ML22063B117
| ML22063B117 | |
| Person / Time | |
|---|---|
| Site: | 07103036 |
| Issue date: | 12/31/2021 |
| From: | Kyoto Univ |
| To: | Office of Nuclear Material Safety and Safeguards |
| Bernie White NMSS/DFM/STL 301-415-6577 | |
| Shared Package | |
| ML22063B110 | List: |
| References | |
| EPID L-2020-DOT-0000 | |
| Download: ML22063B117 (196) | |
Text
()-B Thermal analysis
()
()-B Thermal analysis B.1 General description This analysis shows that this package maintains the integrity and satisfies the thermal performance under normal and accident conditions specified in IAEA Regulation.
This packaging is dry type. The package is transported by vertically fixed on the tie-down device. Consequently, the thermal analysis is carried out as the package is located as vertically.
B.1.1 Thermal design The configuration of this packaging is shown in ()-Fig.B.1. As shown in this figure, this packaging consists of the main body, inner lid, fuel basket, and outer lid. Since the fuel basket No.2 has the same shape as the fuel basket No.1 and there is no significant difference. Therefore, the fuel basket No.2 is handled in the same manner as the fuel basket No.1.
The design features of this packaging are described below.
(1) There are 21 types of fuel elements and so on as shown in the paragraph D of section (). The heat generation from the radioactive contents is ignored in this analysis, since the decay heat generated from unirradiated fuel elements are negligibly small.
(2) Heat transmission (Refer to ()-Fig.B.2)
(a) Heat gain from the surface of package consists of solar insulation and heat during fire under accident conditions.
(b) The heat on the external surface of package is transmitted into the internal surface of inner shell or inner lid by conduction.
(c) The heat on the internal surface of inner shell or inner lid is transmitted to the external surface of fuel basket by natural convection and conduction.
(d) The interior of the basket is not taken into account in the thermal analytical model, the temperature on the outer surface of the basket
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represents the temperature on the fuel elements and basket on the assumption that insulation is effective.
i) No temperature gradient occurs since minimal heat is generated in the basket under normal test conditions.
ii) Only external heat affects the package under accident test conditions, rendering the external maximum temperature higher than the internal maximum temperature.
(3) The balsa used as shock absorber maintains its insulating characteristics even under accident test conditions.
(4) the outer shell and the external sheet have fusible plugs through which any vapor or gases emitted by the shock absorber and heat insulator under accident test conditions are discharged, preventing the inner pressure from rising.
(5) The O-ring provided on the inner lid to maintain the leak tightness of the packaging is protected from the heat resulting from fire under accident test conditions by the heat insulation effect of the heat insulator and shock absorber.
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()-Fig.B.1 Component of packaging
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Unit (mm)
()-Fig.B.2 Concept of thermal transmission
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B.1.2 Conditions and methods of thermal analyses (1) Conditions of thermal analyses
()-Table B.1 shows the thermal conditions used for the normal and accident test conditions.
()-Table B.1 Conditions of thermal analyses Condition Accident test Normal test conditions conditions Before During After Item fire fire fire Decay heat 0 0 0 0 0 0 Environmental Ambient 38 38 -40 38 800 38 conditions temp. Stagnant Stagnant Stagnant Stagnant 30min Stagnant air air air air air Solar rad.
No Yes No Yes Yes Yes heat Ambient rad. 1.0 1.0 1.0 1.0 0.9 1.0 factor Radiation factor for 0.4 0.4 0.4 0.4) 0.8) 0.6) packaging surface a): Surface radiation factor for steel (SUS304) not exposed to fire.
b): Surface radiation factor for steel (SUS304) being exposed to fire.
c): Surface radiation factor for steel (SUS304) exposed to fire.
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- 2) Methods of thermal analyses
()-Table B.2 shows the methods by which thermal analyses are performed.
()-Table B.2 Methods of thermal analyses Item Description Specifications of contents See Section D of Chapter (I) of fuel elements Maximum decay heat (W) 0 Calculation model Packaging Axially symmetric two-dimensional model Contents Temperature calculation Simplified analyses TRUMP, non-steady state thermal analysis code(see B.6.2)
Physical properties used See Section B.2Thermal Properties of the (thermal properties) Materials.
- :Under normal test conditions.
- Under accident test conditions.
B.2 Thermal properties of the materials The materials used for the package are described in Chapter I.
Of these, the materials shown below were used in the thermal analyses.
Stainless steel Air Shock absorber (balsa)
Heat insulator (hard polyurethane foam).
This section will describe the thermal properties of these materials.
(1) Stainless steel The thermal properties of the stainless steel used are shown in (1)
()-Table B.3 Stainless steel is used as the main structural material for the principal elements of the packaging.
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()-Table B.3 Thermal properties of stainless steel Specific weight 7.875g/cm3 Temperature Specific heat Thermal conductivity
() (kJ/kgK) (mW/mK) 50 0.469 1.477x104 100 0.490 1.558x104 200 0.519 1.697x104 400 0.561 1.953x104 600 0.594 2.232x104 800 0.640 2.488x104 (2) Air (2)
()-Table B.4 shows the thermal properties of the air used.
()-Table B.4 Thermal properties of air Specific weight 9.16x10-4 g/cm3 Temperature Specific heat Thermal conductivity
() (kJ/kgK) (mW/mK) 0 1.005 24.07 40 1.009 27.21 100 1.013 31.63 140 1.017 34.54 200 1.026 38.61 500 1.093 56.17 800 1.156 70.94
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(3) Shock absorber (balsa) 6)
()-Table B.5 shows the thermal properties of the shock absorber (balsa).
This material, which is used as the shock absorber in the upper and lower part of the packaging, has a heat insulation capability.
()-Table B.5 Thermal properties of shock absorber (balsa)
Specific weight 0.16 g/cm3 Temperature Specific heat Thermal conductivity
() (kJ/kgK) (mW/mK) 0 1.750 187.2 50 1.695 187.2 100 1.796 175.6 150 1.988 200 1.905 195.5 250 1.955 275 1.867 255.8 320 1.453 255.8 350 0.917 255.8 500 0.130 255.8 900 0.071 255.8
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(4) Heat insulator (hard polyurethane foam)
()-Table B.6(3) shows the thermal properties of the heat insulator (hard polyurethane foam).
()-Table B.6 Thermal properties of heat insulator (hard polyurethane foam)
Specific weight 0.04 g/cm3 Temperature Specific heat Thermal conductivity
() (kJ/kgK) (W/mK) 20 1.193 0.535 50 1.402 0.581 100 1.645 0.675 250 1.859 0.937 300 1.344 0.937 400 0.193 0.937 800 0.151 0.937
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B.3 Specifications of components The following components are taken into account in the thermal analyses.
(1) Silicone rubber O-ring
()-Table B.7(4) shows the specifications of the silicone rubber O-ring.
()-Table B.7 Specifications of silicone rubber O-ring Item Specifications Material Silicone rubber Hardness Shore hardness: 70 Normal service temperature -47 to 150 Service temperature and period under accident 250, 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> conditions (2) Fusible plug
()-Table B.8 shows the specifications of the fusible plugs.
()-Table B.8 Specifications of fusible plug Item Specifications Material Solder (JIS Z 3282 H63A)
Melting point 183 to 184
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B.4 Normal test conditions The following sections will show how the package meets the requirements of the technical standards under normal test conditions.
B.4.1 Thermal analytical model Since the decay heat of fuel elements is minimal, the heat emitted by the contents is not taken into account in the analyses.
No heat is generated by the contents of the package and no solar radiation enters, in the shade with a 38 ambient temperature, the temperature on the outer surface does not exceed 38.
Increase in temperature of the package under normal test conditions is caused by entry of solar radiation heat with a 38 ambient temperature.
This analysis uses a vertically positioned package model.
Solar radiation heat enters it and is transmitted by natural convection and radiation.
In this analysis, simplified calculation methods are used (B.6.1, APPENDIX).
B.4.1.1 Analytical model This section will describe the following items related to the calculations.
Geometrical model Conditions for analyses Heat transfer in the package.
(1) Geometrical model The geometrical model for thermal analyses under normal test conditions supposes that no deformation occurs in the cylindrical packaging that is 840 mm in diameter and 1,800 mm in height.
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(2) Conditions for analyses
()-Table B.9 shows the thermal conditions under normal test conditions.
()-Table B.9 Thermal conditions under normal test conditions Item Conditions Decay heat (W) 0 0 0 Environmental Ambient Stagnant Stagnant Stagnant conditions temp.() air air air 38 38 -40 Solar rad. 400 ,
2 0 0 heat (W/m ) 800 Ambient rad.
1.0 1.0 1.0 factor Radiation factor for packaging 0.4 0.4 0.4 surface
- Although the radiant heat on the surface of an article that is vertically transported is 200 w/m2, 400 W/m2 shall be conservatively set as the value for other surfaces.
- "The surface of an article that is horizontally transported" and "the surface turned upward" (3) Heat transfer in the package (see ()-Fig.B.2)
With regard to heat transfer in the package, the following conditions apply, (a) Deformation is not taken into account since deformation under normal test conditions is minimal.
(b) Steady state thermal calculations are performed for the package surface of the model in which heat entry (solar radiation heat) and heat emission (natural convection to the atmosphere and radiation) are in equilibrium.
(c) The maximum temperature on the package surface (paragraph b) represents the maximum temperature in the package.
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(d) Only solar radiation heat enters the package. This heat is transferred to the outer surface of the package by natural convection and radiation.
(e) Heat reaching the outer surface of the package is transferred to the inner surface of the inner shell by thermal conduction.
Based on these conditions, steady state thermal calculations were performed by simplified methods.
The details of the results are given in B.6.l, APPENDIX.
B.4.1.2 Test model An analytical model is used, and a test model is not used.
B.4.2 Maximum temperatures
()-Table B.10 shows the maximum temperatures on the main parts of the package under normal test conditions.
()-Table B.10 Maximum temperatures of each part of package Item Normal test conditions No solar rad. Solar rad. No solar rad.
heat heat heat Ambient Ambient Ambient Parts temp.: 38 temp.: 38 temp.:-40 Ext. surface 38 65 -40 of basket Inner lid 38 65 -40 O-ring Inner surface 38 65 -40 of inner shell Outer surface of 38 65 -40 main body
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The maximum temperature of the package under normal test conditions is uniformly 62.1 at these parts as shown in section B.6.1 Appendix. The value of 65 is adopted here as a conservative figure.
B.4.3 Minimum temperatures Since the small amount of decay heat of the contents is not taken into account, temperatures at various parts of the package are uniformly -40 under the conditions of no solar radiation heat and -40 stagnant air. Under this thermal condition, the packaging maintains its capabilities, since the value -40 lies within the normal service temperature range of the silicone rubber O-ring (-47 to 150). The structural material is stainless steel and does not embattle.
Hence, the packaging maintains its integrity.
B.4.4 Maximum internal pressure As described in Section B.4.2, the maximum and minimum temperature of the package is 65 and -40 under normal test conditions. In the evaluation of the maximum inner pressure under normal test conditions, pressures due to thermal expansion of the air contained in the packaging are taken into account on the supposition that the temperature of each part of the package is uniformly 65, as shown in Section B.6.4., APPENDIX.
The inner pressure in the packaging is thus 0.016 MPaG in the temperature range of -40 to 65. Even when the temperature changes from -40 to 65, the inner pressure is 0.046MPa [gauge]. Since these values are far lower than the design pressure of 0.0981 MPaG, the package maintains its integrity.
B.4.5 Maximum thermal stress Thermal stresses under normal test conditions do not adversely affect the structural strength of the package as shown in Section A.5.1, Chapter ().
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B.4.6 Summary of the results and evaluation We confirmed that the structural strength and containment of the package are not adversely affected by the normal test conditions, as shown by the following evaluations of the thermal analyses. As already mentioned above, the evaluation was carried out with the minimum temperature of the components of the package being -40 and the maximum temperature of 65.
(1) Surface temperature of package The surface temperature of the package is 65, lower than the allowable reference value 85.
(2) Structural strength The various parts of the package were analyzed for their maximum inner pressure, thermal stress and maximum temperature, which constitute the main factors for structural strength. For the maximum internal pressure, the internal pressure rises by 0.016 MPaG in the packaging, far lower than the design pressure of 0.0981 MPaG and does not adversely affect the structural strength.
Thermal stresses do not adversely affect the structural strength of the packaging, as described in Section A.5.1, Chapter ().
(3) Containment The inner lid O-ring, functioning as containment border and thus constituting the most important part for containment, was evaluated for its temperature, deformation and maximum internal pressure.
The temperatures of the O-ring, containment border, are within the range from -40 to 65. Since this range is within its normal service temperature range (-47 to l50), the O-ring does not deteriorate.
No deformation occurs that might adversely affect the containment border.
When the external pressure drops to 60 kPa, the internal/external pressure difference is 0.056 MPa. On the other hand, since the internal pressure of the inner container is 0.0981 MPa, which is the design pressure, the structural soundness and sealing performance of the sealing device are ensured even when the external pressure drops to 60 kPa.
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B.5 Accident test conditions This section will describe how the package meets the technical standards under accident test conditions.
B.5.1 Thermal analytical model Thermal evaluations were conducted for accident test conditions, using the three-dimensional, non-steady state thermal analysis code TRUMP.
B.5.1.1 Analytical model The section concerns the following items used in the calculation by TRUMP.
Geometrical model Conditions for analyses Heat transfer in the package.
(1) Geometrical model As shown in Section A (Structural Analyses), Chapter (), the packaging maintains its integrity in spite of small local deformations in the drop tests under accident test conditions, as required for Type B(U) packages.
Since the drop test I showed the deformation imposed was 126.7mm in vertical direction, being 81.6mm in horizontal direction, the thermal analysis under the specific testing conditions adopted the dimensions of shock absorber and heat insulating material reduced up to 51mm (deformation;135mm) despite 186mm before the deformation imposed in the former in axial direction and up to 82mm (deformation:95mm) despite 177mm before the deformation imposed in the latter in radial direction respectively.
However, the drop test showed the deformation rather localized and it seemed there were no significant effects considered thermally, so that no particular modeling was considered.
()-Fig.B.3 shows the geometrical model (axially symmetrical, two-dimensional model) under the accident test conditions.
In this geometrical model, a circular section was adopted despite the actual angular section, as shown in B.6.3.
The following parts were evaluated.
Fuel basket Inner surface of the inner shell Inner lid O-ring Outer surface of the main body.
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Unit (mm)
()-Fig.B.3 Two dimensional axis symmetrical model
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(2) Conditions for analyses The following thermal conditions were used in the analyses.
The decay heat of the contents is minimal and is not considered. The thermal analyses for the accident test conditions suppose that the package is placed under fire accident conditions subsequently to the mechanical test conditions under accident test conditions. The temperature distribution for the normal test conditions is used for the packaging which has not undergone the fire conditions.
The thermal conditions during fire accident are, ambient temperature of 800 , period of 30 minutes, fire radiation factor of 0.9, and radiation factor for the package surface of 0.8. The package is supposed to suffer solar radiation heat. Both radiation and convection are taken into account with regard to the heat transfer from the ambient environment to the packaging.
The thermal conditions after fire accident are, ambient temperature of 38 , radiation factor for outer surface of the main body as packaging surface of 0.6, and radiation factor for ambient environment of 1.0.
Natural convection and radiation are taken into account with regard to the heat diffusion from the outer surface of the packaging. Solar radiation heat is also taken into account.
()-Table B.11 shows the above conditions for analyses.
The evaluation takes into account any entry of heat due to a fire resulting from combustion of the heat decomposition gas from hard polyurethane foam.
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()-Table B.11 Thermal conditions under accident test conditions Item Initial During After conditions fire fire accident accident Decay heat (W) 0 0 0 Environmental Ambient Stagnant 30 minutes Stagnant Conditions temp.() air air 38 800 38
() ()
Solar rad. 400 , 400 , 400(),
heat (W/m2) 800() 800() 800()
Ambient rad.
1.0 0.9 1.0 factor Radiation factor for packaging 0.4() 0.8() 0.6()
surface (a): Surface radiation factor for steel (SUS304) not exposed to fire.
(b): Surface radiation factor for steel (SUS304) being exposed to fire.
(c): Surface radiation factor for steel (SUS304) exposed to fire.
(d): Although the radiant heat on the surface of an article that is vertically transported is 200 w/m2, 400 W/m2 shall be conservatively set as the value for other surfaces.
(e): "The surface of an article that is horizontally transported" and "the surface turned upward" (3) Heat transfer for package (see ()-Fig.B.2)
For the heat transfer for the package, the evaluation supposes that, (a) External heat is transferred to the outer surface of the package through natural convection and radiation.
(b) The heat on the outer surface of the package is transferred to the inner surface of the inner shell through thermal conduction.
(c) The heat on the inner surface of the package is transferred to the outer surface of the fuel basket by radiation and thermal conduction.
(d) The interior of the basket maintains its heat insulating capability as under the normal test conditions.
The relational expressions used in the analyses of these heat transfers are shown in Section B.6.3, APPENDIX.
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(4) Thermal analyses of fissile packages As shown below, the deformation of the fissile package which suffers composite effect of different drops under normal test conditions plus a 9 m drop is smaller than that obtained for this thermal analytical model(()-Fig.B.3) except in the case of the vertical drop, in which deformation slightly exceeds 1.6 mm.
Vertical Item Horizontal Lid side Bottom side Minimum thickness before deformation 186 199 177 (mm)
Deformation at 9m 126.7 106.3 81.6 drop as BU package (59.3) (87.7) (95.4)
(mm)
Deformation at 9m 136.6 117.6 88.8 drop as Fissile package (49.4) (76.4) (88.2)
(combination) (mm)
Deformation of 135 95 thermal analytical model (51) (82)
(mm)
Numbers given in brackets( )indicate remaining thickness.
In addition, a combination of drops test I and causes no deformation in the inner shell, and deformations are local.
There supposed to be no significant difference between the thermal analytical model taking into account the composite effect of various conditions on fissile packages and the thermal analytical model, for this reason, the package is not analyzed here for thermal conditions under the accident test conditions.
B.5.1.2 Test model An analytical model is used, and a test model is not used.
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B.5.2 Evaluation conditions for packages In the evaluation, the conditions shown in ()-Fig.B.3, which take into account deformations resulting from drop tests under accident test conditions, were used.
B.5.3 Temperatures of packages
()-Fig.B.4 shows the results of the calculations using the analytical model described in Section B.5.1.1. Temperature evolutions for various parts of the package under accident test conditions are plotted here in relation to time. ()-Table B.12 shows the maximum temperature of each part and the period of time required from the occurrence of fire to the attainment of the maximum temperature.
()-Table B.12 Maximum temperatures of package under accident test conditions Item Accident test conditions Maximum temp. Time required from fire occurrence of fire to attainment Parts of maximum temp.
Fuel basket 209.9 Approx. 1.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Inner lid O-ring 187.8 Approx. 0.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> Inner surface of 483.2 Approx. 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> inner shell Outer surface of 1,226.6 Approx. 0.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> main body Note: The fact that the maximum temperature of the outer surface of the main body exceeds the ambient temperature of 800 is explained by the combustion of the gases generated from the heat insulator passing through fusible plugs.
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()-Fig.B.4 Temperature time history under accident test conditions
B.5.4 Maximum internal pressure The evaluation of the maximum internal pressure under accident test conditions takes into account the pressure due to thermal expansion of the air contained in the packaging. The calculation methods shown in Section B.6.4, APPENDIX, were used.
The value 0.065 MPaG was obtained for the internal pressure in the packaging. Since this value is lower than the design value, the packaging maintains its integrity at its different parts.
B.5.5 Maximum thermal stresses Thermal stresses occurring in the package under accident test conditions do not adversely affect its structural strength, as shown in Section A.6.3, Chapter ().
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B.5.6 Summary of results and evaluation We confirmed that the structural strength and containment of the package are not adversely affected by the accident test conditions, as shown by the following evaluations of the thermal analyses.
(1) Temperatures
()-Table B.12 shows the maximum temperatures of various parts of the package under accident test conditions, and ()-Fig.B.4 shows the recorded temperatures of various parts under accident test conditions.
The fuel basket under accident test conditions reaches its maximum temperature of 209.9 1.6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the occurrence of fire. Since this evaluation supposes maintenance of heat insulation in the basket, the temperature of plate-shaped fuel elements to be actually contained does not exceed 209.9.
This value is lower than the temperature of occurrence of blistering (allowable temperature for fuel) of 400 for plate-shaped fuel elements used in the experiment and research reactors of the Japan Atomic Energy Research Institute and Institute for Integrated Radiation and Nuclear Science, Kyoto University. Therefore, the contents maintain their soundness.
The inner lid O-ring reaches its maximum temperature of l87.8, 0.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> after the occurrence of fire. This value is lower than the service temperature 25O for the silicone rubber O-rings under accident. Thus, the O-ring maintains its integrity even under the accident test conditions, and the packaging retains its containment.
(2) Pressure As described in the preceding section, the temperature of the parts of the package rises under the accident test conditions. This rise in temperature causes the air in the packaging to thermally expand, raising the internal pressure.
The packaging is evaluated for its internal pressure, supposing the maximum temperature of the outer surface of the basket to be 209.9. ()-Table B.13
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shows the maximum pressure in the packaging under accident test conditions.
()-Table B.13 Maximum pressure in packaging under accident test conditions Conditions Maximum pressure under accident Position test conditions (MPa[gauge])
Inside the packaging 0.065 The pressure 0.065 MPa[gauge] is lower than the design pressure for the packaging, 0.0981 MPa[gauge].
Thus, the packaging maintains its integrity.
(3) Structural strength This section concerns the maximum inner pressure, thermal stresses and maximum temperature in the packaging, which are to be examined for structural strength of the packaging.
