..........
$ 125 0- ; ,' -
E '. 100.0- ; - t
.I 75 0- f -
l 1
/
50 0- ,. -
/
i I 25.0- h -
! t c o d4X -" .. ;X ,
iX , iX , IX. . ..: 00 10.0 20 0 30.0. 40.0 60.0 60.0. 70.0 ' 80.0' 90.0'
- TIME (103) s Figure 5.2-31 Cesium Mass Distribution 5 170
s 500.0-- , , , ,
- i. . i .. .i 450.0- -
400.0- *
**'~ ~
Ex-Containment 'Iblal Environmental Release 300 0- -
'EB R
7m-250.0- - e , M l t 200.0- l-1s0.0 - 3 c Xenon Class - - C 100.0- ," .. l l 60.0- .
** LASALLE MELCOR LOW PRESSURE' '
SHORT TERM STATION BLACKOUT
- 0.0 C 76.0 77.0 '78.0 79.0 80.0 61.0 . . 82.0 83.0 84 0 - .- 85.0 - 88.0-.
2: TIME (103 y , Figure 5' 2-32. Xenon Source Term to Environment 5-171
i l i t i e i I - 1.8 , , 1 i i i i i 1.ASAl.II MElf0R I,0W PHESSURE Sil0RT TI:RM STATION DLACKOUT ~ t.e - - ' X X 14- , I [' + Ceslum X Darium o Cesium Iodide ' ' '" i Ex-Containment Total
~
Environmental Itclease
,i I 1.0 - ! -
33 ,g ___ . !
qr R
e i X 0 8- - l ,,_ ,. .. ..
..x. ......
...X .-- - g h - - &W"
-W X ' -
, o4- ,
- ,: . :/
g,. . . ). .
. . 4. . . ..+.. .. .....+ ... ....
02- . ~ _, I , .q. . ...q... . ... q .. . .. . . q . .... . . . O .-* o o -tl % 25- t , t- ,- , , ., , 76 0 77.0 70.0 79.0 80.0 81 0 82.0 03 0 h4 0 85 0 86 0 TP'" '103) s Fi 6ure 5.2 33' Cesium, Barium, and Cesium Iodide Source Terms-5 172
l i l 1 l l 1 i i i i I I 0450- , , , , y i , , , 0.046-t I q Of40-
=
5 l i 1 a.o35_ lodinc . g o.oso- - m Ex-Containment Total - T -- Environmental Release o 0.026- - C e X o.020- - c.ois- - 4 0.010 - } - t 5 0006" { LASALLE Mh0R LOW PRESSURE - j SHORT TERM STATION BLACK 0 tit 0.000 , 1- , , i: , , , , 78 0 ' */. .s 78.0 79.0 80.0 81.0 B2.0 83.0 84 0 - . 66 0 86.0 ) 0 TlMC (10 s) Figure 5,2-34 Iodine Source Term L 5 173
,.4mgn,-.ma e 4a"e m - 59e-g r*q '
,.-y wp u -- ei y.. -.yyvre rjr-ywy .pw ry yT ' vvw- rT y v rew y mT*_
"' f y 4' --'
-FNTt' '"r
- . ...-..-.. . - . . . . . - = . . - - . . . . . . . . - . ..~ . - ... - .- . . .~ - .. .. ~ . . . . - . - - . . - ~ - -
0 250 ,- -, , , , y--- - 1 r* - i LASALLE ME14'OR LOW PRI'SEUHE g/ FIIDHT TERM STATION DLACXOUT
***6' + Tellurium X Uranium -
i Ex-Contsinment Total ' o.roo- l - -- Environmental Releasg D175- - 0.160- -
't2 M 4 T 0125- / -
a 0100-
/ -
/
0076- - s P 0.050- *
, , 4_, . . .
. .y. ... .. -} . + ~=~ $*
O.025- -
.. , .g. .. .. ..y...,.. ..... X. .. .. +.X--- * ** X **
O.000- f[ ' , -- +3 - T" r i i i i i 76 0 77.0 70.0 79.0 80.0 61.0 82.0 83 0 84 0 - 86,0 B6.0 TIME (103) s Figure 5.2-35 Tellurium and tiranium Source Terms 5-174 - _ . , _ .. , - . -. . _ ,__._._._-a., . _ _ ,. _. _.
.1
e 0 70 , , , , ,. 1 i i i LASALLE WE120R IDW PRESSURE S110RTTERM STATION DLACKOUT '
, ,3 _ ,
Ex-Containment Total o so. - Environmental Release - o ss- - l 060- .- i 5 0.46- - 2 c.40- Ruthenium - m I O 0.35- - i O fn E 0.30- - l
- l i o.es. -
0.20- --
~~
i 0 15- ' , 0.10- l. - = o.os- -
-0.00 , , , , , , , m ,
78.0 77.0 78.0 - 79.0 BC.D Bl.0 82 0 83.0 64 0 . 85 0 86.0 l TIME (103 ,) : I i. l.
.1 j
Figure.5.2-36~ Ruthenium Source Term .
. 5 175
_.._..-.;.______.__,. . _ . _...:_ __ .. _ _-.. _ ._.- , _ , ..2_-_.- _ . _-
0350- - r -t i - LASALLE Ml:lCOR LOW PRESSURl'
, i i i T/ '
, ,, SilORT TERM STATION ULACKOUT ,
+ Molybdenum f o.soo. X Cerium f '
A Lanthanum . . j 1; -Containment Total ~
" "78 ~
Environmental Release /- [ 0ELO-0.226-X 0200-M O 0.175 - / - l
}0160-.
/ -
0.126-
/
f/
- D.100-0 075- /
/ -
' ~
'X'
% .K. --
, . - . .... X. ,
~
-9 M ~+ - 'Y
~
- v.qc6. .
, qch .
. rp ~
0.025-- .- s .' 0 0001iWI i r- m i -7 i i , i-76 0 77.0 70 0 79 0 80 0 81.0 B2 0 83 0' L4 0 05 0 B6.0 TIME (103) s Figure 5.2-37 Molybdenum, Cerium, Lanthanum Source Terms 5-U6
. _ . .- . ~ . . _ . . . ._ _ ~ . . - . . . _ . - . . -. . . . .
0.050- , , , , , , , , , LASAtl.E MEICOR IDW PRLSSURE
'~
S!! ORT TERM STATION ULACKOUT 0.046- - 0.040- - 0.036- - I 0.030- - v 3 + X Cadmium Tin 0888~ ~
~
3 Ex-Containment Total a """ Environmental Release c.or0- - 0.016- -
~ + " " " " "' " N ~
,, ,,,, ,,,,,,..........t ............. +......-
0.005- .
-"4' '
l l'
- o. goo gg . , . . i" " " X " - ~ ~ " " " - X " - " - - X "," " " ~ X - " ~ ~ ", X "
76.0 77.0 78.0 79.0 80.0 81.0 62.0 83.0 04.0 66.0 86.0 TIME (10 3 ,) l i i l l l l Figure 5.2 '8 Cadmium and Tin Source Terms 5-177
~
!$ 0 4. , ? i i r- i i i <
4 X _ . , I. xe ) , in s.- --- Cs - 1() i [4 .. l l! --- Te !
,. Sn ) ~
E 12 0 " .
-X- Csl (10) g :X .-
c ' 8 - g 10 5 - a I x h ( D D- f p d:' r-g
/y $ 's- \\
4 : l
.X '~X #'
/ X e
, m,
[ e.0- , I, ; g,,.-------'- - A : ,-f E
<s-
, . /'/
.s {, 'm ',-
y ,. '(. )), : - uv.._.,.-.~,---- --
- ~~-.--.
, g . 7.y 2-.,3 B A . ' f g 3.0- f, .ss ,/ -
);, /
} / - - -
, .s . .
l.ASAl.LE MELCOR 1.OW PHF:SSURE SHORT TERM STATION ULACKOUT 00 . .pg_L _, - r_ _ , _,. , , , _, , 70 0 77.0 78 0 79.0 20 0 81 0 82 0 333 0 e4 0 AS 0 80.0 TlME (103) 3 Figure 5.2 39 deactor/ Turbine Building Decontamination Factors 5-178
t b i l i i I 1 t' i 1.s , .- , , , , ,- 8'-
--..=*
[3nse .**{ h Pedestal Wall Ihilure l'.. ; ~- 200.o 1,3 -
- l ,
e i e i 32 ,' l - 176.0 t, , l 0 . 4 s 1.1 ~ ,I l -
,' . ,. ! 1603 gn. .
