TRANSPORTATION CASK AND CONCRETE MODULE DESIGN FOR MANAGING NUCLEAR SPENT FUEL PRODUCED IN BUSHEHR NUCLEAR POWER PLANT A.M. TAHERIAN Iran Radioactive Waste.

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TRANSPORTATION CASK AND CONCRETE MODULE DESIGN FOR MANAGING NUCLEAR SPENT FUEL PRODUCED IN BUSHEHR NUCLEAR POWER PLANT A.M. TAHERIAN Iran Radioactive Waste Management Company Co. (IRWA) ataherian@aeoi.org.ir

Spent Fuel Management Policy Fresh Fuel Closed Fuel Cycle Nuclear Power PERCEPTION POLICY (Asset ) Recycle SPENT FUEL Do & See (Liability) Disposal Open Fuel Cycle 2

Spent Fuel Management Steps

Power Plant and Fuel Specification Bushehr Nuclear Power Plant: Officially opened: Sep 2011 Commercially operated: Sep 2013, Russian type VVER 1000/446, Reactor core: 163 fuel assemblies fuel assemblies fuel life time: 3 up to 4 years Up to now: 3 refueling in reactor core (3*163 fuel assemblies)

Source Terms Calculation Considerations for determining source term: 47000 MWD/MTU burnup, 4.2 % enrichment, 3 up to 4 years’ fuel life time, 3 and 6 years cooling time

Source Terms Calculation Energy (MeV) Gamma Flux (photons/sec) After 3 years After 6 years 0.01~0.05 4.85E+15 2.1E+15 1.33~1.66 3.03E+14 1.06E+15 0.05~0.10 1.10E+15 4.82E+14 1.66~2.00 1.49E+14 2.22E+14 0.10~0.20 1.14E+15 5.72E+14 2.00~2.50 6.70E+12 6.45E+12 0.20~0.30 9.54E+14 4.12E+14 2.50~3.00 1.94E+13 1.51E+12 0.30~0.40 6.52E+14 2.72E+14 3.00~4.00 1.75E+11 6.93E+10 0.40~0.60 6.83E+14 3.00E+14 4.00~5.00 2.07E+10 8.98E+09 0.60~0.80 5.42E+14 2.21E+14 5.00~6.50 6.18E+04 48270000 0.80~1.00 3.04E+14 1.25E+14 6.50~8.00 7.00E+03 5567000 1.00~1.33 3.70E+15 4.24E+15 8.00~10.0 7.98E+02 639600 Total Gamma Flux: 1.44E+16 photons/sec for 3 years cooling time 1.04E+16 photons/sec for 6 years cooling time PHOTON FLUX SPECTRUM AFTER 3 AND 6 YEARS COOLING TIME FOR 12 FUEL ASSEMBLIES

per fuel element (neutrons/sec) Source terms (α, n) reaction Spontaneous fission Radionuclides per fuel element (neutrons/sec) After 3 years After 6 years PU-238 3.50E+06 3.45E+06 PU-240 1.22E+06 PU-239 1.16E+05 PU-242 9.97E+05 2.31E+05 2.32E+05 CM-242 2.54E+06 1.34E+05 AM-241 4.60E+05 6.89E+05 CM-244 1.21E+09 1.12E+09 AM-243 3.30E+04 CM-246 1.57E+07 5.24E+05 2.77E+04 CF-252 7.23E+06 4.27E+06 CM-243 4.56E+04 4.34E+04 1.01E+07 9.31E+06 NEUTRON SOURCE DUE TO (α, n) REACTION AND SPONTANEOUS FISSION AFTER 3 AND 6 YEARS COOLING TIME FOR 12 FUEL ASSEMBLIES

Source Terms Calculation Thermal power results: 2.13 kW after 3 years cooling time, considered for cask design. 1.2 kW after 6 years cooling time taken into account for module design. Thermal power for spent fuel assembly after removing from the core