As far as the maximum inner pressure is concerned, the pressure rise in the packaging 0.065 MPa[gauge] is lower than the design pressure 0.0981 MPa
[gauge] and does not adversely affect the structural strength of the packaging.
As shown in Section A.5, Chapter (), thermal stresses do not adversely affect the structural strength of the packaging.
(4) Containment The maximum temperature of the O-ring provided on the inner lid, which constitutes the containment border, is l87.8. This value is lower than the service temperature 250 of silicone rubber O-ring under accident conditions, and the package thus maintains its containment.
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B.6 Appendix B.6.1 Maximum temperature of package under normal test conditions
- ()-B-26 B.6.2 Outline of TRUMP -- General purpose program for heat transfer
- ()-B-29 B.6.3 Input data for TRUMP used for temperature calculations for accident test conditions ********************************* ()-B-35 B.6.4 Internal pressure of the package ************************* ()-B-40 B.6.5 Validity Justification of thermal analysis methods ******* ()-B-41 B.6.6 Bibliography ********************************************* ()-B-44 B.6.1 Maximum temperature of package under normal test conditions The maximum temperature is obtained, using the thermal balance for a steady state, as follows.
The quantity of entering heat Q in [kcal/h] only consists of solar radiation heat, and the quantity of emitted heat Q out [kcal/h] is the sum of radiation heat Q 1
[kcal/h] and emitted heat due to natural convection Q 2 [kcal/h]. The packaging reaches its maximum temperature, when Q in = Q out .
It is obtained with the outer surface temperature of the packaging t [],
supposing, t o : Ambient temperature, t o = 38 []
A v : Vertical area to which heat is transferred, A v =x0.84x1.8 = 4.750 [m2]
A h : Upper horizontal area to which heat is transferred, A h = 0.842x/4 [m2] = 0.554 [m2]
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(1) Radiation heat from solar heat Q in Q in = 400 [W/m2]xxA v + 800 [W/m2]xxA h
= 344 [kcal/m2h] kxxA v + 688 [kcal/m2h] xxA h ...(6.1-1) where : radiation factor for the packaging outer surface,
= 0.4 (2) Radiation heat from package Q 1 Q 1 = (A v + A h )xxx(T4 - T o 4)
= (4.75 + 0.554)x0.4x4.88x10-8x{(t + 273)4 - (38 + 273)4 }
= 10.353x10-8x{(t + 273)4 - 3114 } .....................(6.1-2)
T = t + 273 where T: Absolute temperature [K]
t: Outer surface temperature for the package []
- Stefan -Boltzmann constant [kcal/m2 hk4]
(3) Emitted heat due to natural convection Q 2 Heat transfer of natural convection of vertical cylindrical surface is given (5) by Mc Adams formula as follows.
Nuv= 0.13 (Gr/Pr)1/3(5) [109<Gr/Pr<1012] .....(6.1-3)
Nuv: Nusselt number, Nuv = h v L/K .....(6.1-4)
Gr: Grashof number, Gr = gL3t/2 ....(6.1-5)
Pr: Prandtl number, Pr = C p /K .....(6.1-6) where h v : Heat transfer coefficient for vertical, cylindrical surface [kcal/m2h]
L : Representative length [m]
K : Heat transfer coefficient for air [kcal/mh]
g : Gravitational acceleration, 9.8 [m/sec2] = 1.27x108 [m/h2]
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- Coefficient of cubical expansion for air [l/K]
t: Difference of temperatures (t-t o ) []
- Coefficient of kinematic viscosity for air [m2/h]
C P : Isopiestic specific heat for air [kcal/kg]
- Viscosity of air [kg/mh]
The Nusselt number Nu is obtained by Equations (6.1-3), (6.1-5), and (6.1-6),
and the heat transfer coefficient for vertical, cylindrical surface h v by Equation (6.1-4). The heat transfer coefficient for horizontal surface h h is similarly obtained by Equations (6.1-7) and (6.1-8).
Nun= 0.14(GrPr)1/3 [2x107< GrPr<3x1010] ........(6.1-7)
Nun= h h L/k ......................................(6.1-8)
The emitted Heat due to Natural Convection Q 2 is Q 2 = (h v A v + h v A v )(t - t o ) .....................(6.1-9)
(4) Calculation of the maximum temperature t max When the air temperature is 38, each value is, L = 1.8 [m]
g = 1.27x108 [m/h2]
k = 0.0271 [kcal/mh]
= 1/(273+38) = 3.22x10-3 [l/K]
t= t max - t o []
= 0.0623 [m2/h]
a = 0.0882 [m2/h]
Hence, by equations (6.1-6), (6.1-5), and (6.1-3),
Pr = /a = 0.0623/0.0882 = 0.706 Gr = gL3t/2
= 1.27x108x3.22x10-3x1.83xt/0.06232
= 6.14x108xt
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Nu = 0.13(GrPr)1/3 = 98.4xt1/3 Thus, using equation (6.1-4),
Nuv k hv = = 1.481xt1/3 [W/m2K]
L and Nuh k hh = = 1.595xt1/3 [W/m2K]
L The thermal balance in the steady state is Q in = Q out . The convergence calculation for the difference of temperatures t, using Equations (6.1-1),
(6,1-2), (6.1-9) and the heat transfer coefficient h, leads to the maximum temperature t max .
t max = 62 []
The value of 65 is adopted here as a conservative figure.
B.6.2 Outline of TRUMP --General purpose program for heat transfer (1) General TRUMP is a program developed in 1968 by the Lawrence Livermore Laboratory for heat transfer calculations based on a node method.
(2) Functions The program TRUMP is designed to handle heat generation, chemical reactions, phase changes, and heat transfer. This program can cover 3-dimensional objects by dividing them into elements by means of rectangular, cylindrical, rotating body or polar coordinates.
Material properties such as heat transfer coefficient and specific heat are given as functions of temperature or time.
The program can handle heat transfer between elements resulting from thermal conduction, natural convection, forced convection, and radiation as well as that resulting from natural or forced convection and radiation as boundary condition. In this program, boundary temperatures can be expressed as functions of time. Initial temperature can vary with position in the space.
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TRUMP outputs can be obtained, such as temperature distribution for determined time points and thermal balance for each element.
(3) Calculation methods (see ()-Fig.B.5)
The TRUMP solves simultaneous partial differential equations that have four independent variables regarding space coordinates and time as well as a total of three dependent variables, temperature and two densities of reactant.
In case of normal three dimensions, the equations for heat generation, thermal conduction accompanied by chemical reactions and mass transfer are given in the form of normal vector operations:
DT T
= +vT Dt t 1 Qa a Qb b
= KT+G-C C t C t Da a
= +va Dt t E
=-aexp Z a a R T Db b
= +vb Dt t E
=-bexp Z b b R T T T K 1 1 i = h i (T 2i - T 1i ) = K 2 2 i r r h i = h io + h ic [(T 2i - T 1i )2]Pi/2 + F i (T 1i + T 2i )(T2 1i - T2 2i )
The conductance h i for the boundary surface is expressed in a common form that takes into account contact conductance, natural convection, forced convection, and radiation. is the Stefan-Boltzmann constant and F is the overall radiation morphological coefficient.
T K s = U sb (T b - T s )
r where T b : external temperature
()30
U sb : surface conductance.
As in the case of mass phase, U sb is U sb = h so + h sc [(T b - T s )2]Pi/2
+F b (T s + T b )(T s 2 + T b 2)
The TRUMP solves actual equations in relation to minute periods of time. In fact, the time differential u/t should be replaced by (u' - u)/t in the preceding equation. U' and U are the initial and final value of the period of time t.
(4) Utilization of TRUMP The TRUMP program, developed by the Lawrence Livermore Laboratory, has been and is being used in many laboratories in the United States.
()31
Start Data input Physical properties Description of node shape (volume)
Description of internal/external heat contact Description of initial conditions/boundary conditions(external temperature)
Description of attributes of chemical reactions Description of internal heat (arbitrary) generation Description of mass flow Initial setting Entered data printout 3
Selection of period of time t Calculation of node attributes Data such as node such as heat transfer coefficient, temperatures are mass, thermal capacity, quantity of heat, printed at specified latent heat, mean temperature, etc. times.
Internal Yes heat generation Generated heat is calculated, No heat flow due to heat generation emitted from/applied to nodes is calculated.
Yes Chemical reactions Chemical reaction attributes such No as chemical reaction heat are calculated, heat flow due to chemical reactions emitted from/
applied to nodes is calculated.
1
()-Fig.B.5 TRUMPflowchart (1/3)
()32
1 Yes Internal heat contact No Contact conductance for heat radiation, contact conductance between nodes with different heat transfer coefficient, and heat flow through nodes in mutual contact are calculated.
Yes Mass flow No Mass flow rate, quantity of heat transferred by mass flow, latent heat for diffusion or absorption, and enthalpy are calculated, density evolution and density are also calculated.
Yes External contact No Quantity of distributed heat flow caused by thermal contact, external temperature, heat transfer coefficient, and quantity of heat flow caused by thermal contact are calculated.
Yes Phase change No Latent heat for diffusion or absorption in nodes with changing phase 2
()-Fig.B.5 TRUMPflowchart (2/3)
()33
2 Yes Special node No In case special nodes are contained in external thermal contact, internal thermal contact, contact through mass flow, each quantity of heat flow is calculated, in case of contact between special nodes, calculation is performed repeatedly under the convergence condition for temperature changes.
Yes Phase change No Quantity of phase change is calculated.
No 3 Task end Yes End
()-Fig.B.5 TRUMPflowchart (3/3)
()34
B.6.3 Input data for TRUMP used for temperature calculations for accident test conditions (1) Modeling of fuel basket (see ()-Fig.B.6)
The fuel basket was modeled to a cylindrical shape the wall thickness of which is equal to the smallest gap between the inner shell and the fuel basket.
The heat capacity of the fuel basket was corrected to be equivalent by compensating the specific weight.
The inner diameter of the Main body of inner shell
()-Fig.B.6 Fuel basket model (a) The outside radius R 1 of the cylindrical model R 1 is, R1 = R - G where R: Inside radius of inner shell, R = 230 [mm]
G: Gap (minimum), G = 230 - 150x 2
= 17.87 [mm]
R 1 = 230 - 17.87
= 212.13 [mm]
()35
(2) Heat transfer between package outer surface and ambient environment (a) Convection heat transfer coefficient The heat transfer coefficient for natural convection on the outer surface of the package is obtained by McAdams equation5),
(i) Outer surface of vertical cylinder g L3 t Gr = ....(6.3-1) 2 Nuv= 0.13(GrPr)1/3 [109<GrPr<1012] ....(6.3-2)
Nuv k h = ....(6.3-3)
L (ii) Upper horizontal, flat surface g L3 t Gr =
2 Nuh= 0.14(GrPr)1/3 [2x107<GrPr<3x1010] ....(6.3-4)
Nuh k h =
L Where h : Convection heat transfer coefficient [cal/cm2 s]
k : Heat transfer coefficient of air [cal/cm s]
During fire: 7.094x10-4 (at 800)
After fire: 2.706x10-5 (at 38)
L : Representative length [cm]
Vertical surface: 152 [cm]
Horizontal surface: 65 [cm]
g : Gravitational acceleration; g = 980 [cm/s2]
- Coefficient of cubical expansion [1/K]
During fire: 1/(273+800) = 9.23x10-4 After fire: 1/(273+38) = 3.22x10-3
- Coefficient of kinematic viscosity [cm2/s]
During fire: 1.37
()36
After fire: 0.173 Gr: Grashof number Pr: Prandtl number; Pr = 0.706 Nu: Nusselt number
()-Table B.14 shows the results of a calculation in which the preceding values were substituted for the corresponding letters of Equations (6.3-1) to (6.3-4)
()-Table B.14 Convection heat transfer coefficient between package surface and ambient environment Position Vertical cylindrical Upper horizontal Condition surface surface During fire 6.459x10-5t13 6.956x10-5t13 After fire 1.480x10-4t13 1.594x10-4t13 (b) Radiation heat transfer The radiation morphological coefficient is 1
F 12 = .....(6.3-5) 1 / 1 + 1 / 2 1 Where F 12 : Radiation morphological coefficient 1 : Radiation factor for surface No. 1 2 : Radiation factor for surface No. 2
()-Table B.15 shows the radiation factors for both surfaces and the radiation morphological coefficient obtained using Equation (6.3-5).
()-Table B.15 Radiation factor and radiation morphological coefficient Condition During fire After fire Item Radiation Package Surface 0.8 0.6 factor Ambient environment 0.9 0.1 Radiation morphological coefficient 0.735 0.6
()37
(3) Heat transfer between basket and inner cylinder (a) Convection heat transfer coefficient The heat transfer coefficient for the closed fluid layer between vertical, concentrical cylinders2) is obtained by means of the following equations.
Nu = 1.0 [Ra<103] ...........(6.3-6)
Nu = 0.28Ra1/4(L/D)1/4 [103<Ra<107] ........(6.3-7) g 3 t Ra = .................(6.3-8) a Nu k h = .................(6.3-9)
D where Nu : Nusselt number Ra : Raleigh number g : Gravitational acceleration, g = 980 [cm/s2]
- Coefficient of cubical expansion,
= 1/(273+250) = 1.912x10-3 [1/K]
D : Thickness of fluid layer, D = 23 - 21.213 = 1.787 [cm]
t: Temperature difference between inner and outer cylinder []
a : Thermal diffusivity, a = 6.194x10-1 [cm2/s]
- Coefficient of kinematic viscosity,
= 0.426 [cm2/s]
L : Length of fuel basket [cm]
k : Heat transfer coefficient, k = 4.175x10-4 [cal/cm s]
h : Heat transfer coefficient for natural convection,
[cal/cm s]
980 x 1.912 x 10 3 x 1.787 3 Ra = t 6.194 x 10 1 x 0.426
= 40.52xt
()38
When 200 [] is substituted for t the Reynolds number Ra is, Ra = 40.52x200 = 8.104x103 [103<Ra<107]
Nu is obtained by Equation (6.3-7)
Nu = 0.28Ra1/4(L/D)-1/4
= 0.28x(8.104x103)1/4x(125.6/1.787)-1/4
0.918 Thus, using Equation (6.3-9), the heat transfer coefficient for natural convection h is Nu K 0.918 x 4.175 x 10 h =
D 1.787
= 2.145x10-4 [cal/cm2s]
(b) Radiation heat transfer The radiation morphological coefficient for the gas layer between concentrical cylinders is obtained using the following equation:
1 F 12 = ..........(6.3-10) 1 / 1 + (A 1 / A 2 )(1 / 2 1)
A1 / A2 = r1 / r2 ..............(6.3-11) where F 12 : Radiation morphological coefficient 1 : Radiation factor for surface No.1; 1 = 0.4 2 : Radiation factor for surface No.2; 2 = 0.4 r1 : Inside radius of external cylinder; r 1 = 230 [mm]
r2 : Outside radius of internal cylinder; r 2 = 212.13 [mm]
1 F 12 =
1 / 0.4 + (230 / 212.13)(1 / 0.4 1)
= 0.242 (4) Entry of Heat due to Fire Resulting from Combustion of Hard Polyurethane Foam The package was analyzed on the assumption that the heat resulting from combustion of hard polyurethane foam (23.45 kJ/g) exists on the outer surface of the package in the form of fire.
()39
B.6.4 Internal pressure of the package The internal pressures of the package under normal and accident test conditions are calculated.
(1) Operating pressures The operating pressure of the air in the packaging is obtained.
(a) Initial pressure The initial pressure in the packaging is equal to the atmospheric pressure
(=0.101 MPa abs).
(b) P 1 The pressure resulting from air expansion P 1 is obtained using the following equation based on the Boyle-Charles law.
T1 P1 = P0 .......(6.5-1)
T0 where P 0 : initial pressure (at 20);
P 0 = 0.101 [MPa]
T 0 : Initial temperature; T 0 = 273 + 20 = 293 [K]
T 1 : Air temperature under specific conditions [K]
()-Table B.16 shows the results of this calculation.
()-Table B.16 Calculation result for packaging internal pressure Position Air in the packaging Test conditions Normal Accident Pressure (MPa 0.016 0.065
[gauge])
Temperature() 65 209.9 When the lid is virtually tightened at -40 and the temperature rises to a maximum temperature of 65, the internal pressure is 0.046 MPa [gause] based on the formula (6.5-1) as T 0 : 233 [K] and T 1 : 338 [K], and it is less than the design pressure of 0.0981 MPa.
()40
(2) Design pressures The conservative design pressures shown in ()-Table B.17 are used for the various parts of the package evaluation.
()-Table B.17 Design pressures for specific test conditions Inside the packaging Normal test conditions 0.0981 MPa[gauge]
Accident test conditions 0.0981 MPa[gauge]
B.6.5 Validity Justification of thermal analysis methods This section describes the examination of the analyses simulating the fire test (herein referred to as analyses ) on the basis of the results of the fire test on a prototype packaging, analyses carried out to verify the justifiability of the thermal analysis methods described in this section.
(1) Prototype Packaging and Test Methods Prototype packaging: see Reference [7].
Test methods: see Reference [7].
()41
(2) Examination of analysis results
()-Table B.18 and ()-Fig.B.7 show the test and analysis results.
Analyses were performed using the conditions described in Section B.6.3.
The measurements recorded in the tests of prototype packaging were used as input data for the initial temperature and the temperatures in furnace in order to simulate actual test conditions. This indoor test does not take into account solar radiation heat.
As shown in ()-Table B.18 and ()-Fig.B.7, the analytical values are conservative and the thermal analysis methods shown in Sections B.5 and B.6.3 are valid.
()-Table B.18 Comparison of prototype packaging test results with analysis results Conditions Maximum Time required before temperature the maximum temperature
() hour(h)
Evaluation position Test Analysis Test Analysis Near O-ring 88.6 161.0 2.0 1.0 Inner surface 396.2 464.1 0.6 0.5 of inner shell Outer surface 123.3 182.5 1.0 1.6 of fuel basket Outer surface of 1051.6 1229.7 0.1 0.1 packaging
()42
()43
()-Fig.B.7 Comparison of prototype packaging test results with analysis results
B.6.6 Bibliography (1) Study on an application of inelastic structures analyzing methods (), a report at Section Meeting for Application of Inelastic Structures Analyzing Methods (EPIOC), Mechanical Engineering society of Japan, 1977.
(2) Material for Heat Transfer Engineering, Edition, Mechanical Engineering Society of Japan, 1975.
(3) In-house data of Nihon Asbestos Co., Ltd.
(4) In-house data of Nippon Valqua Industries, Ltd.
(5) McAdams, Heat transmission.
(6) In-house data of Mitsubishi Heavy Industries, Ltd.
(7) Prototype container test report for JRF-90Y-950K type transport container, Japan Atomic Energy Agency, May 1990.
()44
()-C Containment analysis
()-C. Containment analysis C.1 General The following part relates to the sealing performances of this packaging tested under normal and accident test conditions. The containment system is considered as the part which ensures the sealing of the packaging. The containment system of this packaging consists of an inner shell comprising a main body and a lid, and the contact between the main body and the lid is sealed by a silicon rubber O-ring (inner shell lid O-ring).
The leakage rate of the containment system is checked by leak tightness test and must meet the reference value during the manufacturing process and the maintenance period. The leakage rate of the O-ring of the inner shell lid is checked by a leak tightness test carried out before shipment of the package and must be confirmed meet the reference value.
C.2 Containment system C.2.1 Containment system (1) Structure The containment system of this packaging is composed, as shown in ()-Fig.C.1, of an inner shell main body and an inner shell lid.
(2) Materials The material used for the fabrication of the main body and the lid of the inner shell is stainless steel, and the sealing part of the inner shell lid is a silicon rubber O-ring.
(3) Design pressure and design temperature As shown in the ()-Table C.1, the leakage rate is evaluated according to the design pressure and design temperature.
1
()-Table C.1 Design pressure and design temperature of containment system Conditions Item Containment System Normal test Design pressure (MPa[gauge]) 0.0981 conditions Design temperature () 65 Accident test Design pressure (MPa[gauge]) 0.0981 conditions Design temperature () 209.9 2
Leak test orifice O-ring made by silicon rubber (Inner shell lid O-ring)
Inner shell lid Weld Inner shell main body Weld The range that the surrounded with a slanted line shows a seal border.
()-Fig.C.1 Containment boundary of packaging 3
(4) Seal Since the inner shell lid is covered by the outer shell lid, there is no possibility of the clamping bolts being removed inadvertently.
Moreover, after installation of the clamping bolts to fix the lid to the main body of the inner shell, the lid is sealed and locked.
(5) Manufacture and checking Manufacture and checking of the structural parts of the containment system are conducted by a suitable method which ensures sealing performances.
C.2.2 Penetration of the containment system Since the on1y opening of this packaging is the inner shell lid, this item is not applicable.
C.2.3 Gasket and weldings of the containment system (1) Containment system gasket For a gasket of the containment system gasket a silicon rubber O-ring is used.
With this O-ring no chemical or electrical reaction should occur, as explained in ()-A.4.l. Moreover, this ring shows excellent sealing performances under the pressures and temperatures in normal and accident test conditions.
(2) Specifications of the gasket (C-4) (C-3)
The dimensions and material of the gasket are shown in (II)-Table C.2 The silicon rubber O-ring can maintain the sealing performance of the inner shell lid under the normal and special test conditions and at the lowest temperature of use, with its heat-resistant property(See B.3 Specifications of components) and cold-resistant property(See A.4.2 Low temperature strength).