,.4
~
0.D - , a - i to ' i o 125.0 ' C l :- q- ' g 0.8- , , - u m l !, 'E w f A O l l l
~
100 0 ' 00- l 5 a 0' 06- l' I 70 0 A E 04~ . - 60 0 03- ,. , d 02- '
.. l LASALLE WELCOR LOW PRESSURE- . Eb o EHORT TERM STATION BLACKOUT (m
01 , , ,- , , , . 0.0 10.0 20.0 .30.0 40.0 50.0 - 60.0 70 0 80 0 90 0 !. TIME (103) s , l-l l l l 4 I 1 1 Figure.5.2-40 Wetvell Pressures 5-179 i _ ~
, g ,,i4, -.m.P9eits t+ n P W W "'-'T*'%"MM'WT~TIDD"" W W3
1 5.3 Intermediate Tern Station Blackout calculation The reacto- tripped and isolated and all ECC failed except for RCIC in the intermediate term station blackout calculation. The ADS operated normally but all containment heat removal systems failed. A model of the 11PCS system was used to simulate RCIC since it could also be used in the long. term calculation descrf. bed in Section 5.4. IIPCS was failed. shortly after the supprossion pool became saturated to simulate RCIC failure on high back pressure and/or DC battery depletion. This sequence is described in more detail in Section 4.3, 5.3.1 overview , 1 The progression of the accident is illus.trated with the list of key events ; and the timing of those events found in Table 5.3 1. Af ter the accident I was initiated and the reactor tripped and isolated, the decay heat began to i boil water within the core and the ADS - was actuated to reduce primary-system pressure and llPCS was actuated to provide emergency coolant. The , ADS was operated gradually, such that, the pressure was reduced at a rate l corresponding to a reduction in the saturation temperature of 100 *F per hour until full depressurization was achieved. The ADS steam flow was piped into the suppression pool where virtually all of the steam was , condensed, heating the pool, llPCS injected water into the reactor vessel upper plenum - with the water supply coming from the CST until the suppression pool level _ exceeded two inches above the normal level, then the water was pumped from the ' suppression pool. The HPCS flow rate was governed by the reactor vessel water level. Ecth IIPCS and ADS were failed at about 9 hours to simulate failure of DC power due - to - bat tery -- depl etion - (i . e . , - remembez llPCS - is simulating RCIC which requires DC power to operate) . - shortly af ter the suppression pool became saturated at about 8 hours. The primary system repressurized following the failure of ADS. The repressurization took- " approximately 4.8 hours with the first opening of a safety relief valve at 13.76 hourt. The core began to uncover at 17.07 hours.followed by a high pressere core meltdown. Coro degradation began in the-upper central portion of the core. ~The first fission products were released from the fuel rod gaps at .18,80 hours, the first core material relocated . at 19.19 hours , and the first core channel blockage occurred at.20.03 hours. As the core meltdown progressed -molten and solid debris accumulated on the core plate, heating the plate resulting in core plate failure at 22;61 hours. The core debris fell into the nearly full lower plenum after the corp plate failed. Large quantities of steam were generated as the debris l partially quenched. ~ The water in the lower plenum completely boiled away 80 seconds after the core plate failed. The part; ally quenched hot debris resided on the lower head, heating-the lower head and penetrations. The inner penetrations reached their failure 5 180
-+-seea u--r Y esaw'-i Fr +g9* vi u. Y am m, y* O s--i- - ee usev'--ivsw=< t 9 C rry -- 'p
,.,.e--r7 -
.-=Ne d 9 =--uP' t *P- w- w**1 f r "Y'9 T eTret
l l temperature at 22.93 hours (19.1 minuten after core piste failure), followed by the depressurization of the primary system. Then, the core debris continued to heat and when the requirac. molten conditions for pouring were reached, the first debris poured from the vessel at 24.20 hours. This molten core debris began attacking the pedestal concrete producing gases which contributed to L.ic pressurization of the containment. intermittont debris pours continued until 26.22 hours. The containment exceeded 60 psig (the venting pressure criteria) at 22.62 . hours and continued to pressurize reaching pressures of 85 (ADS failure pressure), 145 (lower limit of containment failure pressure), and'195 psig , (median containment failure pressure) at 23.01, 31.14, ar.d 34.18 hours, respectively. When the wetwell reached 195-psig, the containment failed with a rupture in the wetvell vall, followed by -rapid containment depressurization. The suppression pool flashed during the depressurization contributing to the depressurization flows. Flow from the contaitunent depressurization entered the reactor- building in the upper Unit 2 . volutne . Three deflagrations occurred in the reactor-building. 5.3.2 Primary System Hydrodynamics The primary system pressure, as shown in Figure 5.3-1, was initially depressurized using ADS, repressurized following.the failure of the ADS, fluctuated with the opening and closing of the SRVs, depressurized again , following reactor vessel failure, partially repressurized as the a containment continued to 1.ressurize, and finally depressurized following > containment failure. The hydrogen partial pressure .in Figure 5.-31 corresponds to the time of the core meltdown. . The reactor vessel water levels during the - core meltdown are shown in Figure 5.3 2. The water level in the core (collapsed) dropped below the top of the active fuel at 17.07 hours and the exposed fuel began to heat. The core was co'spletely dry at about 19 hours. The primary system integrated hydrogen flow rates are shown . in Figure 5.3 3. A natural circulation loop was ectablished'between the core. channel. and the core bypass volumes with the flow going up the channels- and down i the bypass. 5.3.3 Core Meltdowr. j_ - The physical destruction of the reactor core began in the _ upper central portion. of the core _at 18,80 hours with the failure of the fuel rod cladding to contain trapped fission product' gases. located in the fuel rod - gaps. The first core material relocated at 19.19 hours, and the first, core channel blockage occurred at 20.03 hours. As the core meltoown progressed, 5 181
molten and solid debris accumulated on the corc plate, heating the plate resulting in core plate failure at 22.61 hours. The total maaset; of zirconium, zirconium oxide, steel, and steel oxide are shown in Figure $.3 4, along with the mass of hydrogen produced in t)e primary system. The masses of zirconium and steel in the core were reduced by oxidation and debria ejection. A total of 796 kg of hydrogen was produced in vcssel (5.6% of it after the vessel failed). 5.3.4 Containment flydrodytun.ics The containment contained the released fission products until it failed due to overpressurizr: ion caused by the addition of steam, non condensible gases, and heat. The containment pressures are shown in Figure 5.2 6. The pressures increased slowly at first while the suppression pool heated to saturation, then rapidly as SRV flows pr.ssed through the suppression pool without condensing. The spike in the drywell pressure at 23 hours was due to reactor vessel depressurization following lower head failure. Then the containment pressurization continued due to the boilitig of water in ' the pedestal and the production of gares from the interaction of core debris with concrete. Containment failure was predicted at 34.18 hours with a rupture in the wetwell wall. The atmosphere and surface temperatures of the dryvell and wetwell are shown in Figures 5.3 6 and 5.3 7, respectively. Tne pedestal temperatures are shown in Figure 5.3 8. The wetwell pedestal atmosphere temperr.ture peaked at 2020 K, hot enough to melt steel and ablate concrete. Pedestal temperatures decreased and fluctuated af ter the containment failed due to the deprescurization flows through the pedestal. Steam condensation within the containment was an important factor in the containment pressurization and is illustrated in Figure 5.3 9. The water masses for' pools accumulated ir the dryvell, drywell pedestal -and wetwell , pedestal are shown in Figure 5.3 10. The collection of condensate and pump seal and control rod drive leakage accumulated into a substantial pool in 4
, in the drywell pedestal (30% from leakage). This pool interacted with the core debris.
