The document discusses potential strategies for decommissioning the damaged Fukushima Daiichi nuclear power plant reactors and spent fuel pools, referred to as "Plan-B Liquidation Strategies". It notes that dismantling the damaged reactors will be extremely challenging due to the harsh radiation environment. Leaving the reactors and spent fuel pools as they are is proposed as one option, but it would require public and regulatory approval.

1. “ Plan-B” An Alternative Liquidation* Strategy of Fukushima Daiichi NPP May 21, 2011 Satoshi Sato [email_address] International Access Corporation *: A term “liquidation” is used in this document to generally mean various activities directly and indirectly associated with restoration of safe state of each affected reactor in Fukushima Daiichi NPP. This follows a precedent in which workers involved in the emergency actions on the Chernobyl site during the accident and the subsequent clean-up operations were called “Liquidators”.

2. Abbreviations Standby Gas Treatment System SGTS Emergency Procedure Guidelines EPG BWR Owners Group BWROG Fission Product FP Overhead Crane OHC Fuel Handling Machine FHM Core-Concrete Interaction CCI Reactor Pressure Vessel RPV Spent Fuel Pool SFP Independent Spent Fuel Storage Installation ISFSI Greater Than Class C (Cask for High Level Radiation Waste) GTCC Spent Nuclear Fuel SNF Nuclear Power Plant NPP

8. Summary - Reactor Original Function as Secondary Containment totally lost due to H2 explosion on Refueling Floor. Remaining part of building still reasonably good. Overhead Crane (OHC) and Fuel Handling Machine (FHM) not available. Barrier integrity no longer maintained. Details not confirmed. Barrier integrity no longer maintained. Bottom Head Penetrations severely damaged. Totally destroyed 1 Function as Secondary Containment still reasonably maintained even after H2 explosion. OHC and FHM still fully functioning. Barrier integrity severely degraded due to H2 explosion inside or outside Torus. Ditto Ditto 2 Ditto Not affected Not affected Empty 4 Same as Unit 1, except that some portions lower than Refueling Floor also degraded due to H2 explosion Same as Unit 1 Ditto Ditto 3 Reactor Building Primary Containment Reactor Pressure Vessel Reactor Core Unit Severely damaged Severely damaged Function Severely damaged Possibly still partly maintained but not confirmed Barrier Integrity

9. Unit 4 Unit 3 Unit 2 Unit 1 State of Reactor Building, Unit 1 to 4 looking from east as of March 20

10. Unit 3 Unit 4 State of Reactor Building, Unit 3 and 4 looking from west as of March 20

12. 12~13m ~33.5m ID 8.9m ~46m ~23.5m ~15m ~11m ~40m 16~17m Refueling Floor Rx. Bldg. (Secondary Containment) Primary Containment Reactor Pressure Vessel Suppression Chamber (part of Primary Containment) Drywell Pedestal Typical Configuration （ Unit 3, 4 ）

13. Melt-Down through Core Plate Predicted to occur 2 hours following complete loss of cooling capability. Several previous experiments suggested steam explosion not likely. Core Shroud Core Plate Reactor Core Molten Core Water No Doubt

14. Further Melt-Down through Core Plate Actual Complete Loss of Cooling Capability (Official Announcement by Government on May 16, 2011) No Doubt 06h43m 3 06h29m 2 14h09m 1 Duration Unit

15. Degradation of Reactor Pressure Vessel Bottom Head Creep rupture begins to occur at ~240-deg C below melting point (1500-deg C) of vessel material (low alloy steel), allowing some leakage of highly contaminated water containing fractured pieces of fuel pellets.       Highly Likely

16. Locations of Potential Leakage (Typ.) Vulnerability of Bottom Head Leakage

17. Further Degradation of Reactor Pressure Vessel Bottom Head           Drywell Sump Pit Pedestal Pedestal Doorway Possible

18. Major Degradation of Reactor Pressure Vessel Bottom Head and Core-Concrete Interaction (CCI), Resulting in Significant Amount of Release of Radioactive Aerosol       Pedestal Doorway Pedestal       H 2 O, CO 2 H 2 O, CO 2 H 2 , CO Aerosol Aerosol Aerosol Aerosol Not very likely, but could have happened depending on cooling evolution during early stage.

19. Beginning of Primary Containment Failure       Pedestal Doorway Pedestal       Aerosol Aerosol Aerosol Aerosol H 2 , CO H 2 O, CO 2 H 2 O, CO 2  Not very likely, but could have happened depending on cooling evolution during early stage.

20. Pedestal Doorway Pedestal Wall Source: NUREG/CR-6042 Rev.2

21. Beginning of Primary Containment Failure Aerosol Not very likely, but could have happened depending on cooling evolution during early stage.