4
(II)-Table C.2 The dimensions and material of the gasket Positions dimensions material Note Inner shell Inner side 5.4xI.D. 473 silicone O-ring lid Outer side 5.4xI.D. 513 rubber (3) Weldings The weldings of the flange, of the barrel, and of the bottom plate are performed as explained in Chapter ()-A. Weldings are subjected to a non-destructive test during the fabrication process, as explained in Chapter
()-B, the integrity of the weldings is checked and a pressure resistance test is carried out to check the absence of leakage.
C.2.4 Lid The inner shell lid is equipped with 2 drains for the 2 silicon rubber O-rings, as shown in the ()-Fig.C.1. Moreover, the inner shell lid has been designed to be resistant under normal and accident test conditions and to maintain its performances. To preserve the sealing performances of the packaging, the inner shell clamping bolts are tightened to an appropriate torque as shown in ()-Table C.3.
Even when the external pressure drops to 60 [kPa], the opening amount of the inner container lid is less than the initial tightening allowance of the O-ring as shown in A.5.1, and the sealing performance is maintained.
()-Table C.3 Inner shell clamping bolt Designation Size Number Tightening torque (Nm)
Inner shell clamping bolts M24 16 Approx. 280 5
C.3 Normal test conditions The integrity of the containment system of this package remains unchanged after an impact under normal test conditions as required for all type B(U) packages, and as shown in the results of structural analyses in () - A.
Moreover, the results of thermal analyses in () - B show that variations in pressure or temperature under normal test conditions do not affect the integrity of the containment system.
Therefore, as the sealing performances of the containment system remain unchanged under normal test conditions, in these analyses, the evaluation of the sealing performances based on the leak tightness test of the O-ring of the inner shell lid which must meet the reference value, conducted before shipment of the package, shows that the leakage rate of radioactive substances under normal test conditions is lower than the IAEA regulation standard value.
C.3.1 Leakage of radioactive materials C.3.1.1 Volume of Leakage from the Inner Shell The containment system is checked against leakage by a leak tightness test carried out during the manufacturing process and the maintenance period.
For sealing performance, it is confirmed, further, on each shipment that the leakage rate of the package is lower than the reference value.
The leakage rate of radioactive materials is analyzed, on the assumption that, regarding the air supplied to the seal of inner shell lid on a leak tightness test, the pressure change corresponding to the maximum permissible leakage rate is detected in a certain time.
Radioactive materials exist in the gas of containment system and its leakage rate is different from that obtained from the air leakage rate.
Therefore, leakage rate of the gas under normal test conditions is first determined from the maximum permissible air leakage rate and then the obtained 6
leakage rate is applied to acquire leakage rate of the radioactive materials from concentration of radioactive materials in the gas. It is finally confirmed that the leakage rate of the radioactive materials is below the reference values specified by the regulation and notification.
(1) Maximum permissible leakage rate of the air The maximum permissible leakage rate of the air L a specified in design criteria for containment analysis is given as the leakage rate of the air in (II)-Table C.4 (II)-Table C.4 Maximum permissible leakage rate of the air Item Containment boundary (inner shell lid O-ring)
L a: maximum permissible leakage rate of the air(std cm3/s) 1.08x10-1 (2) Leakage rate at leak tightness test and the test conditions (a)Leakage rate at leak tightness test Leakage rate at leak tightness test by pressure drop is given by the following formula.
1)
VTS P1 P2 LR = (C.3-1) 60 HPS T1 T2 where, L R :Leakage rate(std cm3/s) under normal condition at 25, 0.101MPa(1 atm abs.)
V: Volume of the testing system(100m3)
H: Testing time T S :Reference temperature 298(K)
T 1 :Air temperature at the beginning of the test(K)
T 2 : Air temperature at the end of the test(K)
P S :eference pressure(0.101MPa, (1 atm abs.))
P 1 :Air pressure at the beginning of the test(MPa) 7
P 2 :Air pressure at the end of the test(MPa)
The formula (C-31) above will determine the air leakage rate with the following leak tightness test conditions, and the deduced value will be confirmed to be lower than the maximum permissible leakage rate, the reference value.
(b)Leak tightness test conditions
()Air pressure at the beginning of the test is fixed at 0.493(MPa)
()Air pressure at the end of the test is fixed at 0.297(MPa)
()Testing time is fixed at 30 min.
(iv)In calculation the temperature is set T 1= T 2= T S:= 298(K) (25).
These conditions are applied to the formula(C-31) to obtain the maximum permissible air leakage rate. The results of calculation is given in (II)-Table C.4.
(v)In consideration of the conditions (i) to (iv) above and the volume of the testing system, the testing time H and pressure dropP(P 1 -P 2 ) is fixed to confirm that the air leakage rate L R (L R =L R i) at O-ring of inner shell lid is lower than the maximum permissible air leakage rate L R( L R= 2.21x 10-2cm3/s at 0.493 MPa (1.08 std cm3/s), 298K).
(3) The maximum gas leakage rate under normal test conditions The maximum gas leakage rate under normal test conditions is obtained on the basis of the maximum permissible air leakage rate L R t as follows.
(a)Diameter of leak The leak is assumed to be a round hole which crosses the sealing part along the shortest path. Fluid is considered to pass through the leak in the form of free molecular flow or continuous flow and its leakage rate is given by the following formula.
L=(F c +F a )(P u -P d ) 2)(C.3-2) where, L: Volume leakage rate at pressure P a (cm3/s at P a , T a )
8
P a : Average pressure of flow (M P a )
Pa =
1 (Pu + Pd ) (C.3-3) 2 T a : Average temperature of fluid Pu: Pressure on upstream side Pd: Pressure on downstream side Fc: Flow heat conduction coefficient for continuous flow(cm3/MPas)
Fm: Flow heat conduction coefficient for free molecular flow(cm3/MPa. s) 2 D4 FC = 2.49 x 10 x (C.3-4) aµ T
D3 Fm = 3.81 x 10 3 x M (C.3-5) aPa Where, D: Diameter of leak (cm) a: length of leak (cm)
- Viscosity coefficient of the air(MPa.s)
T: Temperature of fluid(K)
M: Molecular weight(g/mol)
Diameter of leak hole is obtained by the following formula and the formula (C-3-2)
Ps Ta L = L R i (C.3-6)
Pa Ts Where, L R i: Air leakage rate at containment boundary(std cm3/s)
Ta: Average temperature(=TS)(K)
The maximum diameter of leak of inner shell lid on leakage rate test is given in (II)-Table C.5.
Note: the formula ANSI 4.5 is converted into SI unit.
9
(II)-Table C.5 The maximum radius of leak hole on leakage rate test Positions O-ring parts Items L R i: Air leakage rate at containment 1.08x10-1 boundary(std cm3/s)
P u : Pressure on upstream side(M P ) 0.493 a
P d : Pressure on downstream side(M P ) 0.101 a
P a : Average pressure of flow (M P a ) 0.297 T a . T: Temperature of the air(K) 298 L: Air leakage rate on leak tightness 3.673x10-2 test(cm3/s at P a , T a )
- Viscosity coefficient of the 1.85x10-11 at 25*1 air(MPa.s) a: length of leak hole(cm) 0.54 (note)
M: Molecular weight(g/mol) 29.0 F c : Thermal conductivity coeffim3/MPa 2.49x109D4 s)
F m : Thermal conductivity coefficient for 7.61x104D3 free molecular flow D: Diameter of leak (cm) 2.490x10-3 Note: Diameter of cross section of O-ring is employed.
1 : Since viscosity coefficient of the air increases with temperature, it is conservative to employ the low temperature.
10
(b)The maximum gas leakage rate under normal test conditions The maximum gas leakage rate under normal test conditions is obtained by substituting the values of pressure, gas and the maximum diameter of leak under normal test conditions into the formula (C.3.2) to (C.3.5).
Gas leakage rate L x calculated from(C.3.2) is converted to leakage rate Ls, x under normal test conditions, at 25, 0.101MP abs(1 atm abs), by the following formula.
Pa , x 298 Ls , x = L x x x (C.3-7) 0.101 Ta , x where, Subscript x: Normal test conditions but it is assumed that T a,x =T u,x Gas leakage rate under normal test conditions is provided in (II)-Table C. 6. The maximum gas leakage rate at O-ring is employed for calculation.
(II)-Table C.6 The maximum gas leakage rate under normal test conditions Position Containment boundary Item (O-ring of inner shell lid)
D: Diameter of leak (cm) 2.490x10-3 a: Length of leak (cm) 0.54
- Viscosity coefficient of the 1.85x10-11at 25*1 gas(MPa.s)
P u,x : Pressure of containment system 0.199 under normal test conditions(MPa abs)
P d, x : External pressure under normal 0.060 test conditions(MPa abs)
T u,x : Gas temperature under normal test 338 conditions(K)
M: Molecular weight(g/mol) 29 L x : Leakage rate under normal test 1.38x10-2 conditions(cm3/s at P a,x T a,x )
L s,x : Leakage rate under normal test 1.55x10-2 conditions(cm3/s at 25 0.10MPa) (5.58x101cm3/h) 1 : Since viscosity coefficient of the air increases with temperature, it is conservative to employ the low temperature 11
C.3.1.2 Evaluation of the volume of leakage radioactive substances (1) In transporting the fresh fuel elements and critical assembly fuel (a) Evaluation of the radioactive substances contained in the inner shell concerning the leakage.
Since there is no possibility of degradation of the fuel plates under normal test conditions, as described in part ()-A, it is considered that there is no leakage of the enriched uranium contained in the fuel plates. It is supposed that the only radioactive substances that may have leaked are the uranium particles which adhere to the surface of the fuel elements during the manufacturing process, in other words, uranium surface contamination.
It is supposed that the level of contamination is 8.00x10-2 Bq/100cm2[235U](1 g235U/100cm2) for the whole surface of the fuel elements, which is the reference value of the surface contamination test during manufacturing process.
It is supposed that the contaminated surface uranium are 93% enriched uranium, 45% enriched uranium, 20% enriched uranium and 93% enriched uranium 234 of the degraded uranium for which the rate of U/235U is at its maximum level.
The weight of radioactive nuclides of 93% enriched uranium adhering to one fuel element is calculated according to the usual method as follows.
235
() Quantity of U: This quantity is calculated by using the level of 8.00 x10-2 Bq/100cm2 and of the whole surface of the fuel element (l 235U/100cm2).
238 234 236
() Quantity of U: The quantity of U and U being considered as nil, 238 the quantity of U is calculated by using the lower limit of the 235 tolerance of the enrichment (93.15 +/- 0.15 wt%) of U calculated in
().
234 236 234 236
() Quantity of U and U: No weight limit has been fixed for U and U, because these are decided during the fuel manufacturing.
12
The quantity of 234U and 236U is calculated using the maximum weight proportion recorded in the past material record and in rounding off these figures (x
- 2) according to the usual method.
Moreover, the total weight of uranium needed for these calculations is 235 obtained by adding the quantity of U calculated in () and the quantity 238 234 236 of U calculated in (). The weight proportions of U and U used for the calculations are shown in the ()-Table C.7.
The Surface contamination level is shown in ()-Table C.8.
234 236
()-Table C.7 Weight proportions of U and U used for calculations Maximum weight Weight proportions Enrichment Isotope proportions in the used for calculations (wt%)
mill sheets (wt%) (wt%)
234 U 1.08 2.2 93.15+/-0.15 236 U 0.47 0.94 13
()-Table C.8 Surface contamination level per fuel element Radioactivity (Bq)
Fuel element 234 235 236 238 U U U U Total JRR-3 Standard Type (Uranium silicon 1.58x103 2.31x101 7.00 2.71x10-1 1.61x103 aluminum dis-persion type alloy)
JRR-3 Follower Type (Uranium silicon 8.80x102 1.29x101 3.90 1.51x10-1 8.97x102 aluminum dis-persion type alloy)
JRR 4B Type 9.34x102 1.37x101 4.14 1.60x10-1 9.52x102 JRR 4L Type (Uranium aluminum 9.34x102 1.37x101 4.14 1.60x10-1 9.52x102 dispersion type alloy)
JRR-4 (Uranium silicon 9.34x102 1.37x101 4.14 1.60x10-1 9.52x102 aluminum dis-persion type alloy)
JMTR Standard 1.22x103 1.78x101 5.41 2.09x10-1 1.24x103 JMTR 8.74x102 1.28x101 3.88 1.50x10-1 8.91x102 Follower Fuel KUR Standard & Half 1.12x103 1.64x101 4.97 1.93x10-1 1.14x103 Loaded KUR Special 8.01x102 1.17x101 3.56 1.38x10-1 8.17x102 KUCA Coupon (120 coupons as one 3.69x102 5.40 1.64 6.35x10-2 3.76x102 fuel element)
KUCA Flat (30 plates as one 1.25x102 1.83x101 5.56 2.16x10-1 1.28x103 fuel element) 14
(b) Evaluation of the leakage volume of radioactive substances under normal test conditions.
The uranium responsible for the surface contamination which adheres to the surface of the elements is assumed to be powder. For the evaluation of the leakage rate, this uranium is supposed to be completely separated and uniformly dispersed in the cavity of the inner shell.
The leakage rate under normal test conditions is calculated by multiplying the concentration of each radioactive nuclide existed in the cavity of the inner shell by the leakage rate calculated in C.3.1.13 (b). By using the JRR-3 standard fuel element, (Uranium silicon aluminum dispersion type alloy) which is highest surface uranium contamination fuel, and by calculating the leakage rate of radioactive substances, the results are obtained as shown in ()-Table C.9.
As shown in the ()-Table C.9, the level of the leakage rate of radioactive substances under normal test conditions is lower than the standard value.
()-Table C.9 Leakage rate of radioactive substances under normal test conditions Nuclide Radioactivity Leakage rate Standard Rate Value (A 2 x10-6)
(TBq/cm3) (TBq/h) (TBq/h) 234 U 1.07x10-13 5.97x10-12 6x10-9 9.96x10-4 235 U 1.56x10-15 8.71x10-14 0 236 U 4.73x10-16 2.64x10-14 6x10-9 4.40x10-6 238 U 1.83x10-17 1.02x10-15 0 Total 1.00x10-3
- : Use 1.48x105cm3 for the inner air volume.
15
(2) In transporting of lowly irradiated fuel element (a) Evaluation of radioactive material in the inner shell concerning leak.
As shown in ()-A, the fuel plate does not failure under the normal test condition, the enriched uranium contained in the fuel plate does not leak.
The radioactive material concerning leak, the surface contaminated uranium, adheres when the fuel element is produced, is similarly assumed as the previous section of C.3.1.2(1), (a). The spectrum converter case is treated as the same assumption.
The water in the reactor is assumed to adhere in a thickness of 1 mm on the all surface of the lowly irradiated fuel element.
Therefore, the radioactive material to be considered in studying the seal function is the radioactive nuclide contained in the water of the reactor.
Since the spectrum converter is not used in the reactor, the reactor water is not considered for the spectrum converter case.
The leak of the radioactive material is evaluated by assuming that the radioactive concentration of the water in the reactor is 12Bq/cm3, which is two times of the maximum value of the measured data of the No.1 canal water, obtained for past twenty years.
The radioactive concentration of the water adheres on the fuel element surface is 12Bq/cm3, the nuclide is 60 Co and shown in ()-Table C.10.
The surface radioactivity per one lowly irradiated fuel element is shown in ()-Table C.11.
()-Table C.10 Nuclide of JMTRC fuel surface water and radioactive concentration Radioactive concentration Nuclide (Bq/cm3) 60 Co 12 16
()-Table C.11 Surface activity per one fuel element of lowly irradiated fuel element Activity (Bq) 234 235 236 238 U U U U Total JMTRC Standard fuel element (Uranium aluminum alloy) 1.19x103 1.74x101 5.28x100 2.04x10-1 1.21x103 (A,B,C type)
JMTRC Standard fuel element (Uranium aluminum alloy) 1.19x103 1.74x101 5.28x100 2.04x10-1 1.21x103 (B,C type)
JMTRC Special fuel element (Special A type) 1.24x103 1.82x101 5.50x100 2.13x10-1 1.26x103 (Uranium aluminum alloy)
JMTRC Special fuel element (Special B type) 4.65x102 6.82x100 2.06x100 7.98x10-2 4.74x102 (Uranium aluminum alloy)
JMTRC Special fuel element (Special C, Special D type) 1.27x103 1.86x101 5.62x100 2.17x10-1 1.29x103 (Uranium aluminum alloy)
JMTRC Control rod fuel Follower (HF type) 8.41x102 1.23x101 3.73x100 1.44x10-1 8.57x102 (Uranium aluminum alloy)
JMTRC Standard fuel element (MA, MB, MC type) 1.19x103 1.74x101 5.28x100 2.04x10-1 1.21x103 (Uranium aluminum dispersion type alloy)
JMTRC Special fuel element (Special MB, Special MC type) 1.25x103 1.83x101 5.55x100 2.14x10-1 1.27x103 (Uranium aluminum dispersion type alloy)
JMTRC Fuel follower (MF type) 8.41x102 1.23x101 3.73x100 1.44x10-1 8.57x102 (Uranium aluminum dispersion type alloy)
Spectrum converter 8.82x101 1.29x100 3.91x10-1 1.52x10-2 8.99x101 17
(b) Radioactive material leak evaluation under normal test condition It is similarly assumed as the previous section of C.3.1.2(1),(a) that the all surface contaminated uranium adheres on the fuel surface is separated and uniformly dispersed in the air in the inner container.
The radioactive concentration of the water adheres on the fuel element surface is 12Bq/cm3, and the nuclide is 60 Co.
The leak rate of the radioactive material under the general test condition is obtained by multiplying the concentration of the nuclide existing in the air of the inner shell by the leak rate obtained in the section of C.3.1.1(2).
The radioactive concentration on the surface of the fuel for the HEU special fuel element (Special C,D types), which has the largest surface area, is shown in ()-Table C.12.
The leak rate of the radioactive material is obtained by assuming that the radioactive material is uniformly dispersed in the air of the inner container of seal boundary, and is shown in ()-Table C.12.
As shown in ()-Table C.12, the leak rate of the radioactive material is smaller than the allowable value under the normal test condition.
()-Table C.12 Leak rate of the radioactivity under normal test condition Radioactive Leak Allowable Nuclide concentration rate Value(A2x10-6) Rate (TBq/cm3) (TBq/h) (TBq/h) 60 Co 1.89x10-12 1.06x10-10 4.0x10-7 2.64x10-4 234 U 8.64x10-14 4.82x10-12 6.0x10-9 8.04x10-4 235 U 1.27x10-15 7.09x10-14 0 236 U 3.83x10-16 2.14x10-14 6.0x10-9 3.56x10-6 238 U 1.48x10-17 8.26x10-16 0 Total 1.07x10-3 18
C.3.2 Pressurization of the containment system Since this package is transported in 'dry' condition, it does not contain any water which becomes a cause of pressurization by the effects of radiation or heat.
Therefore, the only cause of pressurization in the inside part of the package is expansion of the air caused by a temperature rise. This case is explained in ()-Table B.16.
Concerning the analyses of the pressure resistance of the containment system, a safe margin has been taken from the results of internal pressure of the
()-Table B.16 and these analyses have been conducted against the design pressure of the ()-Table B.17.
C.3.3 Coolant contamination Since coolant is not used for this package, this item is not applicable.
C.3.4 Loss of coolant Since coolant is not used for this package, this item is not applicable.
19
C.4 Accident test conditions The integrity of the containment system of this package remains unchanged after an impact under accident test conditions, as required for all types of B(u) packages and the results of structural analyses is shown in ()-A.
Moreover, the results of thermal analyses in ()-B show that variations in pressure or temperature under accident test conditions have no effect on the integrity of the containment system.
Therefore, as the sealing performances of the containment system remain unchanged under accident test conditions, in these analyses, the evaluation of the sealing performances based on the leak tightness test of the O-ring of the inner shell lid which must meet the reference value, conducted before shipment of the package, shows that the leakage rate of radioactive substances under accident test conditions is lower than the legally established standard value.
C.4.1 Fissile gas (1) In transporting fresh fuel element Since the contents are composed of non-irradiated fuel elements, no fissile gas will appear.
(2) In transporting lowly irradiated fuel elements Under the accident test condition, as described in the section ()A.6, since the failure of the fuel element does not occur and the fissile gas contained in the fuel plate does not leak, the enrichment of the fissile gas in the sealed shell is the same value as for the normal test condition, shown in the ()-Table C.10 and in the ()-Table C.11.
20
C.4.2 Leakage of radioactive materials C.4.2.l Leakage from the inner shell The maximum gas leakage rate under the accident test conditions The maximum gas leakage rate under the accident test conditions can be obtained by substituting the relevant values of pressure, gas and the maximum leak hole diameter under the same test conditions into the formula(C3-2) to (C.3-5) and (C. 3-7).
Gas leakage rate under the accident test conditions is shown in (II)-Table C. 13. The maximum gas leakage rate is calculated concerning the inner shell lid.
(II)-Table C. 13 The maximum gas leakage rate under the accident test conditions Position Containment boundary Item (O-ring of inner shell lid)
D: Diameter of leak (cm) 2.490x10-3 a: Length of leak (cm) 0.54
- Viscosity coefficient of the gas(MPa.s) 1.85x10-11 at 25*1 P u,x : Pressure of containment system under normal 0.199 test conditions (MPa abs)
P d,x : External pressure under normal test conditions 0.060 (MPa abs)
T u,x : Gas temperature under normal test 482.9 conditions(K)
M: Molecular weight(g/mol) 29 L x : Leakage rate under normal test conditions 1.38x10-2 (cm3/s at P a,x T a,x )
L s,x : Leakage rate under normal test conditions 1.09x10-2 (cm /s at 25 0.10MPa) (6.59x103cm3/week) 3
- 1: Since the viscosity coefficient of air increases as the temperature rises, it is conservative to use the low temperature.