Tnis water U.oved into the wetwell pedestal, af ter the cavity drain pipes failed, filled the wetwell pedestal to its access ports, and . then spilled over into the suppression pool. The vetwell pedestal was nearly - full at tha time due to condensation and to a swelling er.d overflow:-in the already saturated suppression pool, which occurred during the-boil off of water in the reactor vessel lower nienum, _ The timing of-the drywell pedestal-floor failure was more important to this calculution than it was to the abort-term station blackout calculations, i.e., if all the drywell pedcstal water had been boiled off before the floor failed, this water would not have over flowed into the suppression pool and the containment would have failed earlier. W 5-182
. ~.- _.- - . _ ~ . . -- -.- ~. -- - .. .. -.- - - - t i The suppression pool _ mass increase and temperature are shown in Figure 5.3 11. The pool mass decreased initially as a result of IIPCS pumping from the pool, then increased as t'ie ADS and SRV flows condensed in the pool. When the contain:u nt failed, the mass of the pool decreased due to flashing + of the satursted water. The downcomer and wetwell suppression pool water levels are shown in Figure 5.3 12. The downcomers clear at vessel failurs and at containment failure. The first molten core debris poured from the reactor vessel lower head into the drywell pedestal at 24.20 hours and consisted of liquid steel along with a small amount of steel oxide. Trenty :inutes later the floor failed, as modeled, and the debris transferred completely to tho wetwell pedestal. The masses and temperatures of this steal and steel oxide _ are shown in Figure 5.3 13. Later pours from the vessel passed directly through the , drywell pedestal to the vetwell pedestal. The debris masses for the wetwell pedestal core / concrete _ interaction calculation are shown in Figure 5.3 14. The first debris arrived at 24.53 hours consisting of steel and steel oxide and the.first fuel a short time later. The debris teapert.tures _ are shown in Figure 5.3 15.- _Durin5 the period of rapid zirconiwa oxidation. the temperature of the heavy oxides peaked at 2587 Yu Unlike the other calculations, this calculation did not predict a layer flip. i The maximum radial and axial molten pool penetration distances into the concrete structure are shown - in- Figure 5.3 16. The end of calculation penetration distances were 0.34 m axially and 1.71 m radially. -The radial i penetration distance exceeded 1 the pedestal wall thickness of 1.47 m at-33.67 hours. The vertical penetration of .34 m was much :less than the other calculations and rsason for the difference is not apparent. i Carbon dioxide, carbon monoxide, hydrogen, and steam gases produced as by- t products of the concrete ablation and the resulting chemistry of the molten pool are shown in Figure 5.3 17. The end of calculation totals were 17190, 848, 713, and 3180 kg, respectively. The in-vessel and ex-vessel hydrogen production are compared in Figure 5.318. A total of 1509 kg of. hydrogen was produced (796 kg in-vessel). The combined integrated mass and release rate of steam.and' carbon' dioxide'- gases released from ' heated concrete by dehydration are shown in Figure 5.3-19. The end' of calculation total mass was 37800 kg. The mass and mole fraction distribution of hydrogen within' the- containment is shown in _ Figure 5.3-20 and . the integrated hydrogen flow rate is 1 illustrated in' Figure 5.3 21. I 5.3.5 Reactor Building liydrodynamics The containment failed with a rupture in the wetwell wall at 34.18 hours. e The containment depressurization Lows are s,.1own in - Figure 5.2-23. The hydrogen mole fracs. ion of the depressurization flow in also shown. I i 5-183-u+w---s - y e- imw -w tw,tw- e n- w r--tvewe-- w w awv.. -,.e.-amw- ---++-------si++--iar-ree-w e-e-- <=, -Aw e+-+r en-"2-w+e<.--w-mew 5w+++w,.
Flow from the containment depressurization entered the reactor building in the upper Unit 2 volume and three deflagrations occurred in the reactor building; two in the upper Unit 2 volume and a third in the lower Unit i volume. The other volumes did not ignite either due to insufficient hydrogen or to being rendered inert by steam and carbon dioxide gases. The reactor building combustion dynamics are illustrated by: (1) the - atmosphere temperatures shown in Figure 5.3-23; (2) the mole fractions for the gases in upper Unit 2 shown in Figure 5.3-24; and (3) the hydrogen mole fractions for each volume shown in :'igure 5,3 25. 5.3.6 Radionuclide Transport The transport of fission products throughout the plant and the resultin6 aource terms to the envirotunent are illustrated by a summary of the end of calculation results presented in Tables 5.3 2 through_S.3-7 and by examples of time dependent results _ shovn in _ Figures 5. 3 26 - through 5 3 39. _ The f~rst release of fission products from the cladding-gaps occurred at 18.80-houts. Two examples of the time dependent releases are the cerium class fissjon products shown-in Figure 5.3 26 and the ex-vessel releases _of non-radioactive structural aerosols shown in Figure 5.3 27. 8 The pritnary system retention factors are listed in Table 5.2 3 and shovn-fr selected classes in F1 6ure 5.3-28. The retention factor for uranium could not be calculated because uranium was not released in-vessel, The fractions of the total fission. products released from the fuo remained in the containment (excluding the primary system) are listeu in Tsble 5.3 7. The distribution of these fissiot- products within' the containment is also given. The time dependent mass distributions for the xenon and cesium classes are shown in Figures 5. 3-29 and 5. 3-30, respectively. The xenon-class gases readily transport from the primary _ system and. containment and virtually all of them would eventually escape to the envirotunent. . The inajority of the cesium class deposited either on surfaces or in the = suppression pool resulting in only a small amount being released to the environment. The fission pcoc'uct source _ terms to the= environment are shown in a series of eight figures, Ligures 5.3-31 through 5.3-38. The masses escaping the containment are showe. along with the masses entering;the environment. The reactor building decontamination factors, shown for selected classes ' in Figure 5.36 39, demonstrate the transient nature of the release and transport of fission products. 5 184
l I i Table 5.3-1. , Intermediate Term Station Blackout Events i Event Time Event Descrintion Seconds Minutes Hours
- 0. O. O. Accident initiated, Reactor Tripped, Isolated 28800. 480. 8.00 Suppression Pool Saturated 32400. 540. 9.00 ADS and HPCS Failed 49550, 826. 13.76 Primary System Fully Repressurized 61463. 1024, 17.07 Collapsed Mater Level Below Top Active Fuel 67689, 1128. 18.80 Fission. Products Released from Ring 1 Rod Gap (First Fission Products Released) 6848t. 1141. 19.02 Fission Products Released from Ring 2 Rod Cap 69100. 1152. 19.19 First Co".e Materials Relocated 72120. 1202. 20.03 Core Channel in Ring 1~ Blocked 72205. 1203. 20.06 Fission Products Released from Ring 3 Rod Gap 73901. 1232. 20.53 Core Channel in Ring 2 Blocked 77/90. 1297. 11.61 Core Channel-in Ring 3 Blocked 78245. 1304. 21.73 Fission Products Released from Ring 4 Rod Cap 81405, 1357. 22.61 Core Plate in Ring 1 Failed 81416 1357. 22.62 Containment Wetwell-Pressure _ Exceeded 60 psig; 81485. 1358. 22.63 Lower Plenum Dried out 82553. 1376. 22.93 Lower-Head Penetration in Ring 1 Failed (Reactor Vessel Failed) 82839. 1381. 23.01 Containment Wetwell Pressure Exceeded 85 psig-87030. 1451, 24,18 Core Plate in Ring-2 Failed
- 87114. 1452, 24.20 First Debris Ejected from RV to DW Pedestal 87194. 1453. 24.22 Lower Head Penetration in Ring 2 Failed 08199. 1470. 24.50 Core Plate in Ring 3 Failed 88309. 1472. 24.53 Drywell Pedestal Floor Drain. Pipes Failed 88314. 1472. 24.53 Wetwell Pedestal Received Core-Debris 88389. 1473, 24.55 Lower Head Penetration in Ring 3 Failed 88660 1478. 24'.63 Debris Ejected 88692. 1478, 24.64 . Debris Ejected -
92770. 1546, 25.77 Debris Ejected 94377. 1573. 20.22 Debris Ejected 112088. 1868. 31.14 Containment Wetvell Pressure Exceedou 145_psig-121194 2020. 33.67 Wetwell-Pedestal Vall Penetrated 123049, 2051. 34.18 ' Containment Failed with Wetwell Break 123161. 2053. 34.21 Deflagration Burn in Upper Reactor.Bld. Unit 123301, 2055, ;34.25 Deflagratina Burn in Upper P.eactor Bld. Unit 2 128009, 2133. 35.56 Deflagration Burn in Lower Reactor Bld. Unit 2-131748. 2196. 36.60 Calculation Terminated l i. 1 1 5 185 y-w- w -w-e- p y --ys' y - g-e-~--wy--- +
Tabin 5.3 2. Summary of Radionuclide Releases and Transport Initial Fraction Ratio of Ratio of Fraction Reactor Bid.. C'.a s s Core Released. In to Ex Total to Retained -Decontamination NO. Rep. Inventory
- from Fuel Vessel Radioactive by Primary Factor (Kg) 1 Xe 463.7 . 962 13.7 1. 2.6E 4 1.5 2 Cs 268.4 . 962 4.52 1.13 .81 2.3 3 Ba 207.5 . 567 .356 1. .77 1.9 i 4 1 20.93 . 956 18.7 1. .96 . 2.9 l 5 Te 40.79 . 922 27.3 1.12 .81 2.3 l 6 Ru 307.0 6.83E 5 146. 754. .66 9.3 1 7 Mo 350.6 . 0122 452. 305. 75 2.9 l 8 Ce 594.0 2.66E L 2.50 120. .62 2.9 '
9 la 571.1 . 0508 .0781 1. .71 2.3 10 0 132390. 8.99E-5 0. 1.13 - 2.6 11 Cd 1.407 . 440 1130. 1. .21 3.1 12 Sn 8.587 .452 351. 1. .80 2.2 16 CsI 0. .81 2.1
- Radioactive Fission Products (except Uranium)
Table 5.3 3. Fission Product Distribution (kg), Classes 1+4 location Radionuclide Clagg Xe(1) Cs(2) Ba(3) 1(4) Releases from Fuel Primary 4.159E+02 .1.919E+02 3,090E+01 6.519E 05 Cavity 3_.033E+01 4.543E+01 8.670E+01- O. , L Overall Distribution Primary 1.073E 01. 1.546E+02. 2.376E+01= 6.241E 05 Containment ' 1.873E+01 5.467E+01 6.542E+01 2.781E 06 Reactor Bld 1.279E+02 1.803E 01 3.b48E+00 1.965E 10 Turbine Eld 1.655E+01 1.099E-02 3.611E 01. 7.571E-14 Environment 2.829E+02 1.473E-01 4.443E+00 9.640E 11 Containment Distribution Drywell 1.75AE+01 2.260E+00 .4.160E+00 4.052E 09 Upper Pedestal 4.622E-01 4.207E 01 2.042E+00. 2.703E 06 Lower Podestal 9.572E-02 8.626E 04 '4.214E+00 0. l Wetwell' 6.374E 01 5.199E+01 5.501E+01 7.411E-08 Suppression Pool- 0. 4'.908E+01 5.009E+01 7.400E-08 5 186 , h.1 _mgs- - g4 '
_ _ _ _ _ - _ _. _ _ _ . _ . _ _ _ _ - - . - _ . _ _ . _____.__.m.__.. _. _ l Table 5.3-4. I Fission Prodret Distribution (kg), Classes 5 8_ Location Radionue11de Cing.g Te(5) Ru(6) Mo(7) Ce(8) Releases from Fuel Primary 3.627E401 2.084E*02 6.465E 02 1.130E 02 Cavity 1.327E400 1.429E 04 4.210E+00 4.518E 03 Overall Distribution Primary 2.923E401 1.380E 02 4.892E 02 6.983E 03 Containment 2.968E+00 5.301E 03 2.754E+00 6.635E 03 i Reactor Bld 1.636E 01 1.122E 06 - 7.426E 01- 3.329E 04 Turbine Bld A.521E-02 7.307E-09 1.070E 02 3.992E 06 Environment 1.417E 01 '. 355E 07 3.987E 01 ;1.755E 04 L Containment Distribution - i Drywell 9.641E 02 8.231E 06 6.096E 02 2.500E 04 l Upper Pedestal 3.498E 02 1.659E 06 3.126E 02 1.348E 04 Lower Pedestal 4.258E 03 8.402E 05 1.857E+30 4.301E-04 . Wetsoll 2.833E+00 5.207E 03__8.053E 01 5.820E 03 , Suppression Pool 2.731E+00 5.194E 03 7.178E 01 5.535E 03
.1 Table 5.3 5. _ .