22. Not likely Failure due to Creep Rupture Gross Failure of Primary Containment due to Steam Explosion

23. Gross Failure of Primary Containment due to Melt-Down Aerosol Aerosol Aerosol Aerosol Not likely

24. Complete Melt-Down through Man-made Rock (Basemat) Not likely Man-made Rock

25. Residual Heat Generation 2 months after shutdown 0 2,381 2,381 1,380 Thermal Output (MWt) 0 2.4 2.4 1.4 Estimated Residual Heat (MWt) 0.1% of rated Thermal Output 0 784 784 460 Electrical Output (MWe) 0 548 548 400 # of Fuel Assembly in Rx. 4 3 2 1 Unit

27. Unit 4 SFP Top view of fuel rack by remote underwater TV camera. Difficult to draw any conclusion about fuel integrity only based on this information. Fuel inspection by “sipping” is warranted.

28. Residual Heat Generation 11/30/’10 6/19/’10 9/16/’10 3/25/’10 Hottest Spent Fuel Discharged (Date of beginning of last refueling outage) Spent Fuel Pool New Fuel Storage Vault 1331 514 587 292 1.8 0.23 0.46 0.07 Estimated Residual Heat Generation Rate (MWt) 204 52 28 100 Number of Fuel Assembly 4 3 2 1 Unit

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31. Drywell Flooding

34. Delayed wall creep rupture would eventually occur in the vicinity of gas pocket.

38. Steam Dryer

39. Moisture Separator

40. Fuel Assembly

41. Lower Plenum (Region below Core Plate)

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54. 86% completion as of April 14, 2004

55. Implosion on Containment Bldg., September 17, 2004 97% completion as of January 19, 2005

56. Essentially 100% completed, as of May 5, 2005 “ Green Field” achieved on July 25, 2005

57. Implosion Technique Applied for Turbine Bldg.

58. ISFSI Pad and Spent Fuel Storage Casks Vertical Horizontal

60. Back to “Green Field” as of 9/5/2007 Actual and Future Yankee Rowe Decommissioning Schedule

61. Cased in container on 11/20/1996 Departed from site on 4/27/1997 Loaded on to railcar for 1800km transportation Arrival at Barnwell Site for subsurface repository on 5/7/1997 Reactor Vessel Disposal 3.6m-dia. x 8.1m-tall, weighing 165tons 80 tons of concrete poured inside and outside vessel

62. Large volume of subsurface soil found contaminated with tritium (H-3). Numbers indicate H-3 concentration in groundwater in pCi/L. EPA drinkable level is 20,000pCi/L.

64. Basically just cooling-down Dismantling Activities

65. 30-year long project! Finally Back to Green Field in 2028 SNF Disposition Campaign

67. Swedish Plan (active)

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71. Note that the dose limit for Emergency Workers is 250mSv. 8,400 18,200 46,500 mSv/h 3 2 1 Unit Possible for all buildings other than Rx. Bldg. of Unit 1 to 3 after some decontamination efforts. Not practical for Rx. Bldg. of Unit 1 to 3 due to high contamination level. Implosion Already done for Unit 4 Rx. Not practical for other units due to too much activity load. Chemical Decontamination Possible for Unit 4. Not practical for other units due to harsh radiological environment for workers. (Drywell Dose Rate as of May 20, 2011.) Separate RPV from All Other Connecting Systems Already done for Unit 4. Not practical for other units due to high contamination level. Remove Rx. Internals by High Pressure Abrasive Water Jet Application for Fukushima NPP Units Technique

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76. Radionuclide Releases During Chernobyl Accident Source: Chernobyl – Ten Years On (OECD/NEA)

77. Source: NUREG/CR-6042 Rev.2 1MCi = 37,000TBq Daily Release During Chernobyl Accident

78. Source: OECD/NEA “Chernobyl Ten Years on Radiological and Health Impact – An Assessment by the NEA Committee on Radiation Protection and Public Health” November 1995 Cs-137 Contamination 10 years later Vicinity of Fukushima NPP 80km Equivalent dose rate of 555kBq/m 2 contamination is 1.8 μ Sv/h or 15.8mSv/y. Blue colored region on land represents dose rate greater than 0.3 μ Sv/h as of 3/19/2011. 500km

79. Access Restriction due to High Level Contamination

81. Radiological impact estimated by various organizations Japan - estimated cumulative dose in mSv through 3/11/2012

84. Residual Heat Generation 2 months after shutdown 1 hour after shutdown 0 2,381 2,381 1,380 Thermal Output (MWt) 0 24 24 14 Estimated Residual Heat (MWt) 1% of rated Thermal Output 0 784 784 460 Electrical Output (MWe) 0 548 548 400 # of Fuel Assembly in Rx. 4 3 2 1 Unit