C.4.2.2 Evaluation of the volume of leakage of radioactive materials (1) In transporting the fresh fuel element and critical assembly fuel As described in Chapter ()-A, since no deterioration of the fuel Plates occurs under accident test conditions, it can be supposed, as in the case of normal test conditions, that the only radioactive substances affected by the 21
leakage are the uranium particles which adhere to surface of the fuel elements during the manufacturing process, i.e. uranium surface contamination.
Surface contamination level per fuel element is shown in the ()- Table C.8.
The leakage rate of radioactive substances under accident test conditions is calculated by multiplying the concentration of each nuclide present in the cavity of the inner shell by the leakage rate calculated in C.4.2.1.
()-Table C.14 gives the leakage rate for radioactive substances for the JRR-3 standard fuel element (Uranium silicon aluminum dispersion type alloy),
which is the highest uranium surface contamination element.
As shown in ()-Table C.14, the leakage rate of radioactive substances under accident test conditions is lower than the standard value.
(2) In transporting the lowly irradiated fuel elements As described in the section ()-A, under the accident condition, since the failure of the fuel plate does not occur, the leakage of the enriched uranium contained in the fuel plate is similarly assumed not to occur as for the normal test condition.
The surface radio activity per one fuel element is shown in ()-Table C.10 and in ()-Table C.11.
The leakage rate of the radioactive substance under the accident condition is obtained by multiplying the enrichment of the nuclide existed in the shell by the leakage rate obtained in the paragraph C.4.2.1.
The radioactive enrichment on the fuel element surface for the HEU special fuel element C, D type, having the maximum surface, is obtained by the same method in the paragraph C.3.1.2.(2) and is shown in the ()-Table C.15. As shown in the ()-Table C.15, the leakage rate of the radioactive substance under the accident condition is less than the reference value.
22
()-Table C.14 Leakage rate of radioactive substances under normal test conditions Nuclide Radioactive Leakage rate Standard Rate substance value concentration (TBq/cm3) (TBq/week) (TBq/week) 234 U 1.07x10-13 7.05x10-10 6x10-2 1.18x10-7 235 U 1.56x10-15 1.03x10-11 0 236 U 4.73x10-16 3.12x10-12 6x10-2 5.20x10-10 238 U 1.83x10-17 1.21x10-13 0 Total 1.18x10-7
()-Table C.15 Leak rate of radioactive substances under accident test condition Nuclide Radioactive Leakage rate Allowable Rate concentration value (TBq/cm3) (TBq/week) (TBq/week) 60 Co 1.89x10-12 1.25x10-8 4.0x10-1 3.11x10-8 234 U 8.64x10-14 5.69x10-10 6.0x10-3 9.49x10-8 235 U 1.27x10-15 8.37x10-12 0 236 U 3.83x10-16 2.52x10-12 6.0x10-3 4.21x10-10 238 U 1.48x10-17 9.75x10-14 0 Total 1.26x10-7 23
C.5 Summary of the results and the evaluation (1) In transporting the fresh fuel element and critical assembly fuel Concerning the leakage of radioactive substances, it may be supposed to all the particles of uranium responsible for the surface contamination which adhere to the surface of the elements during the manufacturing process are completely separated, and that these particles are dispersed uniformly in the air in the inner shell. If the concentration of each radioactive substance is multiplied by the leakage rate to evaluate the leakage rate under normal and accidental test conditions, it can be seen, as shown in ()-Table C.9 and in ()-Table C.14, that the leakage rate for radioactive substances is lower than the standard value.
(2) In transporting lowly irradiated fuel element The leak rate is evaluated under general and special test conditions, by multiplying each radioactive concentration by leak rate, by assuming that the all surface contaminated uranium, adheres when the fuel element is produced, is separated and is uniformly dispersed in the air of the inner shell and also by assuming that the all pool water adheres on the surface of the fuel element is evaporated and uniformly dispersed in the air of the inner shell, the leak rates for both test conditions are smaller than the allowable value, as shown in
()-Table C.12 and ()-Table C.15.
24
C.6 Appendix C.6.1 Design temperature for containment analyses The design temperature for the containment analyses is used for the calculation of the internal pressure of the inner shell, and this pressure is calculated from the average temperature of the air contained in the inner shell.
The volume of the air contained in the fuel basket constitutes the largest proportion (77%) of the total air volume contained in the inner shell, and since there is no emission of heat from the fuel, the temperature of the air contained in the fuel basket is lower than the temperature of the fuel basket.
For greater safety, the temperature of the air contained in the fuel basket is regarded as equivalent to the average temperature of the fuel basket (180.7) and the temperature of the air contained in the space between the fuel basket and the main body of the inner shell is regarded as equivalent to the average of the average temperature of the fuel basket and the average temperature of the main body of the inner shell (403.4), namely a temperature of 292.1, proceeding in this way, the average temperature of the air inside the inner shell can be calculated as 206.3, which is lower than the maximum temperature of the fuel basket (209.9).
As explained above, if the maximum temperature of the fuel basket (209.9) is used as the average temperature of the air contained in the inner shell, the internal pressure of the inner shell is overestimated. Therefore, the maximum temperature of the fuel basket is used as the design temperature for the containment analyses.
- Value obtained by the calculation of the average of the TRUMP CODE 25
C.6.2 Reference documents (1) ANSI-N 14.5 American National Standard for Leakage Tests on Packages for Shipment of Radioactive Materials (1977)
American national Standards Institute, Inc.
American national Standards for radioactive materials Leakage test on packages for shipment (1997)
ANS N14.5 - 1997 (2)Document for Heat Transfer Engineering, EditionMechanical Engineering Society of Japan.
26
()-D Shield analysis
()-D. Shield analysis D.1 Outline In the case where the package contents consist of fresh fuel elements 235 238 (including KUCA fuel), U and U are considered as a gamma radiation source, and neutrons emitted by the uranium spontaneous fission is considered as a neutron source.
In case of the lowly irradiated fuel elements (including Spectrum 235 238 Converter), U, U and the radioactive nuclides are considered as a gamma radiation source, and the uranium spontaneous fission is considered as the neutron source.
Regarding the gamma radiation source calculation, we have to consider that under normal test conditions and accident test conditions, the outer shell is subjected to a transformation and that, under normal transport conditions, with normal test conditions and accident test conditions, the dose-equivalent rate is evaluated by assimilation of the inner shell surface to the package surface.
The neutron dose-equivalent rate is calculated by assimilating the uranium content to the point radiation source. There, the content are distributed inside the cavity, but their position is calculated in such a way that the distance between the point radiation source and the inner shell surface is as small as possible. In the same way, the gamma radiation source calculation is evaluated by considering the inner shell surface to be equivalent to the package surface and for safety reasons, by ignoring the inner shell shield effect and considering only the distance attenuation effect.
D.2 Radiation source specification There are unirradiated fresh fuel element and lowly irradiated fuel elements in the package.
235 238 For unirradiated uranium, the radioactive nuclide such as U and U etc.
are considered as the gamma radiation source.
The neutron emitted by uranium spontaneous fission is considered as the 1
neutron source.
2
In case of the lowly irradiated fuel element, the radioactive element such as the 235U, 238U etc. are considered as the gamma radiation source, and the neutrons emitted by spontaneous fission of uranium etc. are considered as the neutron source.
D.2.1 Gamma radiation source (1) In loading the fresh fuel element The uranium isotope contained in the fuel packaged are 234U, 235U, 236U and 238U, and these gamma ray emitting rates are shown in ()-Table D.1.(1)
The gamma radiation source intensity per one fuel element of the JRR-3 standard type (Uranium silicon aluminum dispersion alloy) (Enrichment 19.75
+/- 0.2wt%), which has highest radioactivity, is sown in ()-Table D.2.
The gamma radiation source intensity is obtained as follows.
S E = CWR E Where, S E : Gamma radiation source intensity (Photons/s) of energy E C : Specific activity (Bq/g), shown in ()-Table D.3(2)
W : Uranium isotope weight (g)
R E : Gamma ray emission rate of energy E (photons/decay)
The weight of the uranium isotope is conservatively obtained as follows.
235 235 (a) U :Maximum U contained quantity in the fuel element.
238 234 236 (b) U :By assuming the quantities of U and U are to be zero, the 238 235 quantity of U is obtained by using the quantity of the U obtained above and the lower limit of the enrich tolerance.
234 236 234 236 (c) U, U:As the quantities of U and U are determined when the fuel element is produced, the weight limit is not determined. Therefore the maximum weight rate is selected from the past material record sheet, by using the conservatively rounded up weight rate, the quantities 235 236 of U and U are obtained. In this case, the necessary 235 total uranium quantity is the sum of the U obtained in 238 (a) and U obtained in (b).
234 236 The weight rates of U and U used in the calculation are shown in ()-Table D.4.
The uranium isotope weight for one element used in calculation is shown in ()-Table D.5.
3
()-Table D.1 Gamma radiation emission rate of uranium isotope Gamma radiation Gamma radiation Uranium energy emission rate isotope (MeV) (photons/decay) 234 U 0.05322 0.00119 0.12090 0.000405 235 U 0.10914 0.015 0.14376 0.105 0.16335 0.047 0.18572 0.54 0.20212 0.010 0.20531 0.047 236 U 0.04937 0.00079 0.11275 0.00019 238 U 0.04955 0.0032
()-Table D.2 Gamma radiation source intensity for one fuel element Gamma radiation Energy Source intensity (MeV)
(photons/s) 0.04937 4.716x104 0.04955 7.948x104 0.05322 3.434x106 0.10914 5.820x105 0.11275 1.134x104 0.12090 1.169x106 0.14376 4.074x106 0.16335 1.824x106 0.18572 2.095x107 0.20212 3.880x105 0.20531 1.824x106 4
()-Table D.3 Specific activity used for calculation Uranium isotope Specific activity (Bq/g) 234 U 2.309x108 235 U 8.001x104 236 U 2.397x106 238 U 1.244x104 234 236
()-Table D.4 U and U weight rate used for calculation Isotope Weight rate (wt%)
Mill sheet Value used for maximum value calculation 234 U 0.13 0.5 236 U 0.21 1.0
()-Table D.5 Radioactive nuclide weight per one element used in calculation Uranium isotope Weight (g) 234 U 12.41 235 U 485 236 U 24.81 238 U 1996 5
(2) In loading lowly irradiated fuel element (a) Gamma radiation source by the isotope from uranium The uranium isotope contained in the package fuel are 234 235 236 238 mainly U, U, U and U etc., and these gamma ray emission rates are shown in ()-Table D.6.(1)
The gamma radiation source intensity of the one equivalent fuel element (mixed fuel elements) of the JMTRC HEU fuel element, which has the highest radioactivity and the MEU fuel element, by assuming these fuel elements are packaged together, is shown in ()-Table D.7.
The gamma radiation source intensity is obtained as follows.
S E = CWR E Where, S E : Gamma radiation source intensity of energy E (Photons/s)
C : Specific activity (Bq/g), shown in ()-Table D.8(2)
W : Weight of Uranium isotope (g)
R E : Gamma ray emission rate of energy E (Photons/decay)
The weight of the uranium isotope is conservatively obtained as follows.
235
() U :Maximum contained quantity in the fuel element.
238
() U :The quantity of 238U is obtained, by assuming the quantities 234 236 235 of U and U are to be zero, by using the quantity of U obtained in (1) and the lower limit of the enrichment tolerance.
234 236 234 236
() U, U:The quantities of U and U are determined when the fuel element is produced, the limit of the weight is not determined. Therefore the maximum weight rate is selected from the past material record sheet, and the 234 236 weights of U and U are obtained by using the conservatively rounded up weight rate.
6
In this case, the necessary total uranium weight is 235 the sum of the weight of U obtained in (1) and the 238 234 weight of U obtained (). The weight rate of U and 236U used in the calculation are shown in ()-Table D.9.
The weight of uranium isotope per one fuel element used in the calculation is shown in ()-Table D.10.
7
()-Table D.6 Gamma radiation emission rate of uranium isotope Gamma radiation Gamma radiation Uranium energy emission rate isotope (MeV) (photons/decay) 234 U 0.05322 0.00119 0.12090 0.000405 235 U 0.10914 0.015 0.14376 0.105 0.16335 0.047 0.18572 0.54 0.20212 0.010 0.20531 0.047 236 U 0.04937 0.00079 0.11275 0.00019 238 U 0.04955 0.0032
()-Table D.7 Gamma radiation source intensity per one mixed fuel element (actinoids)
Gamma radiation Energy source intensity (MeV)
(photons/s) 0.04937 2.321x104 0.04955 1.609x104 0.05322 1.923x106 0.10914 3.804x105 0.11275 5.583x103 0.12090 6.545x105 0.14376 2.663x106 0.16335 1.192x106 0.18572 1.369x107 0.20212 2.536x105 0.20531 1.192x106 8
()-Table D.8 Specific activity used for calculation Uranium isotope Specific activity (Bq/g) 234 U 2.309x108 235 U 8.001x104 236 U 2.397x106 238 U 1.244x104 234 236
()-Table D.9 U and U weight rate used for calculation Weight rate (wt%)
Uranium HEU fuel MEU fuel isotope element element 2.2 (2 times of the 0.47 (1.5 times of 234 U actual contained the actual contained substance weight) substance weight) 0.94 (2 times of the 1.7 (1.5 times of the 236 U actual contained actual contained substance weight) substance weight)
()-Table D.10 Radioactive nuclide weight per one element used in calculation Uranium isotope Weight (g) 234 U 7.00 235 U 317 236 U 12.26 238 U 404 9
(b) Gamma ray from the fission product The irradiation time and the cooling time of the JMTRC fuel element is as follows.
() HEU fuel: 302h irradiation (100W equivalence)(0.0013MWd) 15 years cooling
() MEU fuel: 100h irradiation (100W equivalence)(0.0005MWd) 4 years cooling The fission products are calculated by using ORIGEN for the above 2 types of fuel elements with the following assumption.
The peaking factor of the fuel element during operation is 2.00.
The effect of the radioactivity except the main nuclide is considered by scaling the radioactivity of the main nuclide to be 100%.
The radioactivity of the fission product of the mixed fuel elements is determined from the fuel element which has the higher radioactivity.
The radioactivity and the gamma radiation source intensity of the main nuclides are shown in ()-Table D.11.
The radioactivity of the spectrum converter is 2.43 x 106Bq, assuming that it is equivalent to one fuel element, and the JMTRC fuel element evaluation is more conservative.
10
()-Table D.11 Radioactivity rate of the fission products obtained by ORIGEN Gamma ray Emission Radioactivity Main Scaling energy rate by ORIGEN Photons/s nuclide factor*
(MeV) (%) (Bq) 0.662 85.1 1.11x107 Cs-137 0.0364 1.3 1.261x107 1.70x105 0.0322 5.6 7.33x105 Kr-85 0.514 0.43 1.072x106 4.78x103 1.05 1.6 1.050x103 1.74x101 Rh-106 0.662 9.9 1.08x102 0.512 20.4 2.22x102 0.671 1.8 1.452x104 2.71x102 0.636 11.3 1.70x103 1.0391 0.607 5 7.54x102 0.601 17.9 2.70x103 Sb-125 0.463 10.5 1.58x103 0.428 29.6 4.46x103 0.176 6.8 1.03x103 0.0355 4.3 6.48x102 Sr-90 1.234x107 Y-90 1.7608 0.012 1.234x107 1.54x103 Ba-137m 1.179x107 Total 5.017x107 1.21x107
- Ratio of total radioactivity (5.213x107Bq) by ORIGEN versus radioactivity (5.017x107Bq) of main nuclide.
11
D.2.2 Neutron source (1) In loading fresh fuel element As the contents are non-irradiated uranium, it is necessary to take into consideration that, the neutron emission, occurs by spontaneous fission of uranium is considered as a neutron source.
The uranium isotope spontaneous fission speed is shown in ()-Table D.12(3).
()-Table D.12 Uranium isotope spontaneous fission speed 234 235 236 238 Isotope U U U U Spontaneous fission 3.5x10-3 3.1x10-4 2.8x10-3 7.0x10-3 speed (unit/gs) 238 Since among uranium isotope, the spontaneous fission speed of U is the largest, the intensity of the neutron source for one JRR-3 fuel element (Uranium Silicon Aluminum Dispersion Type Alloy) in which uranium content is largest, reaches maximum. This value is 35.6 (n/s). The neutron source intensity by spontaneous fission is calculated by the following formula.
Sn = W f n In this formula Sn : Neutron source intensity for one fuel element (n/s)
W i : Uranium isotope weight for one element (g)
()-Table D.5 f i : Spontaneous fission speed of uranium isotope (unit/gs)
()-Table D.12 n : No. of neutrons emitted by one core fission(4) (2.5)
The energy spectrum of neutrons emitted by fission is shown in ()-Fig.
D.1(4). The higher the neutron energy, the bigger the calculation factor becomes consequently, to evaluate a dose-equivalent factor in a safe way, the neutron total energy emitted should be 10 MeV.
12
From the result of criticality analysis, the k eff effective multiplication factor of one package containing 10 JRR-3 standard elements (Uranium Silicon 235 Aluminum Dispersion Type Alloy) of 20 wt% enrichment without water whose U content reaches a maximum, and is 0.032 by considering 3 . By the same calculation method, if for safety reasons the effective multiplication rate is fixed to 0.1, it is necessary to consider the multiplication effect of neutrons (1/(1-k eff ) = 1.11 ) on the intensity of neutrons radiation.
()-Fig. D.1 Neutron fission energy spectrum (2) In loading lowly irradiated fuel element It is neutron emission by the spontaneous fission of uranium etc. that is necessary to be considered as the neutron source.
The emission rate of the spontaneous fission of these isotopes is shown in ()-Table D.13.(3) 13
()-Table D.13 Emission rate of spontaneous fission of uranium isotope 234 235 236 238 Uranium isotope U U U U Emission rate of Spontaneous fission 3.5x10-3 3.1x10-4 2.8x10-3 7.0x10-3 (Unit/gs)
The neutron source intensity per one mixed fuel element having the highest radioactivity is maximum of 7.46(n/s)
The neutron source intensity by the spontaneous fission is calculated by the same method as the paragraph D.2.3(1).
14
D.3 Model specification D.3.1 Analysis model (1) Gamma radiation dose-equivalent rate The ANISN code(5) is used for calculation of the gamma radiation shield. The evaluation of the dose-equivalent rate is performed by considering that both under normal test conditions and accident test conditions, the outer shell is subjected to a deformation, and that under routine transport conditions, both under normal and accident test conditions, the inner shell surface is assumed to be the content surface. The gamma radiation shield calculation model is shown ()-Fig.D.2.
The Intensity of gamma radiation is identical to 19.75 wt % enriched JRR-3 standard fuel (Uranium Silicon Aluminum Dispersion Type Alloy), but in order to reduce fuel self shielding, the data of JRR-4L type fuel for which aluminum weight is limited is used, and it is supposed that the source area, for one fuel element, is a 6.8 cm long, 8.0 cm wide and 61.0 cm high rectangular solid.
For the lateral part of the radiation source area model, 10 cylindrical fuel elements with equivalent cross section are evenly distributed. At this time, the shielding effect of the basket is ignored, but as shown in ()-Table B.6, the gap between the fuel basket and the inner shell barrel represents, for the lateral model, a 1.8 cm empty space. In view of this space thickness, the model was realized in order that the source area surface could be as close as possible to the detection point.
Since the detection point is one meter from the packaging surface, and for safety reasons, the dose-equivalent rate is evaluated by the ANISN calculation code, we have to proceed with empty space attenuation effect by the following formula.
15
Supposing that the angles flux of the packaging surface obtained by the ANISN code shield calculation is 4( rs , E, ), it calculates the source flux
( r p , E ) and the source volume rate D at rp calculation point of space shown in ()-Fig.D.3 by the following formula.
dS (r p , E ) = ' (r s , E , ') d (' , ) cos 2 d (D.3-1)
S r
D = E K (E ) (rp , E ) dE (D.3-2)
In this formula, d : Surface element of the packaging surface r : Distance between surface element d s and calculation point
= rp rS K(E): Dose rate conversion factor
- Angle between and n, normal vector of d s Unit vector which indicates the angle between d s and calculation point
' Unit vector which indicates the arbitrary angle direction from d s E Energy 16
(1) Upper portion (2) Middle portion (3) Lower portion
()-Fig.D.2 Gamma radiation shield calculation model 17
()-Fig. D.3 Relationship between packaging surface angles flux and calculation point of packaging surface 18
(2) Neutron dose-equivalent rate Neutron dose-equivalent rate, as it is shown in ()-Fig.D.4, is calculated by considering the content of uranium as the point radiation source. The content is distributed in the cavity, but its position is evaluated so that the distance between the radiation source point and inner shell surface is as small as possible. For safety reasons, the evaluation of the neutron shield calculation is performed considering the surface of the inner shell to be equivalent to the surface of the package. Then, proceeding with the evaluation, for more safety, the shielding effect of the inner shell lid, bottom and barrel parts should be ignored, and only the distance attenuation effect should be taken into consideration.
19
(1) Upper portion (2) Middle portion (3) Lower portion
()-Fig. D.4 Neutron shield calculation model 20
D.3.2 Numeric density of atoms in each area of analysis model Density and material for each zone used for calculation of the gamma radiation shield are shown in ()-Table D.14 and the volumetric rate of shield material for each area is shown in ()-Table D.15. The numeric density of atoms for each shield material is shown in ()-Table D.16.
For neutron dose-equivalent rate, the material of the structure is not taken into consideration and then the following tables are not applicable.