I Fission Product Distribution (kg), Classes 9 12- ; Location Badisnyclide Class _ La(9) U(10) Cd(11) Sn(12) Releases from Fuel ' Primary 2.102E+00 0. 6.179E-01 3.773E+00 ; Cavity 2.690E+01 1.190E+01 5.456E 04- 1.074E 01 ', Overall Distribution . _ Primary
- 1,483E+00 1.649E-02 1.307E 01 3.018E+00' i containment 2,291E+01 9.704E+00 4.819E-01 _4.815E 01 ,
Reactor Bld 5./90E 02 1.263r 01 '2.945E 04 2.807E 03-Turbine Bld' 2.648E 03 2,876E 03 4,034E 06 '1.651E 04' Environment 4.828E 02 7.894E 02- 1.408E 04 2.486E 03 Containment Distribution. l Drywell_ . 6.100E 01 2.677E 01- 1.308E 02 1 321E 02 . Upper Pedestal 2.139E 01 9.443E 02 - 5.531E 02 6.844E 03 t Lower Pedestal 1.575E+01 - 6 716E+00 6.045E-04 6.071E 03 Wetwell 6.356E+00- 2.626E+00 4.129E-01 4.554E-01 ;
-Suppression Pool 5.758E+00 2.357E+00 3.445E 4.471E 01:
r o
~
5-187. > _._ _ _ _ ._._.-.._.- . _ _ _. _._ _ ~. ..-. _ _ - - . _ . _ _ - - _ _ _ _ _ . _
Table 5.3 6. Fission Product Distribut.4on (kg). Class 16 Location Radionuclide Class Cs1(16) hieases from Fuel Primary 3,837 E401 Cavity . 2.587E+00 overall Distribution Primary 3.097E4 01 Containment 4.413 E4 00 Reactor Bld 5.449E+02 ; I Turbine Bld 8.928E 04 Environment 5.249E 02 Containment Distribution-Drywell 1.572E 01 Upper Pedestal 4.844E 02 4 Lower Pedestal 1.074E 03 Wetwell 4.206E+00 Suppression Pool 4. 042 E4 00 Table 5.3-7. y Containment Distribution of Fission Products '
- _ ,{
Containment Location within Containment-
- Class Fraction of (Fraction of Containment Mass)
No. Rep. Total Release Drvve11- Wetwell Sunoression Pool 1 Xe .042 961 .039 0 2 CS .230 .049 .951- .898 t 3 Ba .556 .095 .905 .166' 4 1 .043 .973 .027- .027 5 Te .079 .044 .956 .920 6 Ru .253- .002 .998- .980 E 7 Ho .644 .033 ,967 .261-8 Co 419 .058 .942 .834 9- -La-
.791 ,036 .964- .251 :
10 U .816 .037 .963 .243 11 Cd .779 .142- .858 .715 ' 12 Sn1 .124 .042 .958- .929 16 Cs1 .108 .047 .953 .916 '
~
188. s - _ , . _ . - . ._,_,-__..__,_-_._._......_.4 ,. . _ . _ . - _ . . . _ . _ _ _ - - - - . _____
- . . . . - . . . . - _ _ . .- . . . . . . - . - . . - ~ . . - - . - - . - . . - . - . - . . _ . . -
i
\
l 1 eo , , , , , , , , , .- , , , l LAsAuE ME*.COR INTERWEDI ATE TERM STATION BLACKOUT _ g3 l 70- - 1o l 7
- - oo e.o - : --
Total - j: - -- Ilydrogen , 1; ~' - o.e l l m [ 6.0 - ll - 0'
;: '?
w ,. - o
- C S
,, 4,0 -
C : :l. l' 0.*r o~ U O m *l C g : .: 06 C. I'. ' 30- { ! :"*l)j -
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. I*
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- - c.1 j'. - --'/ :
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, , , , , ',- r ,
40.0 60 0 70 0 - 80.0 60.0 90.0 100.0 110.0 120 0- 130.0 140.0 . TIME:(103) 3 t l I t l Figure 5.3 1 .Mre Channel Pressures l'c 5 189 l~
?
- , _ . _ ._ = _- ,
.m- er. .. ,w.- ,m-- , - ..e - .w_,...,.~44,,,..M...-,m- m,. e-,4. . . ,.m~..a.- ,e.- . A,4 r-
160y v , , , i , , , ,
}fj LASALLR WELCOR INTEhWEDI ATE TERM STATION DLACKOUT B50.0 13.5 .
l - Upper Plenurn soo.o 4 -- - Downcomer 1 la o~ -t-- Core Channel r.
% -X- Core Bypass 4so o
\ - Lower Plcrunti se s- V I
- 400.0 l
. v,,
'{ TAF' l
,3
,) m . e
; X D 0- '
--- 3$0.0 v
- Jet INmp~
g Nozzle
,$ .i 3_
,i
_- 300.o ?. c ,; fg .. 't S ' ll lU . '+ s~ lk t o 9::s , 3 - 230.0 f3 eo- fplig; ! ? ...a , DAF
%I,' -> A%4 Xi X4 - 200.0 46- ,
1
- ~ '
, ~ , , * * ' - 160.0
; Downcomer
' O U'"
30- , l 100.0-1.r - l _- 60 0 0.0l, ! Lower liend na 60 0 59.0 88.0 17 0 . 864 05 0 104.0 113.0 122.0 . 131,6. 140.0 - TIME (103) s Figure 5.3-2 Reactor. Vessel Water Levels 5-190 e r -m- Wu-m m+s ' n--
- e--
Yit------( T- -T'.e6 -=6-vm'i- y w-y ,--9 ei y piu' yN4+w'* wt-mve *- ,--,y-% prm.
2
$b) T I
'7 g i i 3 LASAllE MELCOR INTERMEM ATE TERM STnTION BLACK 0UT in- -G-- Upper Plenum to Dome -
--G - Dome to Downcomer
- Seperator Drain -
- ea 100 ~
-+- Jet Pump
-X- Charinel Inlet -
+ Bypass Inlet *- - 20
-V- Channel Outlet - -
0 -+- Hypass Outlet 37s. --- Safety / Relief "alvo Vessel Breach r--------- a _n n
~
3 '
} c so- / ,- g- ' "
-so e
> o ss- <V j y/
-.o, l3 0 ^
o.00 - DNW.* (ir3We [^ i _y
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-0.60 - [~'
?
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9
--e.o
-1.00 ,
80.0 63.0 66 0 69.0
, , , , , c- -
72.0 75.0 78.o 81.0- 84.0 87.0 Go O 3 TIME (10 s) Figure 5.3-3 Reactor Vescel Integre.ted Hydrogen' Flow Rates 5-191
1 o0.0 , , , , , , -, , , LASA _ - - - _1.LE WElf0R INTERWEDIATf: TERM BTAT10'.4 DLACKOUT _ ~__ % s 80.0- g Zr i .-.- Zr02
- 12 70
1 --- Steel ~ i --- Steel Oxide i - Hydrogen
- 'Q '_ \ -
~ so.o. ~ n.o y n n a 1
--___ -_-_---_ o e
y 50.0- n 2 p
....................................... .co ,_
o ....