Principal Specifications of Cask and Canister Component Value Basket Thickness (cm) 2 Material Borated Stainless Steel Canister Cask (inner body) 1.5 33.5 Inner diameter (cm) 130 133 Outer diameter (cm) 200 Thickness of Top Lid (cm) 25.4 Thickness of Outer Top Lid (cm) 26.6 Thickness of Bottom Lid (cm) 20 Carbon Steel Stainless Steel Neutron absorber Cask (Outer Shell) 12.7 225.4 229.4 Ethylene Glycol (67%)

Dose Rate and Criticality Calculation for Canister and Cask Distance Dose rate (mSv/hr) surface 5.23E+05 1 m from the surface of the Canister 1.74E+05 2 m from the surface of the Canister 8.22E+04 dose measurement Location Gamma dose rate (µSv/h) Neutron dose rate Total dose rate Outer surface of Cask 10.10 1.08E-01 10.21 2 m from surface of Cask 4.54 2.02E-02 4.56 The neutron multiplication factor and its relative error were obtained about 0.31463 and 0. 019% for cask filled by the air.

Dose Rate and Criticality Calculation for Canister and Cask

Designing Concrete Module Item Value (cm) Base/roof width 417 Base/roof length 652 Base upper side wall thickness 30.5 Base upper rear wall thickness Roof thickness 135 Rear/end (side) shield wall thickness 114.44 Rear/end (side)/corner shield wall height 564 End (side) shield wall length 297 Corner shield wall width (square) Door inner steel plate thickness 2.5 Door concrete thickness (centerline) 50,67

Dose Rates in 5 Directions in Concrete Module Location Distance (m) Dose rate (mSv/hr) Before adding another module Direction No. 1 On the surface 0.011 1 m from the surface 0.027 2 m from the surface 0.030 Direction No. 2 0.0007 0.0005 0.0004 Direction No. 3 4.3399 2.1914 1.2152 Direction No. 4 0.00010 0.00194 0.00421 Direction No. 5 0.0043 0.1861 0.1923

Dose Rates in 5 Directions in Concrete Module Location Distance (m) Dose rate (mSv/hr) Before adding another module Direction No. 3 On the surface 4.3399 1 m from the surface 2.1914 2 m from the surface 1.2152 Direction No. 5 0.0043 0.1861 0.1923

Dose Rates in 5 Directions in Concrete Module Location Distance (m) Dose rate (mSv/hr) Before adding another module After adding another (new) module Direction No. 1 On the surface 0.011 1 m from the surface 0.027 2 m from the surface 0.030 Direction No. 2 0.0007 0.0005 0.0004 Direction No. 3 4.3399 0.0321 2.1914 0.0206 1.2152 0.0154 Direction No. 4 0.00010 0.00194 0.00421 Direction No. 5 0.0043 0.1861 0.1923

Summary of Results and Conclusions Determination of the source term for spent nuclear fuel, designing the canister and cask for transportation, designing the concrete module for storing spent fuel were carried out using Origen 2.1 and MCNPX 2.6. Canister: thickness of the stainless steel: 2 cm Dose rate on the canister surface: 3.95E+08 mSv/hr on 1 meter from the surface: 1.32E+08 mSv/hr 2 meters from the surface:6.24E+07 mSv/hr at the top of the canister: 7.07E+03 mSv/hr Thermal power: for each spent fuel assembly after 3 years: 2.13 kW for each spent fuel assembly after 6 years: 1.2 kW

Summary of Results and Conclusions Cask: The thickness of carbon steel: 35.5 cm The neutron multiplication factor: 0.31463 for the cask filled by the air Dose rate on the cask surface: 10.21 µSv/h 2 meters from the surface: 4.56 µSv/h Concrete module inlet and outlet vents have a crucial role in the dose rate around the module maximum dose rates on the surface: 0.011 mSv/hr for front side (door side) maximum dose rates 2 meters from the surface: 0.192 mSv/hr for roof side The relative errors were less than 10% for all calculations.

Future Activities Simulation of fire test, drop tests (9m and 2m) and submersion in water for transportation Cask Study of Dual purpose cask (DPC) as a alternative option