()-Table D.14 Material and density Part name Material Density (g/cm2)
Inner shell lid SUS 630 7.85 Inner shell barrel SUS 304 7.85 Inner shell bottom plate SUS 304 7.85
()-Table D.15 Volumetric rate of shield material for each area used in shield calculation Volumetric Area Shield material rate (%)
Radiation source area Fuel core 16.7 (lateral part evaluation) Cladding 14.3 Cavity 69.0 Radiation source area Fuel core 6.43 (lid and bottom part Cladding 5.51 evaluation)
Cavity 88.06 Inner shell lid Stainless steel (SUS 630) 100 Inner shell barrel Stainless steel (SUS 304) 100 Inner shell bottom plate Stainless steel (SUS 304) 100 21
()-Table D.16 Atom density for each material (atoms/barncm)
Radiation source Radiation source area area Nuclide SUS 304 SUS 630 (lateral part (lid and bottom evaluation) parts evaluation)
C 1.18x10-4 2.76x10-4 Al 1.7985x10-4 6.9293x10-5 Si 1.68x10-3 1.68x10-3 Cr 1.73x10-2 1.50x10-2 Mn 1.72x10-3 8.60x10-4 Fe 5.66x10-2 6.22x10-2 Ni 8.86x10-3 3.22x10-3 Cu 2.98x10-3 235 U 7.0793x10-6 2.7275x10-238 U 7.0793x10-6 2.7275x10-6 22
D.4 Shield evaluation (1) Dose equivalent rate by gamma radiation (a) In loading fresh fuel element The ANISN code is used for the shield calculation for fresh fuel loading.
The cross section of the energy group structure (group 18) of the DLC-23E/CASK library(6) is used as the cross section for the gamma ray.
This energy group structure is shown in ()-Table D.17.
The dose equivalent rate calcuration factor for the gamma ray to obtain the dose equivalent rate(7) is shown in ()-Table D.17.
The calculation result is shown in ()-Table D.18.
As for the increasing rate of the does equivalent rate under the normal test condition, by considering the deformation of the outer shell, the surface of the inner shell is considered to be the package surface in the analysis under the usual transport condition, normal test condition and accident condition, so, the does equivalent rate does not increase and is within the allowable value.
23
()-Table D.17 Gamma radiation energy group structure and dose-equivalent rate calculation factor Dose-equivalent rate Upper limit energy Energy groups calculation factor (eV)
((mSv/h)/(/cm2s))
1 1.00x107 8.4944x10-5 2 8.00x106 7.2388x10-5 3 6.50x106 6.1456x10-5 4 5.00x106 5.2036x10-5 5 4.00x106 4.4163x10-5 6 3.00x106 3.7842x10-5 7 2.50x106 3.3385x10-5 8 2.00x106 2.8967x10-5 9 1.66x106 2.4817x10-5 10 1.33x106 2.0800x10-5 11 1.00x106 1.7275x10-5 12 8.00x105 1.4112x10-5 13 6.00x105 1.0523x10-5 14 4.00x105 7.5325x10-6 15 3.00x105 5.4060x10-6 16 2.00x105 3.2205x10-6 17 1.00x105 1.9332x10-6 18 5.00x104 2.6973x10-6 1.00x104 24
()-Table D.18 Dose-equivalent rate by gamma radiation (fresh fuel elements loading)
Dose-equivalent rate Evaluated position (mSv/h)
Package surface Lid <0.001 Side 0.033 Bottom 0.003 1m apart from package surface Lid <0.001 Side 0.004 Bottom <0.001 (2) In loading lowly irradiated fuel element The shield analysis of the gamma radiation for the case where the lowly irradiated fuels are loaded, is conducted by the same method described in the section of previous (1)(a). The result of the analysis is shown in ()-Table D.19.
As for the increase rate of the dose-equivalent rate under the general test condition, the surface of the inner shell is evaluated as the surface of the package under the usual transport condition, the normal test condition and the accident test condition, by considering that the outer shell is deformed under the normal test condition, therefore the increase of the dose-equivalent rate does not occur and satisfies the criteria.
()-Table D.19 Dose-equivalent rate by gamma radiation (lowly irradiated fuel elements loading)
Dose-equivalent (mSv/h) Total Evaluated position Actinides FP (mSv/h)
Package Lid <0.001 0.025 0.026 surface Side 0.022 0.145 0.167 Bottom 0.002 0.069 0.071 1m apart Lid <0.001 0.005 0.006 from package Side 0.003 0.015 0.018 surface Bottom <0.001 0.013 0.014 25
(3) Neutron dose-equivalent rate (a) In loading fresh fuel element Neutrons dose-equivalent rate is calculated by following formula.
S nn Dn = A x xk 4 r 2 In this formula, D n : Dose-equivalent rate (mSv/h)
S n : Neutron source intensity for one fuel element 35.6 (n/s) n : Number of fuel elements for one packaging 10 r : Distance from point radiation source to evaluation point (cm) k : Neutron multiplication effect 1.11 A : Conversion facter of Dose-equivalent rate of 10 MeV energy neutron flux(7) 0.00159 ((mSv/h)/(n/cm2s))
The calculation result of neutron dose-equivalent rate is shown in ()-Table D.20
()-Table D.20 Neutron dose-equivalent rate Calculation result Dose-equivalent rate Evaluated position (mSv/h)
Package surface Lid part 0.002 Middle part 0.007 Bottom part 0.005 Position at one meter Lid part <0.001 from container surface Middle part <0.001 Bottom part <0.001 26
(b) In loading the lowly irradiated fuel element The does-equivalent rate of the lowly irradiated fuel element loading is calculated by the same method in the previous section of (2)(a).
The analysis result of the does-equivalent rate of the lowly irradiated fuel element loading is shown in ()-Table D.21.
()-Table D.21 Dose-equivalent rate of neutron irradiation (lowly irradiated fuel elements loading)
Dose-equivalent rate Evaluated position (mSv/h)
Package surface Lid <0.001 Middle 0.002 Bottom <0.001 1m apart from Lid <0.001 package surface Middle <0.001 Bottom <0.001 27
D.5 Summary of the results and evaluation Dose-equivalent rate results obtained with the present package shield analysis for the fresh fuel element and the lowly irradiated fuel element are shown in ()-Table D.22. and in ()-Table D.23. Gamma radiation dose-equivalent rate is calculated with the one dimensional discrete ordinates transport code ANISN, neutron dose-equivalent rate is easily calculated by using the model of point radiation source.
As shown in ()-Table D.22, and in ()-Table D.23, the result of calculation always satisfies the standard values.
()-Table D.22 Package dose-equivalent rate (fresh fuel element loading) (unit: mSv/h)
Evaluated position Position at one meter Package surface from the packaging surface Item Middle Lid Bottom Middle Lid Bottom Routine Gamma radiation 0.033 <0.001 0.003 0.004 <0.001 <0.001 transport condition Neutron 0.007 0.002 0.005 <0.001 <0.001 <0.001 Total 0.040 0.003 0.007 0.005 0.002 0.002 Standard value 2 or less 0.1 or less Normal test Gamma radiation 0.033 <0.001 0.003 condition Neutron 0.007 0.002 0.005 Total 0.040 0.003 0.008 Standard value 2 or less Accident Gamma radiation 0.004 <0.001 <0.001 test condition Neutron <0.001 <0.001 <0.001 Total 0.005 0.002 0.002 Standard value 10 or less 28
()-Table D.23 Package Dose-equivalent Rate (lowly irradiated fuel element loading) (unit: mSv/h)
Evaluation point Position at one meter Package surface from the packaging surface Item Middle Lid Bottom Middle Lid Bottom Routine Gamma radiation 0.167 0.026 0.071 0.018 0.006 0.014 transport condition Neutron 0.002 <0.001 <0.001 <0.001 <0.001 <0.001 Total 0.169 0.027 0.072 0.019 0.007 0.015 Standard value 2 or less 0.1 or less Normal test Gamma radiation 0.167 0.026 0.071 condition Neutron 0.002 <0.001 <0.001 Total 0.169 0.027 0.072 Standard value 2 or less Accident Gamma 0.018 0.006 0.014 test radiation condition Neutron <0.001 <0.001 <0.001 Total 0.019 0.007 0.015 Standard value 10 or less 29
D.6 Appendix D.6.1 Explanations of ANISN code **************************** ()-D-30 D.6.2 Reference literature ********************************** ()-D-33 30
D.6.1 Explanations of ANISN code The ANISN code developed by ORNL in the USA, is a numerical calculation of the one dimensional Boltzmann transport equation based upon Discrete Ordinates Sn.
The transport equation is a mathematical representation of the balance between formation and disintegration of particles inside a volume element phase space resulting from position, energy and the direction of progression, the equation is given by the following formula.
(r,E,)t(r,E)(r,E)
= (r,E,)s(r,E E, ) dE,dS(r,E,)(D.6-1) where, (r,E,) Angle neutron flux (number of particles passing per unit time through the surface perpendicular to the unit vector and per unit solid angle in the direction of unit vector at position r) t(r,E) Total macro cross section s(r,EE,) Dispersion macro cross section or creation of a macro cross section of secondary gamma radiation from neutrons S(r,E,) External radiation source The Sn method is a numeric evaluation of the transport equation discretely dealing with position, energy and direction of progression. It is called the Sn method because of the special way evaluating the angle division point(Sn division point).
This technique uses the fundamental cell to express the transport equation for the direction of progression of each energy group, then calculates until convergence, by iterations of the difference equation.
31
To express the primary transport equation (r , r ), ( ,
) with the adjacent mesh that determines the fundamental cell (see ()-Fig.D.5 below)
W(A N - A i N i ) N - N
=V(S - t)NW ************************************ (D.6-2)
()-Fig.D.5 Mesh distribution drawing
- Where, N Neutron flux (including angles distribution)
(for each energy group)
- Cosine A Surface factor for flat plate shape: 1.0 for cylindrical shape: 2r for circular shape: 4r2 W : Weight coefficient of direction cosines W = 1.0 32
V : Volume factor for flat plate shape : -
for cylindrical shape : 2 - 2 for circular shape : 4/3 (r3 - r 3) t: Total cross section S : Radiation source term (external radiation source + dispersion integral term)
- Value given by the following formula
= - W(A A )
= 0.0 The formula (D.6-2) is obtained by multiplying the phase space to (D.6-1) formula, integrating it and substituting the differential value to difference value.
The formula (D.6-2) includes 5 unknown variables (N, N N , N
, N ). To reduce the number of unknown variables, diamond difference calculation method or approximation step function can be used.
Diamond difference calculation: Linear approximation at adjacent meshes intermediate point.
N = 1/2 (N N )=1/2N N Step function approximation : N = N = N for < 0 N = N = N for > 0 For the diamond difference calculation, in case > 0 2
2AN i N N -1/2 SV N = W (D.6-3) 2 2A tV W
33
Then,
= 1/2( n1/2 n1/2 )
A = 1/2(A A )
To calculate this difference equation, an initial value is assigned, then the equation calculated iteratively until it converges. This gives the basic solution.
D.6.2 Reference literature (1) Murakami Yukio:Radioactivity Data BookChijinshokan (1982).
(2) IAEA Safety Guides : Advisory Material for the IAEA Regulations for the Safe Transport of Redioactive Material(1985) IAEA Safety Series No.37 (1985)
(3) Ethesington :Nuclear Engineering Handbook(1965)
(4) Nuclear Handbook. Glaston (1965)
(5) ORNL/RSIC Computer Code Collection ANISN-WA One Dimensional Discrete Ordinates Transport CodeCCC-82 (6) RSIC Data Library Collection DLC-23Cask 40 Group Coupled Neutron and Gamma-Ray Cross Section Data (7) Japan Isotope Association:Conversion factor for use in Radiological Protection against External Radiation ICRP Publication 74 (1998) 34
()-E Criticality analysis
()-E. Criticality analysis E.1 General The criticality analysis on the present package is performed to demonstrate compliance of the package with the technical standards in accordance with the following Regulations:
(a)The Regulations Regarding the Transporting of the Nuclear Fuel Material etc. Outside of the Factory or Workshop(Ordinance No. 57 dated on Dec. 28, 1978 of the Prime Ministers Office, Ordinance No.1 dated on June 15, 2001 of Ministry of Education, Culture, Sports, Science and Technology, Ministry of Economy, Trade and Industry and Ministry of Land, Infrastructure and Transport) (hereinafter referred to as Ordinance) and (b)The Notification Stipulating the Particulars Concerning the Technical Standards for the Transportation of Nuclear Fuel Materials etc. Outside of the Factory or Workshop (Notification No. 11 dated on Dec. 18, 1978 of Science and Technology Agency, Notification No.1 dated on June 15, 2001 of Ministry of Education, Culture, Sports, Science and Technology, Ministry of Economy, Trade and Industry and Ministry of Land, Infrastructure and Transport)
(hereinafter referred to as Notification) 21 types of fuel elements are contained in this package. The numbers of the fuel elements contained in one package is 10. For KUCA fuel, the coupon type has a maximum of 120 plates as one fuel element, and the flat type has a maximum of 30 plates as one fuel element, and the numbers of the fuel elements contained in one package is 10. In addition, one spectrum converter will be stored in the upper part of the fuel basket.
In this analysis, the criticality analysis is conducted for the case where the ten types of fuel elements, excluding the fuel follower, special fuel and 235 half-loaded fuel element, are contained. The weight of contained U per one fuel follower, special fuel and half-loaded fuel element is equal or less than the standard fuel element, therefore, the effective multiplication constant for the package becomes small, and the analysis is not conducted.
For the criticality analysis of KUCA fuel, the coupon type is treated a maximum of 120 plates as one fuel element, and the flat type is treated a maximum of 1
30 plates as one fuel element.
For the spectrum converter, the critical analysis is performed assuming that one spectrum converter is housed in the upper part of the fuel basket since the soundness of the package is maintained as shown in () -Chapter A.
As for the JMTRC fuel elements, two types of fuels of different enrichment (MEU, HEU fuels), are contained and transported. In this analysis, the subcriticality is also confirmed for containing MEU fuel elements and ten HEU fuel elements, and in addition, for containing five HEU fuel elements and five MEU fuel elements as the case of mixed sample.
2
E.2 Parts to be analyzed E.2.1 Content The package is designed to contain ten box-type fuel elements maximum as shown in ()-Table E.1. All fuel elements to be loaded have the same 235 enrichment. The maximum mass of U loaded in a package is 4.85 kg, which corresponds to the JRR-3 standard type fuel element (Uranium Silicon Aluminum Dispersion Type Alloy). The fuel element is composed of the fuel plate which has a fuel meat made of an uranium-aluminum-silicon dispersion alloy. The uranium-aluminum dispersion alloy, the uranium-aluminum-silicon dispersion alloy or the uranium-molybdenum-aluminum dispersion alloy is covered with the aluminum alloy cladding. The spectrum converter is a disk-shaped plate in which the fuel core material of uranium dioxide is coated with an aluminum alloy.
The specifications of fuel plate are shown in ()-Table E.2.
E.2.2 Packaging As described in (I)-A.9, a part of shock absorber and heat insulator of outer shell is deformed under normal test conditions concerning fissile package, but there is no deformation of inner shell, affecting criticality analysis.
Fuel elements or inner shell is not damaged while a part of shock absorber and heat insulator is deformed, under the accident test conditions concerning fissile package.
Therefore, this analysis model, excluding conservatively shock absorber and heat insulator as mentioned in (II)-E. 3.1, can be applied to the undamaged package during transport and the damaged package under the normal test conditions as well as the accident test conditions concerning the fissile package.
(II)-Table E.3 shows the deformation and remaining thickness of the shock absorber under normal transport conditions as well as under normal and accident test conditions of the fissile package.
3
()-Table E.1 Specification of fuel element 235 Item Total Length Cross Section U Maximum Number of Fuel Mass of 235U Enrichment Elements Loaded in Remark (g/one fuel element)
Fuel element (mm) (mm) (wt%) a Package JRR-3 standard type (Uranium silicon aluminum 1150 76.2x76.2 19.95 485 10 dispersion type alloy)
JRR-3 follower type (Uranium silicon aluminum 880 63.6x63.6 19.95 310 10 dispersion type alloy)
JRR-4 B type fuel element 1025 80.0x80.0 93.3 170 10 JRR-4 L type fuel element 1025 80.0x80.0 19.95 230 10 JRR-4 (Uranium silicon aluminum 1025 80.0x80.0 19.95 210 10 dispersion type alloy) 46.0 320 10 MEU 4
JMTR standard fuel element 1200 77.0x77.0 19.95 425 10 LEU JMTR fuel follower 890 64.0x64.0 19.95 280 10 LEU A 285 B 800 90.0 242 10 HEU JMTRC C 199 77.0x77.0 standard fuel element A 317 B 800 46.0 286 10 MEU C 255 90.0 199 10 HEU JMTRC fuel follower 800 64.0x64.0 46.0 210 10 MEU A 970 77.0x77.0 199 B 435 65.7x65.7 90.0 67 HEU JMTRC C 242 10 special fuel element D 285 970 77.0x77.0 B 46.0 286 MEU C 255 KUR Standard fuel element 873.1 75.40x79.18 19.95 218 10 LEU KUR Half-loaded fuel element 952.5 75.40x79.18 19.95 109 10 LEU KUR Special fuel element 873.1 75.40x79.18 19.95 109 10 LEU KUCA coupon fuel - - 19.95 4(one coupon) 1200 LEU KUCA flat fuel - - 19.95 15(one plate) 300 LEU Spectrum converter 310 mm in diam. x 10.7 t 90 100.2(as one element) 10 HEU 3
()-Table E.2 Specification of fuel plate (1/2)
Item Fuel plate Fuel plate Clad Weight per Fuel plate width thickness thickness one fuel total length Remark Name of fuel plate (mm) elements (mm) (mm) (mm) (g)
JRR-3 standard fuel element (Uranium silicon 770 71.4 1.27 0.38 279 aluminum dispersion type alloy)
JRR-3 follower type fuel element (Uranium silicon 770 59.4 1.27 0.38 228 aluminum dispersion alloy)
JRR-4B type fuel element Outer fuel 734 74.5 1.26 0.38 189 plate Inner fuel 630 74.5 1.26 0.38 171 plate JRR-4L type fuel element Outer fuel 734 74.5 1.65 0.38 270 plate Inner fuel 630 74.5 1.65 0.38 266 plate JRR-4 type fuel element Outer fuel 734 74.5 1.26 0.38 262 (Uranium silicon plate aluminum dispersion Inner fuel 630 74.5 1.26 0.38 235 alloy) plate JMTR-standard fuel 271 MEU element 778 70.8 1.27 0.385 287 LEU JMTR-fuel follower 769 58.9 1.27 0.385 235 LEU JMTRC-standard fuel A 204 element B 775 0.380 201 HEU C 199 70.8 1.27 A 212 B 778 0.385 209 MEU C 206 JMTRC-fuel follower 780 58.5 0.380 171 HEU 1.27 750 57.9 0.385 168 MEU JMTRC-special fuel A 778 70.8 199 element B 385 58.5 71 1.27 0.38 HEU C 201 800 70.6 D 204 B 209 800 70.8 1.27 0.385 MEU C 205 KUR Standard fuel element 676.3 max 69.93 1.52 0.51 235 LEU KUR Half-loaded fuel element 676.3 max 69.93 1.52 0.51 235 LEU KUR Special fuel element 676.3 max 69.93 1.52 0.51 235 LEU KUCA coupon fuel 50.8 50.8 2.3 0.4 30 LEU KUCA flat fuel 600 62 1.5 0.5 190 LEU Spectrum converter 310 mm in diam. x 10.7 t 0.7 - HEU 5
()-Table E.2 Specification of fuel plate (2/2)
Item Weight Fuel plate Fuel plate Fuel plate of 235U per core core width core Name of fuel one fuel length (mm) thickness Fuel plate core material Remark element plate (mm) (mm)
(g)
JRR-3 standard fuel Uranium silicon Element (Uranium silicon aluminum dispersion 23.1 750 62.0 0.51 aluminum dispersion alloy alloy)
JRR-3 follower type fuel Uranium silicon element (Uranium silicon aluminum dispersion alloy 18.2 750 49.0 0.51 aluminum dispersion alloy)
JRR-4B type fuel element Uranium aluminum Outer fuel 6.0 alloy plate 600 68.0 0.50 Inner fuel 11.9 plate JRR-4L type fuel element Uranium aluminum Outer fuel 7.4 dispersion alloy plate 600 65.4 0.89 Inner fuel 14.8 plate JRR-4 type fuel element Uranium silicon Outer fuel (Uranium silicon 7.5 aluminum dispersion alloy plate aluminum dispersion 600 65.4 0.50 Inner fuel alloy) 15.0 plate JMTR-standard fuel Uranium aluminum element 16.8 0.50 dispersion type MEU 759 61.6 alloy Uranium silicon 22.4 0.51 LEU aluminum dispersion alloy JMTR-fuel follower Uranium silicon 17.5 750 49.7 0.50 LEU aluminum dispersion alloy JMTRC-standard fuel A 15.0 Uranium aluminum element B 12.7 750 58.0 0.508 alloy HEU C 10.5 A 16.7 Uranium aluminum dispersion alloy B 15.1 759 61.6 0.50 MEU C 13.4 JMTRC-fuel follower Uranium aluminum 12.4 762 45.5 0.51 HEU alloy Uranium aluminum 13.1 730 49.7 0.50 MEU dispersion alloy JMTRC-special fuel A 10.5 750 61.8 Uranium aluminum element alloy B 4.2 375 49.9 0.51 HEU C 12.7 750 61.8 D 15.0 B 15.1 Uranium aluminum 759 61.6 0.50 dispersion alloy MEU C 13.4 Uranium silicon aluminum KUR Standard fuel element 11.83 594.0 63.0 0.50 LEU dispersion alloy Uranium silicon aluminum KUR Half-loaded fuel element 11.83 594.0 63.0 0.50 LEU dispersion alloy Uranium silicon aluminum KUR Special fuel element 11.83 594.0 63.0 0.50 LEU dispersion alloy Uranium molybdenum KUCA coupon fuel 4 44.8 44.8 1.45 LEU aluminum dispersion alloy Uranium silicon aluminum KUCA flat fuel 15 570 56 0.5 LEU dispersion alloy 1002(one Spectrum converter 254mm in diam.x 0.922 t Uranium dioxide HEU body) 6
()-Table E.3 Distance from the surface of the inner shell to the surface of the packaging (Unit : mm)
Conditions Normal transport Normal test Accident test condition condition condition for fissile (undamaged package) for fissile packages packages Item Distance from the surface of the inner shell to that of the 180 180 180 packaging Deformation 0 34.8 102.7 Remained thickness 180 145.2*1 77.3*1
- 1 In the damage system, it suppose distance from the pestle surface to the transportation container surface to be zero.