- a 31 u : o ta O 40.0- ,. -
n ' a - o.e 5 ,' Z e .. y 30.0- l - # l l . oA 20.0- : .-.- l ,'
. , \.
I
- ./ * '
- 02 10,0 - l \ -
. ./ ,
il o0 , . y ,.__, s\y..._._._,.-__...._. aa 50.0 69.0 88.0 77.0 66 0 95 0 104.0 113.0 122.0 - t 31.0 140 0 TIME (103) s Figure 5.3 4 Total Core Masses-5 192
1A , , , i i i i i , i i - r- - -
- 1.ASA1.1.E WELCOR INTERWEL ATE TERW STATION DI.ACKOUT -
I'4 ~ ^ 200.0 1.3 - - 12-Drywell l Dh Steam ~ 8 ' Wetweli 1.1 - -- + WW Steam -
-160.0 1.0 -
[09- -
- 125 0 0 0 e- .
..h l
3 w ,
.. g.
& '. -j - v 3 0.7 -
,+,.
-- 100 0 g 0e- l c 75.0 n3, q 1 -
0.4 - r k, ,l o 0.3 - ,T ' - I - 0,.
..' .y ..
, * - e5.0 01- .. .. . .., ,# -
N 0.0 - - - - ,
- 4, - , , , ,
aa 0,0 10.0 20.0 30.0 40.0 60.0 50.0 70.0 ' 80.0 900 100.0 110.0 120,0 130.0 140 0 -
. TIME (10 3 s)-
Figure 5.3'.5 Containment Prussures 1 5~193 1 _a- --_-a-- - - - - - -
, ..-.-.-- - . . .- . . . .- _ . . . . . - - . - . . _ . - . - . - - - . ~ . - . . . - , - . . - . > - - - - .- a > > - . ..c 1 I 1. i I, i t i i I 600 0- r -1 i r--~1 i 1 i i 79 -i 9 ^ LASA!.11 Mf !r0H INTYNWr:D! ATE TFHM STATION If LACKOT 4th o 480.0-1
- 400 0 4
J 3760 i 460 0- - ! t ' , Db0.0 l 440.0- .' f j [f
~ 326.0 -
'.- e',,,.- % % i (2 4g0 0. / y 300 0 '
{ g' g i i v .N-i ! 8 - I:76 0 i D C l w !
*{ 400.0- ,
- y df. 200 0
" w.1 i
, E , l @ 300 0- ,, 228 0 '! 4 I d } ,
, ' f/ 200 0 ,
f 360.0- -' ' i J
/ > ; oA*, 176 0 340.0 ~ AlTTiosphere -- 3 so r.
y
,- -~ ~ ~ f~ , -
~._
Liner pjoor Reactor Shield 886 0 8 00" i-
~
Misc. Steel
- 100.0 t
r = 300.o r , , , , , , i- , , , , , i 0.0 10,." 20.0 30 0 40.0 60.0 60.0 70 0 80.0 90.0 100.0 '-110.0 120 0 130.0 ~ 140.D I TIME (10 0 s)I 4 d i i Figure 5. 3 Drywell _Teinperatures 3. 4 __5-194 _ _ J avg Wrm-= wv me 41 w- 4-WY mir +r-)DID-rest-'y %---m+--~ e- up-de-.r =+ieMref- e-W++1-y -> +.me*try7e3w'W * 'y+e vatyw-
. . . m ._.m..._.._. _ _ . . . -,.m_ ..__~.m.__.....- , . . . . . _ _ . _ . . _ _._ _ _ -_.
4 I SD0 0 , , , , , , , , , , , , , LASALLE WELCOR INTERWEDIATE TERW STATION DLACK007 426.0.
. r 46U.0 ~
f
- 400.0
~l .
t 460.0- !
- l
[ - D50 0 440.0- - =m [ - g I
* ,, , - 325.0
/ \,-
2 420.0- ,_ / . _. g
-- 300.0 ;
v - s / N ,
- s ,] 'N' g 276.0 3 s c g 400 0- f \ ~
,*f \ 250.0
\
g , -4 *
$ $50.0- '
-- 226 0
/ /
/
r 4 20D.0 380.0- , ,j '* - t
- l' sna .
3404 - .' -
'- 150.0 t
. ./
/ Atmospherc . 333 o sec o- .
Suppression Pool V. .*,.' / -
. . - - . - gal;
.--. . Misc, steel 200.0 300.0 -,
1 , , , , , r- , , , r- i 0.0 10 0 .20,0 30.0 40 0 60.0 60 0 70.0 80 0 90.0 100.0 110.0 120.0 130 0 140.6 TIME (103) s Figure 5.3 7 Wetwell Temperatures
-5 195
-. m c..., y if i--- .. g m y ..rg--- i
.i.yg. g g g --4mv, g 7_,77.sg.,p yyr 9m-,iy ,ey-- r y'1T
3 i i
)
22 , , , , , f , , , LASALLE MELCOH INTI:RWEDIATE TERW STATION IILACKOUT a.ss , e.0 - -X- DW Pedestal Atmosphere +. _ WW Pedestal Atrnosphere P
'8" j X- DW Ped Wall Surface
~
+ WW Ped Wall Surface '-
~
?
-- E ts .
--- WW WWPed Ped WallWall Interior Interior(.30 M (.11 M)) l l
- 1.6 - - _
i . J
- 2.25 -
f 2 , ! 14-O l ,
- 2 00 p
'l O l i [- E : :
' - 1.76 r3 3 1.2 - *
- o
- F .
O 8 : / d { -] '
/ '. I L1 - 1.60 fi 1.0 - f f [} ' ,
/ ',
l,,._ , ' - 1.25 t 0.8 - * / .-
-- 1.00 t s ll
.. -f 8 ll 'I 0.rs 06- 4' ' # '
i
, . .n i . . . M ' ' +g - 0.s0 i- -
,====.gj 0.4 - ,
,- ._ ,,3 0.00 0.2 -- , , , , , , ,
80.0 68.0 76 0 84.0 92.0 100.0 108.0 116.0 1N4.0 . 1$2.0 140 0 TIME (103) 3 t Figure 5.3-8 Pedestal Temperatures 5 196 W 7
_ . _ _ . _ _ . . . . . .. .__ .._ ._ . __ . . . _ _ _ . .. -_._.m . . _ _ _ . _. _. . _ i r 3 60 , , , , , , , , m -- , ,- , , ,u u LASAllE MElf0R INTERMEDIATE TERM STATION DLACKOUT ,/~~
/ 85.0 3.16 - --- Drywell Liner Flux / -
-- Dr well Floor Flux /
W Wall Flux / -'"
- - - Total DW Condensate- /
f.80- /
-- 65 0
/
/
~
2.46" f 9 / 2 i
- 4s.0 (x2.to- / -
' - 40 0- @
n , n
'S- -- 35.0 O O -
- l I
m M M g ,# - 30.0 j A 1.40-R = e - 2s.0 e i
- n ,
1.06- / .,- 20 0 p' n
# - 16 0 0 70-
/ ..i st M 10.0 0 35- ' .
- ."". - 1 -
..] .....,4 s s, (,j Q;4. ft s ! ' ~ ', . ' 4'. . s.0 j ,. ..'4
- ' ' : -' ' ,gg",
0.00 , ,
' , ,' I
, , , ,I ,
an 03 10 0 20.0 30,0 40.0 60.0 60.0 70.0 80.0 90.0 100.0 110.0 120.0 130.0 140.0 TlME (103) s l l 1' l l-i t. Figure 5.3-9 Containment. Condensation l I i .. 5-197
.. _ . . _ _ . _ . . . _ _ _ _ _ _ . _ _ _ _ _ _ _ . _ _ - _ . _ _ .. _ ... _ _ _ _ .__ _ .__ _ _ _ m - _ .
1 1 60 0 , , , , , , , , , , , , , LASALLE WELCOR INTERWEDIATE TERM STATION DLACKOUT 45.0- Drywell Pedestal .- luo u Wetwell Pedestal
- - - Drywell o0.0 40.0- -
I
- 80.0 i I
35.0- * = i j 10.0 l
'-1 30 0 ~ -
9 x
'. :S
; . - 60.0 n . '. n 3 25.0-w l '.
v S y l . - 60.0 20.0- .
'. - . 40.0 - ,
16.0 - ' l. '.
; ', - 30.0 l '.
l '.