85. Major Degradation of Reactor Pressure Vessel Bottom Head and Core-Concrete Interaction (CCI), Resulting in Significant Amount of Release of Radioactive Aerosol       Pedestal Doorway Pedestal       H 2 O, CO 2 H 2 O, CO 2 H 2 , CO Aerosol Aerosol Aerosol Aerosol

86.       Pedestal Doorway Pedestal       Aerosol Aerosol Aerosol Aerosol H 2 , CO H 2 O, CO 2 H 2 O, CO 2  Beginning of Primary Containment Melt-Through

87. Gross Failure of Primary Containment due to Melt-Down Progression Aerosol Aerosol Aerosol Aerosol

88. Gross Man-Made Rock (Basemat) Melt-Through Man-made Rock

89. Source: NUREG/CR-6042 Rev.2 Various Gases and Debris Generated during CCI

90. Source: NUREG/CR-6042 Rev.2 Breakdown of FP Species

92. Radioactive Decay after 70 days I-131, Cs-134, Cs-136, Cs-137, Rb-86, Te-127m, Ba-140, Sr-89, Sr-90, Co-58, Co-60, Ru-103, Ru-196, Am-241, Cm-242, Cm-244, Nb-95, Nd-147, Pr-143, Y-91, Zr-95, Ce-141, Ce-144, Pu-238, Pu-239, Pu-240, Pu-241 32 species gone, 27 species left. 9.5 x 10 -8 3 days 2.9 x 10 -11 2 days 8.5 x 10 -22 1 day Remaining Half-Life

93. Gross Failure of Primary Containment Melt-Through This would be very slow even if it does take place at all.

94. Complete Melt-Down through Man-made Rock (Basemat) This is even more unlikely. 12~13m

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102. Specific Strategies Strategy-A SFP Not Affected Strategy-II 2 Strategy-B Strategy-I Reactor Systems Not Affected 4 Strategy-B Strategy-II 3 Strategy-B Strategy-II 1 Spent Fuel Pool Reactor Unit

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111. Helium is a standard cooling medium for high temperature gas reactors. GT-MHR (Gas Turbine – Module Helium Reactor)

112. A* B* To be added Scrubber/Gas Cooler Ventilation System Mode-1/2 Heat Sink Gravel Flow from Suppression Chamber to Drywell *: See “proposed line-up” for system interfaces for A and B for each unit. Option A

114. Copper Sphere Shell Zeolite Mixing several different constituents may be considered

115. Flow from Suppression Chamber to Drywell

116. Field Assembly of Primary Containment at Browns Ferry Site During Construction Time

117. Unit 1 Core Spray System Helium/Air Injection Point Proposed System Lineup A

118. Unit 1 Shutdown Cooling System B This valve may not be opened.

119. Unit 1 Isolation Condenser (Alternative Option) B X

120. Unit 2/3 Core Spray System Helium Injection Point A

121. Unit 2/3 High Pressure Injection System B

122. Unit 1 Atmospheric Control System (Alternative Option) B X

123. To be added To be added Scrubber/Gas Cooler Ventilation System Rx. Bldg. Truck Bay Mode-1/2 Option B1

124. To be added To be added Scrubber/Gas Cooler Ventilation System Rx. Bldg. Truck Bay Blower Mode-1/2 Option B2

125. To be added To be added Scrubber/Gas Cooler Ventilation System Rx. Bldg. Truck Bay Mode-2 Option C

126. Rx. Bldg. Truck Bay Stack Air Gap for Flow Path Air Flow only by Natural Convection Mode-3 See detail “D” See detail “E”

127. Detail “D”

128. Detail “D”

129. Construction Details of Bottom Portion of Primary Containment Vessel (Oyster Creek) (2) Detail “E”

130. Construction Details of Bottom Portion of Primary Containment Vessel (Oyster Creek) (1) Detail “E”

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138. Spent Fuel Racks Spent Fuel Racks Gate Cask Pit A A Spent Fuel Pool (top view) N

139. Water Level Spent Fuel Racks Spent Fuel Racks Spent Fuel Pool (side view) A-A

140. Finned Heat Sink Chambers (Copper) Step-1 Install Finned Heat Sink Chambers on Spent Fuel Racks.

141. Cross-Tie Pipes

142. Step-2 Install Pre-fabricated Pipe Modules.

143. 35cm Cold (Inlet) 2-inch Sch#40 Stainless Steel Hot (Outlet) 2-inch Sch#40 Stainless Steel Convection Cooling 2-inch Copper Pipe Modules