E.2.3 Neutron absorbing materials The packaging is designed to use no neutron absorbing materials.
7
E.3 Model specification E.3.1 Calculation model This packaging is designed to contain 20 types of rectangular fuel elements and disk-shaped spectrum converter. The fuel follower contains less U235 per fuel element, compared with the standard type fuel element, so that the effective multiplication factor of the packaging will become smaller, and consequently we will analyze, here, 9 kinds of fuel elements, excluding the fuel follower and the special fuel element. The KUCA coupon and flat fuel are included in the analysis. In addition, the spectrum 235 converter is also including in the analysis although U content is low.
In the evaluation of subcriticality, under the assumption that all of the gap existing inside and outside of the packaging are filled with water, investigation will be conducted to select the package under severest condition among the damaged package and undamaged package in isolation and in arrays so that the analysis is to be executed under the severest conditions.
The damaged system used here is defined as the state of the package under general test conditions and special test conditions, and the undamaged system is defined as the state of the packaged sound.
(1) Package in isolation (damaged package vs. undamaged package)
As for the packages in isolation, the zone surrounding the packaging of undamaged package consists of insulaling material and the damaged packages are assumed as those having insulation taken out, to be replaced by water.
In this context, the neutron reflecting effect and neutron moderating effect of the water are greater than those of insulating material so that the conditions to which the damaged packages to be subjected will be severer since they have larger neutron reflecting effect and moderating effect.
(2) Arrays of packages (damaged package vs. undamaged package)
In the arrays of packages, the damaged packages which have no insulating material will be subjected to the severer conditions, compared with the 8
undamaged packages, because the distances between the adjacent packaging in the arrays of packages are smaller and the neutron mutual interference effect is larger.
(3) Damaged packages in isolation vs. damaged packages in array As for the damaged packages in isolation and in array, in case of packaging being filled by water, the neutrons will be sufficiently moderated in this model, and the extent of neutron moderation will be almost same in both of the cases, and the arrays of packages of perfect reflection with no leaks of neutrons at all will be subjected to severer results than the packages in isolation in which the neutrons leaks are considered smaller, taking the reflecting effect into account.
Consequently in this analysis the arrays of packages in radial direction will be taken, as shown in ()-Fig.E.1 and ()-Fig.E.2, as a triangular lattice type having the most densely arranged infinitive arrays composed of packaging having external shock absorber and insulating materials removed completely. In the axial direction, the evaluation will be conducted on the analysis model of damaged packages in array placed under the severest condition having infinite length of fuel part.
Therefore, the moderation of neutrons is at the same level in packages in isolation and those in array.
Packages in array in which no leakage of neutrons is supposed to occur may be subjected to more severe conditions than those in isolation in which less leakage of neutrons is supposed to occur because of the reflecting effect of the water.
Requirements defined in the regulation and analysis conditions is shown in ()-Table E.4 9
()-Fig.E.3 (box type fuel element) shows the model of the fuel element loaded in the inner shell. The inner shell is filled and surrounded with water, the density of which is 1.0g/cm3. Any structure materials except fuel baskets in inner shell are replaced by water to neglect neutron absorption by these materials.
As for the JMTRC fuels, two kinds of fuels of different enrichment are mixed in the package, as a sample of this case, ()-Fig.E.4 shows the criticality analysis model for mixed fuels.
Calculation model of 9-types fuel elements used in these analyses are shown in ()-Fig.E.5 to ()-Fig.E.14. The model of both JMTR standard fuel elements (LEU and MEU) is the same except for fuel meat compositions. JRR
-4 B, JRR-4 L type fuel elements and JRR-4 fuel elements (Uranium Silicon Aluminum Dispersion Type Alloy) have the outer fuel plates which contain less amount of fissile than inner plates.
These outer plates are conservatively assumed to be the same with inner plates.
For the KUCA fuels, the analysis was conducted assuming the fuel core part and the aluminum cladding part were homogenized and the fuel plate was spread evenly throughout the square pipe of the basket. The coupon fuel was evenly arranged in the vertical direction in the square pipe, and the flat fuel was arranged in the horizontal direction in the square pipe.
For the spectrum converter, the analysis was performed assuming one spectrum converter fuel core is set on the upper part of the fuel basket.
E.3.2 Regional densities for each analyzed model region Atomic number density used in the calculation models of the package and the fuel elements are shown in ()-Table E.5 and ()-Table E.6 respectively.
235 Conservatively, the maximum value of enrichment of U considering the tolerance is assumed for each fuel element.
10
Boundary Conditions (full reflection) Boundary Conditions (full reflection)
Water Water Water Water Water Water Inner Shell trunk Boundary Conditions (full reflection)
In this model of criticality, 60cm is chosen for the axial length and full reflection is supposed for the boundary conditions.
()-Fig.E.1 Calculation model of arrayed packages for criticality with 10 box type fuel elements (except KUR) 11
Boundary Conditions (full reflection)
Water Spectrum converter position Water Water KUR KUR KUR KUCA KUCA KUCA Boundary Boundary Conditions KUR KUR KUR KUR Conditions KUCA KUCA KUCA KUCA (full reflection) (full reflection)
Water KUR KUR KUR Water KUCA KUCA KUCA Inner Shell Boundary Conditions (full reflection)
Water In this model of criticality, 60cm in the axial length is chosen for KUR and KUCA flat fuel, 120cm in the axial length is chosen for KUCA coupon fuel and the spectrum converter (SC), and full reflection is supposed for the boundary conditions. In the calculation, metal spacer was not included. The spectrum converter is set at the top of the fuel basket.
()-Fig.E.2 Calculation model of arrayed packages for criticality with 10 box type fuel elements (KUR, KUCA fuels and spectrum converter) 12
JRR-3 JRR-4 JMTR JMTRC KUR KUCA fuel
()-Fig.E.3 Calculation model of package for criticality with 10 box type fuel elements 13
()-Fig.E.4 Calculation model of package for criticality with HEU and MEU 14
()-Fig.E.5 Criticality calculation model of JRR-3 standard fuel element 15
()-Fig.E.6 Criticality calculation model of JRR-4B type fuel element 16
()-Fig.E.7 Criticality calculation model of JRR-4L type fuel element 17
()-Fig.E.8 Criticality calculation model JRR-4 type fuel element 18
()-Fig.E.9 Criticality calculation model of JMTR standard type fuel element 19
()-Fig.E.10 Criticality calculation model of JMTRC standard type fuel element (HEU) 20
()-Fig.E.11 Criticality calculation model of JMTRC standard type fuel element (MEU) 21
Thickness 0.152 Pitch 0.281 Side Plate Fuel Plate water water Fuel Meat Fuel Plate
()-Fig.E.12 Criticality calculation model of KUR standard type fuel element 22
Fuel basket9.4 cm square Coupon 5.08cm square Thickness: 2.4 mm 120 cm Water The fuel core part and the aluminum cladding part were homogenized.
The 120 fuel plates were spread evenly.
The gap of fuels: 0.77 cm The area except for the fuels is water.
()-Fig.E.13 Criticality calculation model of KUCA coupon type fuel 23
Fuel basket9.4 cm square 60 cm The 30 fuel plates were spread evenly.
The gap of fuels:
1.633 cm Water The fuel core part and the aluminum cladding part were homogenized.
()-Fig.E.14 Criticality calculation model of KUCA flat type fuel 24
()-Table E.4 Requirements defined in the regulation and analysis conditions Requirement defined in the regulation Analysis condition Infiltration of Infiltration of Placement of the Transport water into the Approach of water water into the Approach of water Conditions transported product transported reflection transported reflection materials articles articles
- 1. Normal Transportation None None Conditions 2.
1 pc Available Available Independent A
- 3. General 1 pc triangular-lattic test Available Available This is assessed with Isolation e type model, in condition an infinite number, which inner shells
- 4. Special 1 pc Available which is stricter than are infinitely test Isolation Available Available proximity/reflection most condition of water.
densely-arranged,
- 5. General 5N pc* was adopted test (Array No requirements Available condition
- 6. Special 2N pc*
test (Array No requirements Available condition
- : N is Transport limited number. In this transport shell N=Infinite 25
()-Table E.5 Atom density of regions used in criticality calculation (atoms/barncm)
Inner shell Water Nuclide and pipe of (1.0g/cm3) fuel basket H 6.686x10-2 O 3.343x10-2 Cr 1.727x10-2 Mn 1.721x10-3 Fe 5.905x10-2 Ni 7.449x10-3 26
()-Table E.6 Atom density of fuel element used in criticality calculation (atoms/barncm)
JRR-3 Standard JRR-4 JRR-4 L Type Fuel Fuel Element JMTR JMTR JMTRC JMTRC KUR Type JRR-4 Element (Uranium (MEU) (LEU) (HEU) (MEU) (LEU)
Nuc (Uranium B Type Cladd-
-lide Silicon Fuel (Uranium Silicon Standard Standard Standard Standard Standard ing Aluminum Aluminum Fuel Fuel Fuel Fuel Fuel Aluminum Element Dispersion Dispersion Element Element Element Element Element Dispersion Alloy) Alloy) Alloy) 3.0820 5.6159 4.7729 3.8562 5.2434 3.3440 6.0614 5.0660 4.353 5.9922 Al x10-2 x10-2 x10-2 x10-2 x10-2 x10-2 x10-2 x10-2 x10 -2 x10-2 8.6527 6.4118 8.2830 4.932 5.7890 Si 0 0 0 0 0 x10-3 x10-3 x10-3 x10 -3 x10-5 1.3387 Fe 0 0 0 0 0 0 0 0 0 x10-4 235 2.4952 1.5415 1.1563 1.8490 1.8459 2.4515 1.7397 1.8511 1.661 U 0 x10-3 x10-3 x10-3 x10-3 x10-3 x10-3 x10-3 x10-3 x10-3 238 9.8853 1.0930 4.5811 7.3249 2.1395 9.7126 1.9890 2.3294 6.741 U 0 x10-3 x10-3 x10-3 x10-3 x10-3 x10-3 x10-4 x10-3 x10-3 Nuc- KUCA Nuc Nuc Spectrum lide KUCA flat Coupon -lide -lide converter 6.0262 6.0262 O 1.22084 Al Al x10-2 x10-2 x10-2 235 1.5956 2.3213 U 5.49504 Mo Si x10-3 x10-2 x10-3 238 235 1.7266 235 7.0175 U 6.07487 U U x10-3 x10-4 x10-4 238 6.8407 238 2.7802 U U x10-3 x10-3 27
E.4 Evaluation for subcriticality E.4.1 Calculation conditions (1) Content
()-Table E.7 shows the 11 kinds of fuel elements, the content of packaging to be analyzed.
()-Table E.7 Fuel elements to be analyzed Item Enrichment Maximum number of U235 of elements Fuel element (wt%)* per package JRR-3 standard type fuel element (Uranium 20 10 Silicon Aluminum Dispersion Type Alloy)
JRR-4 B type fuel element 93 10 JRR-4 L type fuel element 20 10 JRR-4 fuel element (Uranium Silicon Aluminum 20 10 Dispersion Type Alloy)
JMTR standard type fuel element 46 10 JMTR standard type fuel element 20 10 JMTRC standard type fuel element 90 10 JMTRC standard type fuel element 46 10 KUR standard type fuel element 19.95 10 KUCA coupon fuel 19.95 1200 KUCA flat fuel 19.95 300 Spectrum converter 90 1 (body)
- Nominal value 28
(2) Packaging We evaluated the packaging on the assumption that the surface of the inner shell is the surface of the packaging (see ()-Fig.E.3)
E.4.2 Water Immersion into package In the criticality analysis, the inside of the inner container was evaluated as being filled with water so as to include the evaluation of isolated systems and sequence systems. The maximum Keff is observed at the density of water about 0.02g/cm3, and even in this case, the package is maintained subcritical.
In this calculation the displacement of package or temperature change due to water immersion is ignored. The influence of the packaging material is included in the change in water density.
The evaluation of Optimum Moderating Water Density is shown in E.7.1.
Appendix.
E.4.3 Calculation method Criticality calculations are performed using a combination of the KENO-
.a Monte Carlo computer code[1] with the 137-energy group MGCL neutron cross-section library(2). The explanations of KENO-V.a and MGCL is shown in E.6.2 and E.6.3. The slab geometry Dancoff-Ginsberg correction factor is considered in calculating the resonance self-shielding effects with MAIL code(1) included in the MGCL.
For the KUR fuel element (including KUCA fuel), criticality calculations were performed using the SCALE code system(3). KENO-VI Monte Carlo module together with the 238-energy group ENDF/B-V neutron cross section library of the SCALE code system was used for the calculation of keff. The resonance self-shielding effects were treated using the BONAMI and CENTRM modules of the SCALE code system. The explanation of KENO-VI code is shown in E.6.2.
()-Fig.E.15 shows the procedure of the calculation.
29
For Fuel elements
<for Fuel elementsexcept For KUR, Fuel
<for KUR KUCAelements fuels >and SC except KUR>
KUR and KUCA fuel SCALE code system MGCL 137 ENDF/B-V Group Library 238 group library MAIL Generate Macroscopic Effective Cross-Section BONAMI CENTRM (Resonance Self-shielding)
Macroscopic Effective Cross-Section eff KENO-V.a KENO-VI Monte Carlo Monte Carlo Criticality Calculation Criticality Calculation Effective Effective Multiplication Multiplication Factor Factor
()-Fig.E.15 Schematic flow of criticality analysis 30
E.4.4 Results In the evaluation of subcriticality, arrays of damaged packages were analyzed which could be subjected to the most severe conditions (Section E.3.1).
()-Table E.8 shows the calculation results of the effective multiplication factor in arrays of damaged packages under submergence.
The maximum K +/- is 0.902+/-0.005 (standard deviation of the Monte Carlo calculation) with JRR-3 standard type fuel elements (Uranium Silicon Aluminum Dispersion Type Alloy) in a package. The maximum K at a 99%
confidence level of this result K eff +3 is 0.917, which is less than the standard value of 0.95.
For KUCA coupon fuel, when 1200 KUCA coupon fuels in a package (120 coupons are inserted into one grid of rectangular pipe) and the coupon gaps were 0.4 cm, and the fuel in each grid is at the center of the basket, the maximum K +/- is 0.8080+/-0.0026. The maximum K at a 99%
confidence level of this result K eff +3 is 0.8158, which is less than the standard value of 0.95.
For KUCA flat fuel, when 300 KUCA flat fuels in a package (30 flat plates are inserted into one grid of rectangular pipe) and the fuel plates gaps are evenly spread most, and the fuel in each grid is at the center of each grid, the maximum K +/- is 0.9055+/-0.003. The maximum K at a 99%
confidence level of this result K eff +3 is 0.9145, which is less than the standard value of 0.95.
The spectrum converter was analyzed when one spectrum converter was installed on the top of the fuel basket as described above. The effective multiplication factor is keff +/- = 0.3935 +/- 0.0022, and the effective multiplication factor in the 99% confidence interval at this time is keff
+ 3 = 0.4001, which is well below the reference value 0.95 and is subcritical.
The effect of optimum moderation by water is considered by varying the 31
density of water within and surrounding the inner shell from 1.0g/cm3 to 0.0g/cm3. The calculations are performed for the JRR-3 standard type fuel element (Uranium Silicon Aluminum Dispersion Type Alloy) which shows the highest effective multiplication factor of the 11 types of fuel elements at Max density of water 1.0 g/cm3. The results show that the optimum moderation occurs at a water density of 0.02 g/cm3, and it is subcritical (keff+3=O.939).
For KUCA fuel, the effect of optimum moderation by water is considered by varying the density of water within and surrounding the inner shell from 1.0g/cm3 to 0.0g/cm3. The calculations are performed for the flat fuel which shows the highest effective multiplication factor at Max density of water 1.0 g/cm3. The results show that the optimum moderation occurs at a water density of 0.001 g/cm3, and it is subcritical (keff+3=O.9325).
As for JMTRC fuel, there is a case in which two kinds of fuel of different enrichment (MEU, HEU fuels) are mixed in the package and transported.
In addition, as a result of evaluating the effect of the spectrum converter on the effective multiplication factor when the density of water filling the inside of the inner container and between the package changes from 1.00 g/cm3 to 0.00 g/cm3, the highest effective multiplication factor was obtained. It increases when the water density is 1.00 g/cm3, but even then, keff + 3 = 0.4001, and this package is subcritical.
235 In this case, the quantity of U loaded is less than the case where the MEU fuels are loaded, and the effective multiplication factor becomes smaller than the case of MEU fuels loading.
32
()-Table E.8 Results of criticality analysis when immersed Number Enrichm Mass 235 *1 of Meat ent of U Fuel Element Fuels*2 Keff +/- Keff +/-3 Material of 235U*1 (g/elemen (Unit/pac (wt%) t) kage)
Uranium-Silicon 19.95 485 10 0.902+/-0.005 0.917 JRR-3
-Aluminum Standard Type 0.939*3 dispersion Alloy JRR-4 Uranium-Aluminum 93.3 182 10 0.811+/-0.006 0.829 B Type Alloy JRR-4 Uranium-Aluminum 19.95 245.3 10 0.801+/-0.007 0.822 L Type dispersion Alloy Uranium-Silicon 19.95 210 10 0.799+/-0.004 0.811 JRR-4 -Aluminum dispersion Alloy JMTR 46.0 320 10 0.827+/-0.006 0.845 Uranium-Aluminum Standard Type dispersion Alloy (MEU)
JMTR Uranium-Silicon 19.95 425 10 0.893+/-0.004 0.905 Standard Type -Aluminum (LEU) dispersion Alloy JMTRC Uranium-Aluminum 90.0 285 10 0.783+/-0.004 0.796 Standard Type Alloy (HEU)
JMTRC 46.0 317 10 0.812+/-0.004 0.825 Uranium-Aluminum Standard Type dispersion Alloy (MEU)
Uranium-Aluminum 90.0 285 5 0.796+/-0.004 0.809 JMTRC Alloy Standard Type 46.0 317 5 Uranium- Aluminum (HEU,MEU) dispersion Alloy Uranium-Silicon 19.95 218 10 0.771+/-0.001 0.774 KUR
-Aluminum Standard Type dispersion Alloy KUCA Uranium-Molybdenum 19.95 4 1200 0.8080+/- 0.81584 Coupon fuel -Aluminum (120/ 0.0026 dispersion Alloy grid)
KUCA 19.95 15 300 0.9055+/- 0.9145 Uranium-Silicon Flat fuel -Aluminum (30/ 0.003 dispersion Alloy grid 0.93255 Spectrum Uranium dioxide 90 1002 1 0.3935+/- 0.4001 converter (Total) 0.0022
- 1 : The value utilized in calculation
- 2 : Number of fuel elements loaded in a package
- 3 : Water density 0.02g/cm3
- 4 : Fuels was slide at the center of basket
- 5 : Water density 0.001g/cm3 which is outside of basket 33
E.5 Benchmark test (1) Benchmark test To verify the validity of the criticality analysis method by using a combination of the KENO-a code and the 137 energy group MGCL Library in SCALE which is used in this chapter, the analysis is conducted for the following experiments, and the result is evaluated.
(a) The criticality test (TCA criticality test)(3) conducted in National institute of Japan Atomic Energy Agency (JAEA), in which the lowly enriched UO 2 fuel rods clad by Aluminum are arrayed.
(b) The criticality test (International benchmark test)(4) conducted in ORNL 235 using the SPERT-D fuel (Uranium Aluminum alloy, 93.17% U enrichment)
(c) The criticality test(5) conducted for JRR-4 (20% enrichment, U 3 Si 2 , plate type fuel)
(2) Description of benchmark experiment (a) TCA criticality test The benchmark experiment was performed at Tank-type Critical Assembly (TCA) of JAEA. The critical water heights were measured by the experiment.
The experiment was performed varying fuel type, rod lattice pattern, lattice pitch and fixed poisons. The fuel material is uranium or uranium-plutonium oxide.
The experimental configuration of TCA facility and the dimension of uranium oxide rod are shown in ()-Fig.E.16.
The fuel rods are arrayed on a square pitch in the tank and four kind of lattice pitch, which correspond to the water-to-fuel volume ratio are 1.50, 1.83, 2.48 and 3.00. The number of fuel rods in a tank is changed according to the lattice pitch.
The calculations are performed for five cases of above experiment with 235 low enriched (2.6% U) uranium oxide fuel.
34
(b) International benchmark test OECD/NEA planned ICSBEP (International Criticality Safety Benchmark Evaluation Project in 1994 to verify the criticality safety analysis code, and produced the International Handbook of Evaluation Criticality Safety Benchmark Experiments. In this handbook, the criticality test conducted in ORNL (23 tests) to determine the specification of fuel storage, transport and reprocessing by using SPERT-D fuel (Uranium aluminum alloy, 235 93.17% U enrichment, shown in ()-Fig.E.17, ()-Fig.E.18) is described.
The three cases of criticality data, which are close to the JRR-4, are selected from the above test data as the international benchmark test data, are analyzed by using MGCL library and KENO- a code. The above three cases are described as follows.