- 10 0- i , '. -
l - 20.0
/1 -
; I-l ;
50- {l I I, -
- 10 0
,,.......I-l
'- 00 - , - - ^ ,.~'['",'- 7', - r ,~ ^ 7 - , - ,- , ! ad 0.0 10.0 20.0 30.0 40.0 60.0 60.0 ' 70.0 80.0 90.0 100.0 110.0 120.0 130.0 140.0 , TIME s (100) Figure 5.3 10 Drywell and Pedestal Water Masses 5 198 i- __,_,1_ i ._ . ili' . . . _ _ , . . _ , _ . _ _ i_. . . . _ . _ _ . _ _ _ u.-..._-.._..~ ._
.a.~,.. -~. - . - . . - . . . - . . - . . . - . . . . - _ - . . . - ._ ~ . , . . .. . - . . . . . . . . - - .
i 600 0 i , , , i i , i , e i e i LASALLE WElf0R INTERMEDIATE TERM STATION DL4QK0tg ,,,,,,,,, . ... .. - l 'a 450.0-400.0- ,. '., -- 400.0 3SJ 0- , . ** * . . . . . . . . ' j - 380.0 t 300.0- . 250.0- *
/ -- 360.0 v h.200.0- j /
g ,' - h. g E
% 160.0- l
.! U -~ U O'O 8
8 : o 100.0- / -
,/ - 320.0 50.0-' -
0.0 - -- 300.0
- 60,0 - -
- 28 0.0 --
-100.0- -
ini lal Water Mass Mass increase of 3.682,400 Kg
...*-- T erdu m .
-150.0 , , , , , , , , , , , , ,
0.0 10.0 20.0 30 0 40.0 601 50.0 70.0 80.0 - 90.0 100.0 110,0 120.0 130.0 14 0.0 TIME (100) s i. r l l Figure 5.311 Suppression- Pool Mass and Ternperature : l 5-199
-i j
, . .. .. - - . . . .- . ~ -....n_,n - _ . . . .+ - _ . .. . .
1 -- i 2 1 2 4 d ?- t 2 i i i s ?. i 1 4 -j -15 0 , , , , , , , , ~7- , , y. LASALLE WELCOR INTERMEDIATE TERM STATION DLACKOUT i- Port ik ttom -Elevetion 16.7- - ------ - - - - - - - - - - - - - - ~ - - - - - - - -
- 705 0'
. ,T., i, i.
-l 8.4 - * < - - - '
, w, .
.i ',
- 702.5 s
I .
- 17.1 - ,...
* = -
,,...., ,.... - -- *** l
{ - 700.0
\!c
- 17.8 - -
v . 3 - 697.6 ~h 1
- . :) .;
e
.o-* - 18.5 - ' - as
> > 1 v
i 1, j i' I 606.0 Q s
-19 2 - h s:
Downcomen - 692.5
-19.9 - WCt'#0ll - ! -
, - 690.0
- 2 0. 6 - -
I Downcomer Exn 4
-. - - - - - _. - - - _. - - - - - - - - <-. - - - - ,. . L. _-
-21 3 - -~ 007 0
-22.0 i , , , , , , , , , , , ,
-NO ~l u.0 10.0 20.0- 30.0 4' 60.0 S0.0 . 70.0 80.0 90.0 100.0 110.0 120.0 130.0 140.0 1 TIME (103) s i
Figure 5.b12 Suppression Pool-Water Levels t , 5-200 r_ -)
s 3-30 i i i *=w i i i , i i LASALLE ME!IOR INTERMEDIATE TERM 5"TATION BLACKOUT e
\/ \/
72- ^ ^
, X -
i. 6.4 - e . t - loo ~ g'f.X T se. : s x- -
- '- , , - 1.25 2 4 .a - : ','-
g - g x g' +.,', n a e 94,o- 4, . , , y
, ys -- 1.00 o m
< g g
- ' J, o 3e- ,,
E
, l;;,
- c.7s g!
s . . 2.4 -
* ,', )( -
- 0.50 1.6 --
' .ag
)k i Li t Oxide Mass o.e - t X M O Mass - 028 S p
. 4. . Light Oxide Temperature X ~
- -X- - lietal Temperature -
v n oo- IXlXr iXlX , ,! I ! a aa 88.5 88.7 88.9 87.1 87.3 87.5 - 87 7 i 87.9 88.1' ,'.3 08 88.5 T1ME (103 ,) _ Figure 5.3-13 Drywell Pedestal Debris Masses and Temperatures 5-201 Y4)
B00 0 i i .- , , , , , , 1.ASALLE MELCOR INTERMEDIATE TERM STAT.ON BLACKOUT
? coo- 'Ibtal . .-1s Heavy Oxide Light Oxide
-- Metal . l 640.0- -
;g4 i 560.0- -
- 1.2 m 480 0- -
k 1. 0_ y
;3
% ~400.0-v to m O
- 0.8 .~,,,,, -
rn 320.0- -
* ' ~ ~
, - - _ 06 240.0- f - -. - -
~
,_] ** - 04 '
100.0- _
-J
- 0,2
' ~ ~ ~ ' ' ~ ~ ~ ' ~ ~
80.0- , 0.0 - , ,a f-" "[ , 7 , , , en 80.0 86.0 92.0 '98.0 '104.0 110.0 116.0 ..122.0 128.0 134.0 ;:140.0 TIME (103) s Figure-5.3-14.Wetwell Pedestal Debris Massns 5-202
~ ~
--- _______2
33 , , , , , , , , i LASALLE WElIOR INTERMED;A'lE TERM STATION BLACKOUT 80-Heavy Oxide ~- s.0 IJght Oxide
-- Metal
--- lower Pedestal Atmosphere _ 4.3 2.7 - -
- 4.0 2.4 - -
/
g q 2.1 -
, , f.-
) .,
-as 9, ,; e
~~
s
.s
~
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/
I ' , > ' ** v . g ,, N a t 8- ' f 'f ' -
, gj ; n_,
E i / i, i- - 2.s c O. I y 6 1.5- - l - p ; . I s --
. !! ,;...* / If
-20
,. 1 . pl . j l.
1.2 - ;; ,' , ,/ i g, ,i g l - j! ! .' / gj l'iI - 1.5
;; '- i 4l' gg O.9 - { l gj - i f . g' gl - -
8
- . - to (i
I
' {- l li-} t o.e - n i' ( ' -
j l ll
- 0,s .
. , _ _ , _ _ g- g
--- ~
.. {-l
-l o.a -
80.0 86.0 92.0 98.0 104.0 110 0 116.0 ' 122 0 128.0 134 0 140.0 j- TIME (103) s i Figure 5.3-15 Wetwell Pedestal Debris Temperatures. 5-203
20 , , , , , , , , , LASALLE WELCOR INTERWEDIATE TERM STATION BLACKOUT te- Radial -- eo
--- Vertical 66 1.6 - -
Pedestal Wall Thickness - 60 1,4 -
-,g k
v 8 .2-
--40 C.
bm - 35 N 3 d 1.0 . C 3 - 3.0 b 0.8 - -
$ - 2.6 e
0.8 - -- 2.0
- 16 0.4 - -
- - 1.0 s
0.2 - s ~
,- .' O.5 0.0 .
, -- ',~ ~ ~ ~ , , , , , ,
aa 80.0 B6.0 92.0 98 0 104.0 110.0 116.0 1220 - 12C.3 134.0 140.0 3 TIME (10 $ Figure 5.3-16 Maximum Concrete Penetration Distances
, 5-204
i as 25 0 i , , , , , i i i
-LASALLE MELCOR INTERWEDIATE TERM STATION BLACKOUT ca.s - Total -- au u Carbon Monoxide
--- ' Carbon Dioxide 20.0-Hydrogen '-- 4 6.o
--- Steam
- 4 0.0 '
17.5 - - -
/
- 35.0 15.0- 7E; Q
f-g ./- -- M p, - 30.0 - ] .
# M 3 ? R.5 -
w - O
/ C
$ / 25.0 -
e 5 10.0- / -
- 20.0 -
7.6 - -
- 15 0 1
~
6.0 - g
,j 10.0
,. ,s:
2.5 - #- ,# 1 -/
-~-"
0.0 -, , , . ,
~:s',xl;.w n .
.,3-. a
- 80,0 86.0 92.0 98 0-- 104.0 - 110.0 .116.0 122,0 128.0 134.0 - 14 0_f TIME (103) s Figure'5.3-17 Core / Concrete Interaction Cas Production
.5 205
, - - ~ . , . . . - . - . . - - . .- - .
~ . . . . . - . . .~.....g,.n+_ . . . . . . - . ,. . , ,
c i } i-i-- i i. + i i i f 2 e j 10 -r r -. , , , , i i i
~ ~
LASALLE MELCOR INTERMEDIATE TERM STATION DLACKOUT i d.
* ' *00 l 0 9 --
i 1 f n i 08- ..
- 1.75 '
I' - t 0.7 - J l ..
* - 1.50-j
/
- l. .
I 06- - I
-n ec ,.' -e
- g - 1.25-
-_g l M .' ~-
I 30.6- *
-.n.
/ _
o ( $ - l00 . O e n : . . X > 0.4 - / - i , l l 0.75-e : i 0.3 - * - (
- In-Vessel j
. Ex-Vessel-
- t
- 0.50 02- l- -
t l l / - 0.25 i 01- - - t f I O.0 'i- .r c -N ~ i i i~ i . , 80.C ~ 68.0 78.0 84.0 92.0 100.0 10 a.0 116.0 124.0 .132.0 . 140 0 .