144. Φ10mm (typ. 4) 50mm Approx. 2000mm Approx. 8000mm

145. A A View A-A Main Header Main Header Main Header Distribution Header Top View

147. Step-3 Load Heat Sink Gravel Water Gravel

149. Copper Sphere Shell Zeolite Mixing several different constituents may be considered

150. Water level gradually decreases Step-4 Start Ventilation System

151. To Ventilation Fan and Gas Treatment System Wet Scrubber Water level Operation Mode 1, and Mode 2

152. Operation Mode Favorable Thermal Characteristic of Helium Not Required Required Required Cleanup System 3 2 1 Mode Natural Convection Non-contaminated Air Low (<100kW) Forced Cooling Contaminated Air Medium (100-350kW) Forced Cooling Contaminated Helium High (>350kW) Power Cooling Medium Heat Load J/kg ･ K Air = 1 W/m ･ K 1012 1 0.0316 Air 5192 5.53 0.1663 He Heat Capacity Relative Thermal Conductivity Thermal Conductivity Medium

153. Operation Mode 3 “Natural Convection” Inlet Sleeve Shielded Air Intake Block

159. Ultimate Configuration with Operation Mode 3 All contaminated equipment and materials are permanently buried in-situ. Paradigm Shift !! This concept, in spite of huge cost benefit expected, significantly deviates from the conventional approach.

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162. Water Treatment (1) Highly Contaminated Water Currently Stored in Various Pools at Site Vitrification Canisters On-site Repository Concentrated Radioactive Liquid Treatment System Cement Aggregate Contaminated Concrete Rubble (Optional) Processed Water (slight contamination allowed) Ready-Mixed Concrete < 5,000Bq/cm 3 < 0.065mSv/h 20 v/v% 80 v/v% For Entombment Work

163. Dose Rate Calculation of Homogenously Contaminated Concrete Assumption: 500TBq in 10 5 m 3 , or 5,000Bq/cm 3 of processed water Water Content in Ready Mixed Concrete = 20% Calculation: Low enough!

164. Heat Generation Calculation of Homogenously Contaminated Concrete Assumption: 500TBq in 10 5 m 3 of water, or 5,000Bq/cm 3 Water Content in Ready Mixed Concrete = 20% Energy Release per Disintegration = 1MeV Calculation: Total energy release rate = (1.6 x 10 -13 J) x (5 x 10 14 /sec) = 80W Temperature increase based on black body radiation q” = σ T 4 σ = 5.67 x 10 -8 q” = 80/(4 π r 2 ) r = 28.8m q” = 7.7 x 10 -3 W/m 2 T = 19-deg C Low enough!

165. Water Treatment (2) Desalination System Cement Aggregate Contaminated Concrete Rubble (Optional) Processed Water (still slightly contaminated) Ready-Mixed Concrete < 5,000Bq/cm 3 < 0.065mSv/h 20 v/v% 80 v/v% Contaminated Water within Intake Area For Encasing Concrete Rubble

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168. Intake Area Intake Facility (typ.) Backwash Valve Pit (typ.) Control Bldg. (typ.) Turbine Bldg. (typ.) Rx. Bldg. (typ.) RW Bldg. (typ.) Before Unit 2 Unit 1 Unit 3 Unit 4

169. Entombed Reactors After Protection Fence against Aircraft Impact Stack Concrete rubble generated from demolition of all other structures is encased in the large concrete block(s). Tsunami Barrier Tsunami Barrier Wave Breakers for Tsunami Protection Original Shoreline

170. ISFSI for SNF and any potential GTCC Waste On-Site Repository for Vitrified Canisters Legend: Monitoring Post Ground Water Sampling Point Protected Area Main Gate New Site Boundary Monitoring Facility Conceptual New Site Arrangement

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173. But, man-made structures may not be too bad… Possibly good for centuries or even millennia!

174. Source: “The Future of Nuclear Power” (MIT) Residual Heat 1/20

175. Source: “The Future of Nuclear Power” (MIT) Radioactivity 1/100

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177. District for New Industry/Community Development Entombed Reactors (Units 1 to 4) Survived Reactors (Units 5 and 6) Solar Thermal Power Beyond “Liquidation” Previous Site Boundary

178. Target Overall Schedule FS, Bidding, Design/Engineering Mode 1 Mode 2 Mode 3 Mode 1 Mode 2 Mode 3 Sipping Transportation Campaign Unit 2, SFP Unit 1 to 3, Reactor Construct On-Site Repository Facility Unit 1, 3, and 4, SFP Unit 4, Reactor 4y 6y Public Acceptance (Workshop) 10y 8y 2y Activities EI 2 -D Projects Demolition of other structures Water Treatment, Vitrification Build Liquidators’ Villages Recruit Liquidators Expand On-Site Liquidation Infrastructures New Industry/Community Development Construct ISFSI (for Unit 2 SNF) Construct Intake Area Tsunami Barriers Licensing Review on EI 2 -D (Safety Analysis)

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182. Time Cooling Efforts Abandoned after Plant Shutdown Release (NG, I, Cs) CCI Penetration Depth Required Evacuation Radius

184. Temperature Monitoring Probes

185. To be added Rx. Bldg. Truck Bay