() CASE3 (SPART3)
Shape of lattice :4x3.09 No. of criticality fuel :12.36+/-0.17 Criticality mass (235U) :3.79+/-0.05kg Lattice array :Refer to ()-Fig.E.19 (The figure shows the No. of the fuel plate)
() CASE15 (SPART15)
Shape of lattice :16x3 No. of criticality fuel :48 Criticality mass (235U) :19.62kg Lattice array :Refer to ()-Fig.E.19 (The figure shows the No. of the fuel plates) 35
() CASE23 (SPART23)
Shape of lattice :6x5.55 No. of criticality fuel :33.12+/-0.10 Critical mass (235U) :10.15+/-0.03kg 235 U enrichment :3.99g/
Boron enrichment :0.871g/
Lattice array :()-Fig.E.19 (The figure shows the No. of fuel plates)
(c) JRR-4 critical test JRR-4 is a swimming pool type research reactor of maximum 3.5MW output, and the fuel is lowly enriched uranium silicon aluminum dispersion type fuel.
The fuel elements are arrayed in the 4x5 lattice, and the graphite reflector (Lid tank side, the large reflector is made of Aluminum),
irradiation shell and the neutron source are arranged outside the fuels.
The plate shape 5 control rods and back up safety control rod are located between the fuel elements and the reflector. The moderator and the coolant are light water.
The fuel elements and the core arrangement are shown in ()-Fig.E.20 and ()-Fig.E.21 respectively. The minimum core and total core criticality tests are conducted in July in 1998.
As for the minimum core, the 12 fuel elements are arranged on the cross lines, and the graphite reflector is located outside the fuel elements, and the control rods of C 1 , C 2 and C 3 are being withdrawn by full stroke, and the C 4 control rod and the C 5 control rod are being withdrawn by 369mm and 292mm respectively. The core temperature during the experiment is approximately 20.
The criticality analysis for these minimum core criticality and for the maximum core criticality are conducted by combining the MGCL library and KENO- a code.
36
(3) The result of the benchmark test In order to verify the accuracy of the criticality analysis by combining the MGCL library and the KENO-.a code used in this analysis, the effective multiplication factors by using MGCL and KENO-.a are obtained for the following conditions, and the result is shown in ()-Table E.9.
(a) The criticality experiment (TCA criticality experiment) in which the lowly enriched UO 2 fuel rod with the Aluminum clad, conducted in JAEA.
(b) The criticality experiment (International benchmark experiment) conducted in ORNL using SPERT-D fuel (Uranium Aluminum alloy, 235U enrichment of 93.17%)
(c) The maximum and minimum core criticality experiment conducted in JRR-4 (20% enrichment, U 3 Si 2 , plate fuel)
From these results, the analytical procedure and the nuclear data is judged to bring the valid result.
()-Table E.9 Analysis result of benchmark criticality test Test name Fuel rod (Plate)(Element) array Keff 1 Keff+3 17x171.83 0.9926 0.0042 1.0052 21x211.83 0.9911 0.0043 1.0040 TCA criticality 20x201.50 0.9883 0.0040 1.0003 experiment 18x182.48 0.9859 0.0041 0.9982 17x173.00 0.9981 0.0041 1.0104 (88x681x2) 0.98896 0.00174 0.99418 International (352x88) 0.98865 0.00141 0.99288 benchmark test (132x110111211121211) 0.99110 0.00138 0.99524 JRR-4 (2x44) 0.98901 0.00138 0.99315 criticality test (4x5) 0.98319 0.00116 0.98667
- : Volumetric ratio of fuel and water 37
()-Fig.E.16 Configuration of TCA criticality experiments 38
39
()-Fig.E.17 SPERT-D fuel
()-Fig.E.18 SPERT-D fuel (continued) 40
41
()-Fig.E.19 Core arrangement
()-Fig.E.20 Fuel element 42
()-Fig.E.21 Core arrangement 43
E.6 Summary of results and evaluation If it is assumed that the article is under the general test conditions for fissionable transported articles, the deformation of the shipping casket is the deformation of the outer shell, which is outside a system subject to criticality assessment (surface of the transported article with the state of damage considered).
No dent containing a cube measuring 10 cm on a side would occur in the inner shell that is a system subject to criticality assessment, and each side of a circumscribed rectangular solid would not be below 10 cm.
The maximum effective multiplication factor was obtained when one package contained ten JRR-3 standard type fuel elements (Uranium Silicon Aluminum Dispersion Type Alloy) as shown in ()-Table E.7.
K +3=O.939 (at a water density of 0.02 g/cm3) and the packaging is in subcriticality.
44
E.7 Appendix E.7.1 Evaluation of optimum moderating water density E.7.2 Description of KENO-V a code and KENO-VI code E.7.3 Explanation of MGCL neutron cross section library and MAIL code E.7.5 References 45
E.7.1 Evaluation of optimum moderating water density The effect of water density change to the subcriticality of the package is evaluated under the condition of water immersion in the package.
The water density at optimum moderation depends on the distance and the neutron absorbing materials between fuel elements. In case of this package, there is no considerable difference in the pitch of steel pipe enveloping a fuel element.
Therefore, the evaluation of multiplication factor under the optimum moderation is performed for the case where the most reactive fuel element in the water of 1.0g/cm3 is loaded to the package.
As the JRR-3 standard type fuel element (Uranium Silicon Aluminum Dispersion Type Alloy) is the most reactive in the water of 1.0g/cm3, the critical calculation is performed for JRR-3 standard type fuel elements (Uranium Silicon Aluminum Dispersion Type Alloy) by varying the water density from 0.0 to 1.0g/cm3. In addition, KUCA flat fuel case was investigated by varying the water density from 0.0 to 1.0g/cm3. The calculation model and material compositions except water composition is same as the water density of 1.0g/cm3.
()-Table E.10 and ()-Fig.E.22 show the calculated multiplication factors for various water density. For JRR-3 standard type fuel, the optimum moderation is observed at the condition that the water density is about 0.02g/cm3. The calculated multiplication factor at the optimum moderation is 0.939 in 99% confidence level (keff + 3), lower than reference value of 0.95. For KUCA flat fuel case, the optimum moderation is observed at the condition that the water density is about 0.001g/cm3. The calculated multiplication factor at the optimum moderation is 0.9325 in 99%
confidence level (keff + 3), lower than reference value of 0.95.
This result indicates that the package is maintained subcritical at any water density.
46
()-Table E.10 Effective multiplication factor for various water density
[contained ten JRR-3 standard type fuel elements (uranium silicon Aluminum dispersion type alloy)]
Water Density Keff Keff+3 (g/cm3) 1.00 0.9021 0.0047 0.9162 0.60 0.8391 0.0052 0.8547 0.40 0.8189 0.0041 0.8312 0.20 0.8572 0.0040 0.8692 0.10 0.9028 0.0034 0.9130 0.05 0.9286 0.0026 0.9364 0.02 0.9305 0.0026 0.9383 0.01 0.9294 0.0019 0.9351 0.00 0.9067 0.0017 0.9118 (300 KUCA flat plates in a package)
Water Density Keff Keff+3 (g/cm3) 1.00 0.9055 0.003 0.9145 0.5 0.8321 0.0022 0.8387 0.1 0.9208 0.0019 0.9265 0.001 0.9295 0.0010 0.9325 0.00 0.9190 0.0009 0.9216 47
48
()-Fig.E.22 Relationship between effective multiplication factor (keff+/-3) and water density (contained ten JRR-3 standard type fuel elements (uranium silicon Aluminum dispersion type alloy))
E.7.2 Description of KENO-V. a code and KENO=VI code (1) KENO-V. a code KENO-V. a, developed by the U.S. ORNL, is a Monte-Carlo criticality calculation code. Based on the multigroup Monte-Carlo method, the KENO code is capable of calculating neutron multiplication factors for complicated systems.
As the library for neutrons cross section, the KENO code uses a library with neutron scattering matrix expressed by Legendre's extended terms (P L )
in multigroup form.
The KENO-IV, the version preceding the KENO-V. a, is only capable of handling primary degrees (P 1 ) for extension of scattering matrix, while the latest KENO-V. a is capable of handling any degrees. (However, the application only covers primary degrees.) The KENO-V. a has increased accuracy especially in systems where the anisotropy of neutrons' scattering has a great influence on their effective multiplication factor.
The KENO-V. a uses the same basic calculation method for effective multiplication factor as the KENO-IV. This method is based on the assumption that fissile neutrons generated in a field containing fissile material lose their weight in the course of collision with the medium according to their absorption cross section in the medium.
Neutrons will be traced until their weight falls lower than a specified value or until some of the neutrons begin to leak from the system. In the collision in a medium containing fissile material, the weight of fission is recorded and used for the distribution of neutron generations in the next generation. Generating neutrons (usually 300 neutrons) for one generation and repeating the generation of neutrons according to the weight distribution of fission in the preceding generation will bring about a distribution similar to that of actual fissile neutron generations. The effective multiplication factor of the system is the mean of the effective multiplication factors of the different generations.
49
NPB Ncou f Wt ij j =1 i =1 t Keff = NPB Wtoj j =1 where NPB : Number of neutrons generated in one generation NCOLL: Number of collisions of neutrons Wt : Weight of neutrons at the time of fission Wt : Weight of generated neutrons
- Number of neutron generations per fission
- Macro fissile cross section
- Total macro cross section
- Number of collisions of neutrons
- Number of neutron generated in one generation (2) KENO-VI code KENO-VI is the latest version of the Monte-Carlo criticality calculation code KENO, and is incorporated as Monte-Carlo criticality calculation module in SCALE code system. The calculation procedures are similar to those of KENO-V.a, whereas KENO-VI can handle more complicated geometry.
E.7.3 Explanation of MGCL neutron cross section library and MAIL code MGCL is the multi-group neutron cross section library generated at JAEA by processing ENDF/B-IV(1) evaluated neutron cross section with SUPERTOG, PIXSES and other cross section processing codes. The energy group structure of MGCL master library is 137 groups.
MGCL master library includes the infinite diluted cross sections, resonance self shielding factors and scattering matrix for sixty seven nuclides. The scattering matrix is represented by p 1 approximation.
MAIL is the computer code to generate macroscopic effective cross section from MGCL in the form used by KENO-IV and ANISN. The heterogeneous effect of resonance self shielding is corrected with Dancoff-Ginsberg factor.
50
E.7.4 References (1) Y. Naito, et al. MGCL-PROCESSUR: A Computer Code System for Processing Multi Group Constant Library MGCL, JAERI-M9396 (1981).
(2) L. M. Petrie. et al. KENO-V a: A Monte Carlo Criticality Program with Super Grouping, NUREG/CR-2O0 rev. 3 sec. F-11 (1984).
(3) SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, ORNL/TM-2005/39, Version 5.1, Vols.
I-III, November 2006. Available from Radiation Safety Information Computational Center at Oak Ridge National Laboratory as CCC-732.
(4) Y. Komuro, et al KENO-IV Code Benchmark Calculation (10) (Critical Experiment of Light Water Type Critical Assembly), JAERI-M9147(1980)
(in Japanese).
(5) K. Woods, et al. Critical Experiments of SPERT-D Fuel in Water, NEA/NSC/DOC(95)03/ Volume (1998).
(6) Y. Nakano, et al. Neutronics Characteristics of JRR-4 Low Enriched Uranium Core, Proceedings of 21th International of RERTER (1998).
51
()-F Consideration of aging of nuclear fuel package
()-F. Consideration of aging of nuclear fuel package F.1 Aging factors to consider In this chapter, it is considered that the usage conditions expected during the planned use period of this package and the accompanying secular change Possible causes of aging during the period of use of this package are temperature changes during container storage and use, radiation generated from the stored materials, chemical changes such as corrosion and repeated load.
Therefore, we will evaluate the main materials and elements of this shipping container for the factors that cause these changes over time.
The evaluation will be carried out with the period of use of this package being 40 years after manufacture, the number of times of use being three times a year, and the number of days required for transportation being 100 days.
F1
F.2 Evaluation of the need to consider aging effect in safety analysis The constituent materials of this package are as shown in (II) -Table A.5 of (II)-A. Among these materials, the materials that take aging into consideration are shown below.
Stainless steel Heat insulator (hard polyurethane foam)
Shock absorber (balsa wood)
The aging change of O-ring is not considered in this chapter because it is exchanged for each transport. In addition, only the transport container should be considered since the stored items will be exchanged for each transport The consideration of aging of each material is shown in (II) -Table F.1.
As shown in A.4.4 of (II)-A, the lifting device is no effect of fatigue due to repeated loading during the period of use because allowable number of repetitions is larger than the expected number of repetitions (12000 times) during the planned use period when a lifting load is applied. As shown in A.5.1.3 of (II)-A, the sealing device is no effect of fatigue due to repeated loading during the period of use because the allowable number of repetitions when a design pressure is applied is larger than the expected number of repetitions (500 times) during the period of use.
F2
IITable F.1 Evaluation of requirements for consideration of aging in safety analysis (1/3)
Material Factor Examination of requirement for consideration Evaluation of aging Stainless steel Heat Regarding temperature changes in stainless The maximum temperature during transport is SUS304SUS630 steel, the metal cask structure standard 65, and this material is not affected by aging (design and construction standard (JSME S due to heat since the function of this material NSI-2005)(1) (2007 supplement version)(2) does not deteriorate.
stipulates design strength and physical property values up to 425, stainless steel is not affected by creep or so up to the relevant temperature.
Radiation When the neutron irradiation dose is 1016 n/cm2 Even if neutron irradiation is concentrated on or more, the material strength begins to be a specific 1 cm2 of a structural part under the affected (3). condition that it is used for 40 years, with the three times per year and 100 days for one transport, the neutron irradiation amount is F3
3.69 x 1011 n/cm2. Therefore, this material is not affected by aging due to neutron irradiation.
Chemical Stainless steel is a material that forms a Stainless steel is a material that does not passivation film on the surface layer and is not easily corrode, and has no practical effect on easily corroded. In addition, no significant corrosion due to the adhesion of sea salt corrosion was observed in the exposure test of particles that is expected during transport.
sea salt particles (4). Furthermore, even if corrosion should occur, the presence or absence of corrosion will be confirmed by periodic inspections and maintenance work, and appropriate measures will be taken.
From the above, this material is not affected by aging.
F4
IITable F.1 Evaluation of requirements for consideration of aging in safety analysis (2/3)
Material Factor Examination of requirement for consideration Evaluation of aging Heat insulator Heat It has been confirmed that the amount of change The maximum temperature during transport is (hard polyurethane in weight of polyurethane foam sharply due to 65, and this material is not affected by aging foam) temperature changes when the temperature due to heat since the function of this material (5) exceeds 200 . does not deteriorate.
Radiation No significant changes in mechanical Even if neutron irradiation is concentrated on properties were observed up to a neutron dose a specific 1 cm2 of a structural part under the of 1015 n/cm2 (3). condition that it is used for 40 years, with the three times per year and 100 days for one transport, the neutron irradiation amount is 3.69 x 1011 n/cm2. Therefore, this material is not affected by aging due to neutron irradiation.
Chemical This material may absorb moisture due to Since this material is covered with stainless F5
contact with moisture. steel and is in a closed space, it is not affected by decomposition or corrosion due to moisture absorption. As for the evaluation of stainless steel, as mentioned above, since it is not affected by aging, there is no risk of moisture absorption and corrosion of this material, and it is not affected by aging.
F6
IITable F.1 Evaluation of requirements for consideration of aging in safety analysis (3/3)
Material Factor Examination of requirement for consideration Evaluation of aging Shock absorber Heat It has been confirmed that the temperature The maximum temperature during transport is (balsa wood) change of this material is affected from 115, 65, and this material is not affected by aging but the effect of temperature change is not due to heat since the function of this material (7) confirmed below that temperature . does not deteriorate.
Radiation No significant changes in mechanical Since the neutron irradiation amount from the properties were observed up to a neutron dose neutrons generated from the stored items in of 3 MGy (7). this transport container is about 142 Gy even if it is conservatively estimated, it is not necessary to consider the aging effect due to neutron irradiation.
Chemical This material may absorb moisture due to Since this material is covered with stainless contact with moisture. steel and is in a closed space, it is not affected by decomposition or corrosion due to F7
moisture absorption. As for the evaluation of stainless steel, as mentioned above, since it is not affected by aging, there is no risk of moisture absorption and corrosion of this material, and it is not affected by aging.
F8
F.3 Consideration of aging in safety analysis As shown in the previous chapter, the necessity of considering aging for each material related to this package was evaluated. As a result, it was found that it is not necessary to consider the aging in the safety analysis for this package.
Elements not evaluated in this chapter do not need to be considered in safety analysis because their soundness is guaranteed in the maintenance and handling of the transport container shown in chapter (IV).
References (1) Japan Society of Mechanical Engineers, "Spent Fuel Storage Facility Standard Metal Cask Structural Standard (JSME S FA1-2007)" (2007).
(2) Japan Society of Mechanical Engineers, "Nuclear Power Plant Equipment Standards for Design and Construction Standards (2005 Edition) <Volume I Light Water Reactor Standards> (JSME S NC1-2005 (including 2007 supplementary edition))" (2007).
(3) K.Farrell, et al,An Evaluation of Low Temperature Radiation Embrittlement Mechanisms in Ferritic Alloys, J. of Nuclear Materials, Vol.210,(1994).
(4) R.R.Gaugh, Stress Corrosion Cracking of Prescipitation-Hardening Stainless Steels, Materials Performance, Vol.26, No.2, (1987).
(5) Hokkaido Research Organization, Northern Construction Research Institute, "Construction of Material Index for Foamed Plastic Insulation that Contributes to Fire Protection Performance", Survey Research Report No.352 (2015).
(6) Japan Nuclear Energy Safety Organization, "2003 Metal Cask Storage Technology Confirmation Test Report Final Report", (2004).
(7) Gilbert Gedeon, P.E., Wood as An Engineering Materials: Mechanical Properties of Wood Course No:S04-019 F9
()-G Assessment of the compliance with the regulation and the notification G
(II)-G. Assessment of the compliance with the regulation and the notification This transported article is in conformity to the relevant items of technical standards stipulated in the regulation and the notification as shown in (II) Table G.1.
(II) - G - 1
(II) Table G.1: Assessment of the compliance with the technical standards stipulated in the regulation and the notification Item Item of the Item of the corresponding to Explanation Remarks regulation notification description in the application form Article 3-1-1 Article 3 Not applicable since this transported article is a BU-type transported article.
Article 3-1-2 Article 4 Not applicable since this transported article is a BU-type transported article.
Article 3-1-3 Article 4 The nuclear fuel materials contained in this (I)-B and transported article correspond to those other Appended than special-form nuclear fuel materials, table 1 and are uranium alloy and uranium oxide with enrichment of the fuel material being below 93.3 wt%.
Since the amount of radioactivity contained in the cask exceeds the A2 value, this transported article corresponds to a Article 3-2 Article 5 Not applicable since this transported article is a BU-type transported article.
Since this transported article is a BU-type (I)-D Article 3-3 transported article, it is subject to the technical standard stipulated in Article 7 of the regulation.
In addition, as described below, the evaluation was made as follows because it is required to consider the aging in the conformity evaluation of the technical standards of Article 7.
- 1. The planned number of years of use of the package is set to 40 years, the number of times of transportation per year is set to 3 times, and the number of days required for transportation per transportation is set to 100 days.
- 2. As for the stored items, since there is no aging effect in the relevant transportation days, the transportation container components will be considered.
(II) - G - 2
- 3. The factors of aging are heat, irradiation, chemical change and repeated loading.
- 4. Regarding the effect of radiation irradiation, even if it is irradiated from the stored items during the planned use period, it is well below 1016 n/cm2 where the material strength is affected.
- 5. Regarding the influence of heat, the maximum temperature of transport container components is 65 ° C, stainless steel is within the range where structural strength standards are set by standards, and regarding heat insulating materials and cushioning materials, the temperature which mechanical properties change occurs is below the temperature.
- 6. Regarding the effects of chemical changes, stainless steel forms a passivation film on the surface and corrosion does not proceed, and the condition should be confirmed by visual inspection and repaired if necessary, and the heat insulating material and cushioning material should be covered with stainless steel and they are in a closed space, so they do not decompose or corrode due to moisture absorption, etc.
- 7. Regarding the effect of repeated load, the allowable number of repetitions based on the stress generated by the repeated load due to pressure and handling exceeds the expected number of repetitions.
From the above, the package is not affected by aging.
Article 4 Not applicable since this transported article is a BU-type transported article.
Article 5 Not applicable since this transported article is a BU-type transported article.
Article 6 Not applicable since this transported article is a BU-type transported article.
(II) - G - 3
Article 7-1 The package can be handled easily and (II)-A.4.4 Article 4-1 safely as shown below.
The packaging has an eye plate in the container body so that it can be easily lifted and lowered during handling, and the eye plate can be connected to the hanger and easily handled. In addition, the eye plate has a safety factor of 3 in consideration of handling, and is designed to have the necessary strength against the load considering the maximum weight of the transported object, so that it can be handled safely.
Article 7-1 As shown below, the package is not likely (II)-A.4.7 Article 4-2 (continued) to crack or break due to the expected (II)-A.5 temperature, internal pressure, vibration, etc. during transportation.
- 1. The minimum temperature of each part of the package expected during transportation is -40 ° C, and the maximum temperature is 38 ° C as stipulated in Article 7 (2) of the Regulation. In this case, the decay heat of the stored items is negligible, so the temperature of each part of the transported items is uniformly 65 ° C. At this temperature, the transport container components do not compromise their integrity.
- 2. Significant thermal expansion differences and thermal stresses of package does not occur between transport, even assuming that the ambient temperature of the package changes from -40 ° C to 38 ° C with respect to expected temperature changes during transport.