. TIME (10$ s) i.
I a t Figure 5.3-18 Hydrogen. Production 1 k' i. 5-206 r
.-r. ...w- = -- w w . e .w-.. -
3 r+r
t 45 0 , , , , , , , , , , , , ., LASALLE WELCOR INTERWED! ATE TERM STATION DLACKOUT I
,a. 4.se .
40 0- Total Mass l Total Rate l 35.0-
'f - 3.2
,-l -
-l .n l ;- -
n 30 0- .' . - - 1,0 . g- : :: , - x n
.O
- .: : .4. *
-o
- . m-v 25 0- .i. i. -d 3
$ v - 0.8 ..- *U -
e ; ' :.;l ; ' Cd - X ., :. .
.m-
,. .m m.
fm20.0- l : - 2
; -u b : .
.: - c.a . 0' o
t o.a .
* ~ 4 Q e
- d 15.0- l
- M
.. : : i':
.:, , , - l' i
- i : ;
e
- 0.4
>t o.o - ,
, ~ , +;
. l .
... g,. . *-
f .
.~ n_g 5.0 - f . . -
, , , ..... i 0.0 , . , , , , . , , , , , m-- aa 0.0 10.0 20.0 30.0 40.0 60.0 60.0 70,C 80.0 904 100.0. 110.0 120 0 130.0 140 0 TIME (10 0). s s
Figure 5.3-19 Concrete Degassing Masses l 5-207 i
. . . . . _ - - . . . . . ~ - . - - . ~ . ... ~.. .. . - . ~ - . . - . - . - . - . - - . - -
f-k i i j .. i 4 9. 1-l i 800 0 1 r , i i i t m uo I LASALLE MELCOR INTERMEDIATE TERM STATION BLACKOUT X
-X- Drywel' Mass'-
i_ ,3o n . -+- Welwell Mass _ o ., _
.WW Pedestal Mass
!~
~ - - DW Pedestal Mass
! - - < - Downcorner Mase -
- -X- - DW Mole .
e00.0 - +- WW Mole x oo i ! 500.0- --- C.6 i -C tig s > .o -
*g
. ;g -
-% g.
j l-7M 430.0-e .
.-- 0.4
.[o
} M 'o - l-1
. 'A 300,0- . ji ,
~
-- 0.3 p
l c j ( ( . l I 200.0- -} , ~ ',', - -
-0.2
- j. , ,5b a . . , * ,. ...
+ - . -
---X..,+**',.y,;...ts N..
; ~X'.
100.0- ,.- 4 -*
-X. Oa
- p ,.
a:;{ n
..-A 1
0 0 d@ - 50.0 60.0 78.3 84.0 92,0
-f-
'300.0
' T - ? .-
108.0 ' Ile 0
-124.0 9
132.0- 140.0
^a-f L TIME (10ks) --
) r 1; (' 4 4-t )- Figure.5.3 20 Containment Hydrogen Distribution-1. 4 5_- 2 0 8 _- k, .
, w --,c .Nn v ..-men,, ,e ,. --a -- .t
r, 1.00 , , y i , , , , , LASALLE MElroR INTERMEDIATE TERM STATION DLACKOUT 2 00 o es- - -- DW Pedestal to Drywell -
--- WW Pedestal to Wetwell WW Pedestal to DW Pedestal /
O.76-Vent Exit /' VACUUM areakers ~ l Drywell Drain , ' ,. m .-- j_. - j / - 1.50 . e 0.54 - y r*' e
- e - +
g ,j o 1.25 c, 0 62 -
/
,f, .-.
O -!
% . /t
/
.j'l l
- 1 00 k k 0.40- / ,/ - ~$
l M '/ O 4 e / O o
. 0.75 ^
/ /
l
~-
8 r--" E 02e- i l i - p j j - o.so 1 ,
~ 0.16 - p .
l , J .! ; (
- 025 d 2
,.... . . ... . .... . 7 ,/... ....,...., 24 c.o4 - -
~'. -
, .',, f <
^
\
n
,x _,;3-; -~ .
\
. o_co
\
-0.os -
s
% --0 25
-02o , , , , , , , ., ,
50.0 60.0 76.0 84 0 . 92,0 - 100.0 108.0- 118.0 ' 124.0 132.0 .14 a.0 TIME (109 s) Figure 5.3 21 Containment Integrated Hydrogen Flow Rates 5-209
. n . .- -
2000 , , , , , r* 1 i "- i " LA5ALLE WELCOR INTERMEDIATE TERM STAT!oN DLACKOUT 180.0 8 Total Mass Flow --oo Ilydrogen Mass Flow
- - - - Ilydrogen Mole & action 100.0 -- 0.8 14 0.C ., -
0.7 120 0 -- 0.6 '. R 9
# E 100,0 - l - 0.5 [
e il m B0.0-
\ M
$ -- 0.4 2.
60.c - -- 0.3 40.0- , _ _
-- 02 s
\
\
20.0 g -- 0.1 A N 9 , 122.0 123 0 124.0 125.0 126.0 .127.0 1E0.0 129.0 130.0 131.0 132.0 TIME (103) s Figure 3.3-22 Wetwell Rupture Flow 5-210
v 20- , . . . . .. . , i I LASALLE MElf0R INTERMEDIATE TERM STATION BLACKOUT
~ ' "
W-Lower-Unit 2 Upper-Unit 2 - Umt1 Refueling Bay i og l Steam hnnel/'D!rbine Room 2 J 14 X Steam AnnelAustaine Room 1 I B
- o. 3 - -; -
3.2 l l-2 : o e4-oO : e 0 1.0 e 3
; C
$ 0.75- -$, -
Po n D c
- n. - 0.8 E o se-- :: -
6 ll - 0.e o.s' - :: -- oA8- l *
-- 04 o i- ll Is . .:.
n
! '% - 3 . w . = ;_, .= . ^02 ij
' ~ . , 1=.. .=.. .r " .= " "
. =. ra . = r- , -
g: a. ...- - o.ao-JX i 6 " 122.o s ea.o 124 o tas.o ra's.o ts.o -
;^
tae.o = tas o;
,'x ,
130 o bc:=-132.o tai o TIME (103) s Figure 5.3-23 Raactor/ Turbine Building Atmosphere Temperatures 211-q . . . . .
.-------1
0 65 ' ' 1.ASAl'.LE MELCOft INTERMCDIATE TEllM STAT 10M BLACKOUT O.80- - 0.55- - 0.50- - '
\
O.45- - 0.40- - M C o C 0.35- - O i f4 ) k o 0.30- -
"O 2 H2O 0.as - .... H2
--- 02 o.co - -----
--- CO ~
)
--- CO2-C.15 - -
1 0.10 -- - s ljj \ R.- ~ * '
~~~~
' 0.05- ~ b[ ,t **
k - . . I t u.: a . .: u . ~. .~. ~+ + .~. . a. . . w . _--
- _ _- ,, _~ _ _~ _~ _ _ _ -
0.00 i Uf , , -***t---- 122.0 123.0 124.0 '25.0 '26.3 127.0 128.0 129.0 - 130.0 131.0 .132.0 l: TIMh, '10 3 s)- i-l Figure 5.3-24-Upper Unit 2 Mole Fractions 5 212
0.20 ' ' -' LASAllE MELCOR INTERMCDIATE TEltW S1Efi6NLACKOUT r o.s a- ----- Lower Unit 2 - Upper Unit 2 Unit 1
- -- Refueling Bay ,
.t e - ~
l Steam Tunnel /lbrbine Room 2 ' X Steam 'Ibnnelfibrbine Room l' O.14 - to C o- - 3e 012-o "5 0.!0- - llE .i ,, ........ - .- ----- -. l ..- 5 s to ..' l 8 0.08- ,, l - o. b
~ l
.n , ~. ;
w
} *'
l 0.06- .- I s ,, ~ ~ . . ;; ... **,, t ,/. : ~:~ ~.-. ._,_,__ l 1 I ~~___~~~~~~L._:'_,_.,~~~~~-"~-_ , t 0.0 - -- M --
'/ :/
I .: 0.02- l t 1, " . -
.-t- l l~ {
'm-.l :
=
}
[ 0 00 ':,, " i i i i i , ,- , ,' - 122.0 123.0 - 124.0 -125.0 120.0 . 127.0 128 0 129.0 130.0 ' 13 % ,132.0 TIME (109 s) Figure 5.3-25 Reactor / Turbine' Bdilding flydrogen Mole Fractions - 5-213
. ~
10' 3-- , ---r--- i r , . i .> , a LASAILE 11 Elf 0R INTERMi'DIATE TERM ETAT 10N ULACKOUT
.: - - = . _ _ _:
~
- In-Vessel Total
-- In-Vessel hadioactive 3d _
Ex-Vessel Total
- - - , Ex-Vessel Radioactive !
~
18 : ~ ' j [
- 1 a :)^ ' : ./
--r i
= _ :
x -
/ : :
v , g , _
~ ~
$ ,)
l 2 )~ ~ ~ ' ' ~ ~ ~
~'*-~~~"''T~~'~-~*~~~~*-~~"
10~' t : f I /
'3 l ,' '
-]
f ,
+*
,,,. ] .j l ,
~l 10' * : 1 .* /- -: t S I / . .i 7 g
[ 's
~
l
~
7 ~ l l .-
.l 10 - ,
/- : .
1 l j ! 1 l < . o
/ ~
i t ? ,
' , /- p 10" . .
;j 60.0 ' 68 0 76 0 ~ 84 0 - 92 0 ' 100.0 108.0 ' - 118.0 ' :' 12 4.0 -' 132.0_ = 140.0 .-;
.. TIME (109 s)-
.~
- . 1; Figure 5,3-26 In and Ex-Vessel Releases of;Ceriura .from Fitel I
s l 214- ] i
-i lj
_. _ .-- _ ._ .-- - - - ---..-. - -.-- . -.- .: . . = - - + . - . . .. . - . , - .:
- . ~..,. -.~ ,- _ . - - - . .. n.~ - . . . . . . . ~ . . . . - _ . . . ~ . . . ~~. ...;.-.-
t-3.00 i , _, , , , , 1 LASALLE ME1IOR INTERMEDIAlB TERM STATION BLACKOUT
** ~
- . Concrete -
---- Steel e.so- -
2.25- - 2,00- - Q 1.75- - l' .M n 1A0- -- - M
<d 1
+
':ii 1.25-
- 1.00- * -
I' e l 0.75- ,. a
- l. 0.50- /. ..-
- l. :
l l
= '
0 25 i *
,a 0.00 , y- - . , , , , , , ,
i 80,0 68.0 70 0 84.0 92.0 100.0 - 108.0 -116.0 124 0 - 132.0 14 0.0 l:: TIME (10 3 s)- i l 1 - b l: Figure 5.3-27 Non-Radioactive Ex Vessel Releases 5-215 i
1 1 4 i 4 1.0 - - - -, m-.-, , ,- - - , - - _ , ,
- "~~
fk.- l
}l.
0.9 - Y.! j {' > LASALLE MELCOR INTER,ii'.' LATE TERM STATION DLACKOUT - l. f {, f
~- ~ ~ ~ -
06- - m . s c f, . , _ . _ . . _ . _ . _ - _ . _ . - . * ~ * - - - + - - - - ~ ~ ~ + -
- 0 g.'I' o i .f '
O j o 0.7 - -lD .f , . si n
- ^ X X X
. \
s .' I 8 1 08- .' . j b l' y V
'I Y
- O -
[D
$m C.5-c
.i
- 1 E. .: -
g 04-i l , , c ! I .
$ y .* .
N "~ ~U' ' p , 'h- o p_ i C ',
, l ... . g ) -
~
, .h
- -- Cs ) \ '
g 02 *----Pa ) V ( _.__j U ; Te 0.1 -
-"O '
~~ X-- Ru .
l - %- Cd i : -V- Csl ) 0.0 - M7 , X- ,
, , ,' s uym s 1 ,
80 0 66 0 70 0 84,0 92.0 100.0 108.0 118.0 1240 132.0. 140.0 TIME (100) s h Figure 5.3-28 Primary System Retuation Fractions 1 5-216
1 450.0 , , , , . l . LASA!1E ME140R INTERMEDTATE TERMg STki10N BLACKOUT I 400.0 .I
- I g i I
I I I 350.0- l - 3 % 1 I 1 300.0- ' I - I _i 1 l Q 250.0- f 1 gi X / i v
, 1:',
R I m C. '., , lll! 200.0- J I *, -
- _ i
' !:$.th
' ' Yi . --
Total Released Primary System 5k\ II ' i 150 0- : it
--- Containment -
'j '
~ . + - Reactor /IbrLine Bld li
, i , X Environment ii too.o - 8 i .i !.i . \ i : .. *! s
., g 50.0- -4 9 }
[ -- i N-s. t : o.a- +X
^;>
i,X
.s.
, , L+X, - , ;XV ;--LX - ,
60.0 68.0 76.0 84.0 92.0 100.0 108.0 116.0 124 0 - 132 0 140.0-TIME (103) 3 Figure 5.3-29 Xenon Mass Distribution 5-217 1 l
't
- O ' ' '
LASAEIE MELCOR INTERMCDI ATE TERM STATlos BLACKOUT 3 225.0- - ^ 200.0- - l l i - 175.0 - , \- 150 0- ' -
'S M
125 0- - E 'Ibtal Released - Primary System
- - - Containment 200 0- -
Reactor /lbrtine Bld i
+ Environment
, fA 75.0- ,7 g
/ '\
,-e g- ---- .
so.o- ,- .
-i.
- 2G.0 - _,
I
-. t.
00 lX , - jX , j X, , 'X , , ' 'j( . , ,
- 80.C 68.0 70.0 . 84 0 92 0 100 0 108.0 _.116.0 124.0 1300 140.0
- TIME (Ib3s) h I
( Figure 5.3 30 Cesium Mass Distribution 5 218 L
450.0- y , , , , , . LASALLE MElf0R INTERMEDIATE TERM STATION BLACKOUT-400.0 - - 350.0- -
**~*~
Ex-Containment 'Ibtal ~
----+
Environmental Release -
*~_,,...
Q 250.0-w ,.. ...** - w N e
. X 200,0-150.0- .' .
..' Xenon Class 103.0- ,
4 60.0- _ 4 , 00 , , , , , , , , , 122.0 123.0 '124.0 125 0 1260 127.0 128 4 129.0- :130.0 132.0 131.0 . TIME /103) s
~
Figure 5.3-31 Xenon Source Term to Enviroranent 5-219
2.2 , , , , , , , , , LASALLE MElrOR INTERMEDIATE TERM STATION BLACKOUT - to-
+
~
Cesium X Tellurium
** o Molybdenum ~
Ex-Containment Total
- Environmental Release o.8 - -
o.7 - -
'O x
T o6- - a a o6- -
, 4 .... ** '
o,3 &* c.2 - , .. ,... '. - - o.1 - .- --- N ,,,,,,,,,u n k n'"""""N'"" W',;3.......'. ,~~~~ g
, , . .vt t a. 127.o 144.o tis.o tie.o tiv.o tia.o sis.n ido.o t$1.0 tia.o TIME (103 ,)
a Figure 5.3-32 Cesium, Tellurium, Molybdenum Source Terms 5-220
0.0 i , , , , , , , , LASA'.LE MELCOR INTERMEDIATE TERM STATION Bl.ACKOUT e.o - - 7.0 - --
Ex-Containment Tbtal ~
Environmental Release
,1 .
Te 5.0- - M w g ,, M llE 4.0-3.0 - .' .
..- "' DariuIn 2.0 - .- .
1.c - ..- . V .- 0.0 , , , , , , , ., 7 122.0 - 123.0 124.0 -125.0 126.0 127.0 128 0 1290 130.0 131.0 132.0 TlME (100) s
,i Figure 5.3-33 Barium-Source Term-
. 5 - 2 21 --
. ..m. . . - _ . _ . . . . _ .-,_,_ ,-_-._.~.._..,_.m.,,,_-.m.,_m._..,_._..m.. ..m.-. __.u..... ._m. ..__s..,_.
l'. 1 i= i i 1~ l 4 i-i 1 l 0.300 i i , , , i i i i i LASALLE MELCOR INTERMEDIATE TERM STATION BLACKOUT ~ j o.275- - 4 } _. !- 0'860~ ~ Ex-Containtnent Tbtal Environmental Release j P.225 - j i-- l 4 1 I .i
- 0.200- - i I
i t t l~ 9 0.175 - - 4
! x m
I O C.150 - - I O 'L I m , n .,' h j 0.125 - - l 1 0.100- -
-lodine *-
i 0.075 - ..a* a .*
,e*,<*.
t i 0.050- ,.(~. - 4 4 0.025- ,- - 0.000 , . I d' , , , , , ' ,
, 122,0 123.0 124.0 125 0 ' l28.0 127.0 128.0 ^ 129.0 .: 130.0- '131.0 ~ 132.0 -
$ TIME (100) s Figure 5.3-34 Iodine Source Term- . 5+222
-.w' - - w, y - -- a n r.-, -
i .-- .
- , . .-. .- . .....~. . . . - - -. . .- . . .... ~ . . . _ . . - . - . . . . - - . . , , . . .-
I l l l ! I3 i s i i i . i i i e LASALLE MELCOR INTERMEDIATE TERM STATION BLACKOUT. 1.z - Ex-Containment Total 3,3, Environmental Release - ., t.o - - - 0.0 - - 0.8 - - To M co o7- - - - -
-I
.. o 0.6 - -
W c.5 - -
'4-
-Ruthenium -
0.3 - . 0.2 - - 0.1 - '" ***... .. .... . 0.0
, , , , , , , , ,}}