- 3. Regarding the expected change in internal pressure during transportation, the maximum internal pressure of the package is 0.147MPa (absolute pressure). The strength and sealability of the sealing device are evaluated under the condition that the pressure exceeding these is applied to the inner container, and it is confirmed that the structural soundness and the sealing property are ensured.
(II) - G - 4
Item Item of the Item of the corresponding to Explanation Remarks regulation notification description in the application form
- 4. As a result of calculating the natural frequency of the package for vibration, etc., there is a large difference with respect to the frequency (0 to 50 Hz) expected to be input to the package during transportation, and the load received during transportation is not amplified. In addition, since the input load expected during transportation is included in the load during free fall and stacking test under general test conditions, there is no risk of cracking or breakage of the transported material.
Article 7-1 There are no protrusions on the surface (I)-C Article 4-3 (continued) of this package other than the eye plate used for handling. In addition, the surface of this package is made of smooth-finished stainless steel, which makes it easy to remove contamination.
Article 7-1 The components of this package are (II)-A.4.1 Article 4-4 (continued) made of chemically stable materials such (II)-A.4.2 as stainless steel, hard polyurethane foam, and balsa wood. There is no risk of dangerous physical or chemical action with the materials that make up the container and with the stored items.
- 1. As shown in Article 4-2, there is no significant difference in thermal expansion within the expected temperature range during transportation, and there is no mutual interference due to thermal expansion. There is no risk of dangerous physical action between them.
- 2. This package is a type that does not use cooling water, and there is no risk of damage due to freezing.
- 3. Insulation material (hard polyurethane foam), cushioning material (balsa wood), and silicone rubber O-rings do not cause a chemical reaction even if they come into contact with metal materials.
- 4. Since the hard polyurethane foam and balsa wood are covered with stainless steel and sealed, there is no risk of corrosion.
(II) - G - 5
Article 7-1 Due to the valveless design of this (II)-C.2.1 Article 4-5 (continued) packaging, no technical standards apply.
(II)-A.4.3 Article 7-1 Article 9 It shall be confirmed that the density of the (III)-A.2 Article 4-8 (continued) radioactive material on the surface of this transported article does not exceed the following value in a pre-shipment inspection.
- 1. Radioactive material emitting alpha ray:
0.4 Bq/cm2
- 2. Radioactive material not emitting alpha ray: 4 Bq/cm2 Article 7-1 The loading of fuels in the shipping cask (IV)-A.2 Article 4-10 (continued) is performed in accordance with prescribed procedures. Further, a content inspection is conducted as the pre-shipment inspection of the transported article. Therefore, no material that may impair the safety of the transported article will be loaded.
In this transported article, each side of the circumscribed cube is 10 cm or more as indicated below. Article 5-2 (I)-C JRF-90Y-950K type (I) Fig. C.1 Height: approx. 1,800 mm Outer diameter: approx. 840 mm Although the opening/closing section of the Article 5-3 (II)-A.4.3 transported article is the lid of the inner shell, the lid is covered with the lid of the outer shell. Therefore, it will not be carelessly opened. In addition, the lid of the outer shell is locked and sealed.
Article 7-1 As shown in Article 4-2, the expected (II)-A.3 Article 5-4 (continued) temperature range of the container (II)-B.4.2 components during transportation is in the range of -40 ° C to 65 ° C. On the other hand, in the temperature range of -40 ° C to 75 ° C including the above, the material of the transported component does not cause a significant decrease in strength or embrittlement, and does not affect the required material strength. Therefore, there is no risk of cracking or breakage of the components in the temperature range of -
40 ° C to 75 ° C.
(II) - G - 6
Item Item of the Item of the corresponding to Explanation Remarks regulation notification description in the application form Article 7-1 When the external pressure drops to 60 (II)-A.4.6 Article 5-5 (continued) kPa, the internal/external pressure difference is 0.056 MPa. On the other hand, there is no leakage of radioactive materials even when the external pressure drops to 60kPa because it was confirmed that the structural soundness and sealing performance of the sealing device are ensured when the internal pressure of the inner container is 0.0981MPa, which is the design pressure, and the external pressure has dropped to 60kPa.
Article 7-1 This package is not applied to this Article 5-6 (continued) requirement because it does not contain liquid nuclear fuel material.
Article 7-1 For this package, the following (II)-D.5 Article 5-7 (continued) conservative conditions are set, and the maximum dose equivalent rate on the surface of the package is obtained using the ANISN code. In addition, the analysis evaluated under the condition that new fuel with a larger amount of uranium than the stored items and low-irradiated fuel were stored, and the maximum dose equivalent rate on the surface of the package was 0.169 mSv/h when low-irradiated fuel is stored, and it is lower than 2 mSv/h.
- 1. For 234U and 236U, which have a high contribution to the dose equivalent rate, the amount exceeds the maximum value of the actual results so far.
- 2. It is assumed that the fuel with low irradiation has a larger amount of uranium and higher radioactivity than the actually stored fuel.
- 3. The outer surface of the inner container is regarded as the surface of the transport container, ignoring the structural materials such as the outer container.
(II) - G - 7
Item Item of the Item of the corresponding to Explanation Remarks regulation notification description in the application form Article 7-1 This package evaluates the dose (continued) (II)-D.5 Article 5-8 equivalent rate at a position 1 m away from the surface based on the same conservative conditions as the maximum dose equivalent rate on the surface. The maximum dose equivalent rate at a position 1 m away from the surface is 19 Sv/h, which is below 100 Sv/h.
Article 7-1 The amount of radioactivity contained in (II)-A.6.4 Article 6-5 (continued) this package is 29.8 GBq. Even if it is assumed that all of this radioactivity is 234U (A2 value: 6GBq), which has the minimum A2 value, this requirement is not applied because it is less than 100,000 times.
Article 7-2 Article 19 General test conditions for BU-type Appendix 7 transported articles Appendix a. Thermal testing 4-1 (II)-B.4.1 As a result of evaluating the temperature of the package when the package is placed under a certain condition of solar heat radiation in an atmosphere of 38 ° C under general test conditions, the temperature of each part becomes uniform at 62 ° C.
Since the temperature is below 150 ° C.,
which is the maximum usable temperature of the silicon rubber O-ring, the soundness of the O-ring is maintained. In addition, the internal pressure of the inner container is 0.016 MPa (gauge pressure), but in the strength evaluation of the inner container, under the condition that the package loaded with 0.0981 MPa (gauge pressure), which exceeds the pressure, is uniformly 75 ° C.
Appendix It is confirmed that the inner container is 4-2 within the elastic range. The amount of opening of the inner container lid and flange that form the sealing boundary is less than the initial tightening allowance of the O-ring, so the sealing is maintained.
(II) - G - 8
Article 7-2
- b. Water spray (II)-A.5.2 Appendix Since the surface of the package is smooth 3-1-i stainless steel, it does not absorb water, and there is no risk of corrosion due to water absorption, there is no damage that affects Appendix the sealing performance and shielding 3-1-ii performance.
- c. Free fall (II)-A.5.3 Appendix Since the maximum weight of the package is 950 kg, the drop height is 1.2 m.
3-1-ii (1) Regarding the state of the package when it falls from a height of 1.2 m, the acceleration and deformation amount generated in the package are evaluated using the CASH-II code. In addition, the strength of the transport container and the stored items is evaluated based on the obtained acceleration. The fall posture is intended for vertical, horizontal, corner and tilted falls.
As a result of the analysis, the package is deformed up to 66 mm, but the deformation is limited to the outer container, and the inner container, inner container lid, basket and stored items are not deformed, and the amount of opening of the inner container lid is less than the initial tightening allowance of the O-ring. The structural soundness and sealing performance are ensured.
- d. Stacking (II)-A.5.4 Appendix 3-1-ii (3) Comparing the case where the projected area of the package is loaded with 0.13 kg /
cm2 and the case where the load is loaded with 5 times its own weight, the latter load is more severe, so it was evaluated under the latter condition. As a result, the package is not plastically deformed, the soundness of the transported material is ensured, and there is no damage that affects the sealing performance and the shielding performance.
Appendix e. Penetration (II)-A.5.5 Article 3-1-ii (4) 5-9-ii When a 6 kg steel rod of the outer plate (thickness 3 mm) of the outer container is dropped from a height of 1 m, the falling energy of the steel rod and the energy of penetrating the steel plate of the outer container are compared, and the latter value is larger. Since the outer steel plate does not penetrate, there is no damage that affects the sealing performance and shielding performance.
(II) - G - 9
Item Item of the Item of the corresponding to Explanation Remarks regulation notification description in the application form Article 7-2 Considering that the outer container will be (II)-D.4 Article (continued) deformed under the general test conditions, the maximum dose equivalent rate is 6-2-i evaluated under the outer surface of the inner container is regarded as the surface of the package, which is the same as the evaluation of the maximum dose equivalent rate during normal transportation. As a result of the evaluation, the maximum dose equivalent rate on the surface is 0.169 mSv/h, which is lower than 2 mSv h.
Article 7-2 The structural soundness and sealing (II)-C.3.1 Article (continued) property of the sealing device are ensured 6-2-ii for the package under the general test conditions, and the structural soundness of the stored material is also maintained. In the evaluation of the amount of radioactive material leaks, it is assumed that there is a leak equivalent to the pass criteria for the leak test in the pre-shipment inspection, and regarding the contents of the uranium isotopes that adhere to the surface of the fuel during fuel production, As a result of evaluating the amount of radioactive material leaked per hour, assuming that the internal pressure of the inner container exceeds the maximum internal pressure of 0.199 MPa (absolute pressure) under the condition that it is dispersed inside the vessel, the ratio to the standard (A2 x 10-6) is 7.89 x 10-4, which satisfies the standard.
Article 7-2 Article 15 The package is transported by a dedicated (continued) load. Even when the package placed under Article (II)-B.4.2 general test conditions is placed in the shade 6-2-iii in an environment of 38 ° C, the decay heat of the stored material is negligible, so the temperature of the surface of the transported material that a person can approach is 38 ° C. Therefore, the temperature of the surface of the package does not exceed 65 ° C.
Article 7-2 Article 9 The package which placed under general (I)-A.2 Article (continued) test condition does not release radioactive 6-2-iv substances because the structural soundness and sealing performance of the sealing device are ensured. In addition, since it is confirmed that the surface density is below the surface density limit in the pre-shipment inspection, the surface density limit is not exceeded.
(II) - G - 10
Article 7-3 Article 20 Special test conditions for BU-type (II)-A.6 Appendix 8 transported articles Appendix Drop test I (II)-A.6.1 5-1-i Regarding the state of the package when dropped from a height of 9 m, the acceleration and deformation amount generated in the package are evaluated using a CASH-II code. In addition, the strength of the transport container and the stored items is evaluated based on the obtained acceleration. The fall posture is intended for vertical, horizontal, corner and tilted falls.
As a result of the analysis, the transported material is deformed up to 146 mm, but the deformation is limited to the outer container, and the inner container, inner container lid, basket and stored items are not damaged, and the amount of opening of the inner container lid is less than the initial tightening allowance of the O-ring. The structural soundness and sealing performance are ensured.
Article 7-3 Appendix Drop test II (II)-A.6.2 (continued) Regarding the state of the package when 5-1-ii dropped from a height of 1 m onto a steel rod with a diameter of 150 mm, the amount of dent deformation of the outer container caused by the collision of the steel rod with the package is evaluated. The evaluation is based on the case where the steel rod collides with the outer container lid, the outer container body, and the outer container bottom. As a result of the evaluation, it was confirmed that none of the cases penetrated the outer plate of the outer container and that the steel rod did not reach the inner container and the inner container lid due to the deformation of the outer container.
Structural soundness and sealing performance are ensured.
(II) - G - 11
Item Item of the Item of the corresponding to Explanation Remarks regulation notification description in the application form Article 7-3 Appendix Fire resistance test (II)-A.6.3 (continued) 5-2 -i By using the TRUMP code, using a Appendix conservative model that superimposes the 5-2-ii deformation of the outer container by the drop test on the state of the transported material placed in the thermal test after performing the drop tests I and II. As a result of evaluating the temperature of each part of the transported material, the cushioning material and the heat insulating material are partially burnt. In addition, the temperature of the O-ring is 188 ° C, but it is below the short-term maximum usable temperature, so thermal soundness is ensured. In addition, the internal pressure of the inner container is 0.065MPa (gauge pressure) even when the internal air temperature is the maximum temperature of the basket, which is lower than 0.0981MPa (gauge pressure). The structural soundness and sealing performance of the sealing device are ensured.
Appendix Immersion test (water depth:15 m) (II)-A.6.4 Regarding the state of the transported 5-3 material placed at a water depth of 15 m, as a result of evaluating the structural strength and the opening amount of the inner container lid when an external pressure of 147 kPa is applied to the inner container and the inner container lid which are sealing devices, it was confirmed that the instrument lid was not damaged or buckled and that the opening amount of the inner container lid was less than the initial tightening allowance of the O-ring, structural soundness and sealing performance were ensured.
(II) - G - 12
Item Item of the Item of the corresponding to Explanation Remarks regulation notification description in the application form Article 7-3 For packages placed under special test (II)-D.5 Article 6-3-i (continued) conditions, the maximum dose equivalent rate is evaluated, which is the same is the same as the maximum dose equivalent rate evaluation during normal transportation, considering that the outer container is deformed and the cushioning material and heat insulating material are partially burned. As a result of evaluation using the ANISN code under the condition of the outer surface of inner vessel regarded as the package surface, the maximum dose equivalent rate at a position 1 m away from the surface is 0.019 mSv/h, which is lower than 2 mSv/h.
Article 7-3 Article 17 For package placed under special test (II)-C.4.2 Article (continued) conditions, the structural integrity and 6-3-ii sealing performance of the sealing device are ensured, and the structural integrity of the stored items is also maintained. In the evaluation of the amount of radioactive material leaked, it is assumed that there is a leak corresponding to the acceptance criteria of the leak test in the pre-shipment inspection, and regarding the contents of the uranium isotopes that adhere to the surface of the fuel during fuel production, As a result of evaluating the amount of radioactive material leaked per week, assuming that the internal pressure of the inner container exceeds the maximum internal pressure of 0.199 MPa (absolute pressure) under the condition that it is dispersed inside the vessel, the ratio to the standard (A2 x 10-6) is 9.32 x 10-8, which satisfies the standard.
(II) - G - 13
Article 7-4 As explained in the conformity with the (II)-A.3 technical standards of Article 7 No. 1 (II)-B.4.2 (Article 4 No. 2 and Article 5 No. 4), when the ambient temperature is in the range of -
40 ° C to 38 ° C, at the temperature of each part of the package, the material has no effect on the required structural strength. Further, structural soundness and hermeticity are ensured under the condition that the internal pressure of the inner container exceeds the maximum pressure in the temperature range.
Article 7-5 This package is a natural cooling (II)-B.1 system that does not have a cooling device and so on.
Article 7-6 The maximum working pressure of this (II)-B.4 package does not exceed 700 kPa because (II)-B.5 the difference in internal and external pressure (gauge pressure) is less than 0.0981 MPa, even considering the expected changes in temperature and internal and external pressure during transportation.
Article 8 Not applicable since this transported article is a BU-type transported article.
Article 9 Not applicable since this transported article is a BU-type transported article.
Article 10 Not applicable since this transported article is a BU-type transported article.
Article 11 Regarding the consideration of aging, as (II)-F explained in the conformity with the technical standards of Article 3, Paragraph 3, considering the radiation, heat, chemical change and repetitive load as the factors of aging, the planned number of years of use and the number of times, there is no effect of aging on the shipment.
Regarding the state of the package under the general test conditions for the fissile transport material, the outer container is partially deformed, but the structural soundness of the inner container, the inner container lid, and the basket is ensured.
(II) - G - 14
If it is decided to put it under special test conditions related to fissile transport materials, after receiving the history of general test conditions, regarding the condition of the package placed in the drop test, thermal test, and immersion test, the deformation of outer container and some of the cushioning and insulation will damaged, but the inner container, inner container lid, basket and stored items will not be damaged.
Article 11 Article 23 Since this transported article will contain 15 (I)-B g or more of uranium 235, and the (I)-D enrichment of uranium 235 will be 19.95 to 93.3%, it corresponds to the requirements for fissionable transported articles.
Article 11-1 Article 24 Appendix (General test conditions) 11-1-2 The effect of spraying water equivalent (II)-A.9.1 to a precipitation of 50 mm/h for one hour is assessed.
The maximum total weight of this (II)-A.9.1 transported article is approximately 950 kg, and the drop height is 1.2 m. An analysis is conducted so that the maximum damage caused by the drop can be assessed.
Article 11-1 Appendix Since applying a load equivalent to five (II)-A.9.1 (continued) 11-3 times the transported article in self-weight will represent a severer condition, the strength of the inner shell under this condition is assessed.
In this test, a mild steel bar with a weight (II)-A.9.1 of 6 kg and a diameter of 3.2 cm was dropped from a height of 1 m to the weakest part of this transported article.
(II) - G - 15
Item Item of the Item of the corresponding to Explanation Remarks regulation notification description in the application form Article 11-1 If the package is to be placed under (II)-A.9.1
-i, ii general test conditions, the deformation that occurs in the package is limited to a maximum of 66 mm of deformation of the outer container due to free fall. And the shape of the deformation is a deformation that causes a dent. Since there is no dent in the structure of the container so that it contains a cube with a side of 10 cm. On the other hand, since the outer diameter of the transported object is 840 mm and the total length is 1,800 mm, one side of the circumscribed rectangular parallelepiped is 10 cm or more.
Article 11-2 Article 25 As a conservative effective magnification (II)-E.3.1
-i to v analysis model that includes all five (II)-E.4.4 conditions from Article 11 No. 2 (i) to (v), in (II)-E.5 order to strengthen mutual interference between adjacent packages, the outer surface of the inner container is the surface of the package ignoring the outer container of the transport container, and the transport containers are infinitely arranged with the regular hexagon that circumscribes the outer surface of the inner container as the perfect reflection surface so that neutrons do not leak in the evaluation system, the stored items with a history of irradiation should not be irradiated, and the inside of the inner container is filled with water even in the array system, and the reactivity is maximized by distinguishing the water density inside and outside the basket where the stored items are stored As a result of calculating the effective magnification for all the stored items using the KENO-VI code, the maximum effective magnification is about 0.94, and the subcriticality is ensured regardless of which of the nuclear fuel materials is stored.
(II) - F - 16
Article 11-3 As explained in the conformity with the (II)-A.3 technical standards of Article 7 No. 1 (II)-A.4.2 (Article 4 No. 2 and Article 5 No. 4), when the ambient temperature is in the range of -
40 ° C to 38 ° C, the material has no effect on the required structural strength at the temperature of each part of the package.
Further, when the internal pressure of the inner container exceeds the maximum pressure in the temperature range, structural soundness and sealing performance were ensured.
(II) - F - 16
() Handling methods and maintenance of nuclear fuel package
()-A Package handling methods A.1 Method of loading The contents of this package are loaded in the following manner.
(1) Preparation of the contents Before being loaded, the contents shall pass a content inspection based on the pre-shipment content inspection indicated in ()-A.2.
(2) Loading of contents and installation of inner lid The packaging shall be transferred by means of handling tools to a location for loading and removal of the outer lid and inner lid. After this operation, the contents prepared in advance shall be loaded into a fuel basket and a top spacer shall be inserted.
After completion of the above operations, the inner lid shall be installed and the inner lid clamping bolt shall be fastened at a specified torque.
(3) Leak-tightness inspection on the inner lid Leak-tightness inspection on the inner lid shall be conducted.
(4) Installation of an outer lid The outer lid shall be fitted and fastened by a clamping bolt with a specified torque, and sealed and locked.
()1
A.2 Package inspection prior to shipment Pre-shipment inspection indicated in ()-Table A.1 is performed on each shipment of the package.
A.3 Method for removal The contents shall be removed from the package in the following procedure.
(1) Remove the outer lid and the inner lid.
(2) Remove the upper spacer.
(3) Remove the contents from the package.
(4) Install the inner lid and the outer lid.
A.4 Preparation of empty packaging After the contents are removed from the packaging, conduct radiation control of the inner surface of the packaging, and conduct decontamination as needed. In addition, conduct a visual appearance inspection of the packaging to confirm it has no anomaly, and then store it indoor.
()2
(IV) Table A.1: Procedures for pre-shipment inspection of the package Item of Method for inspection Acceptance criterion inspection Visual Visually inspect the appearance No cracking, abnormal flaw, deformation, etc. is appearance of the main body, inner lid and observed.
inspection outer lid.
Lifting With the package lifted, inspect The eye-plates have no cracking, abnormal flaw, inspection its appearance. deformation, etc.
Weight Measure the total weight of the The weight is not more than 950 kg.
inspection package.
Surface Measure the surface density of The surface density is not more than 0.4 Bq/cm2 for density the package by the smear method radioactive materials emitting alpha ray, or not inspection or the like. more than 4 Bq/cm2 for radioactive materials not emitting alpha ray.
Dose With fuel elements loaded, The sum of the dose equivalent rate for gamma ray equivalent measure the dose equivalent rate and neutron ray is not more than 2 mSv/h on the rate for gamma ray and neutron ray. surface of the package, or not more than 100 Sv/h inspection in a position 1 m distant from the package surface.
Subcriticalit Visually inspect the appearance 1. The fuel basket is installed in the y inspection of the fuel basket. prescribed position.
- 2. No cracking, abnormal flaw, deformation, etc.
is observed.
Content Inspect/measure the type, 1. Type inspection concentration, volume, It must be the design approval conditions.
appearance and surface density.
- 2. Concentration and volume It must be the design approval conditions.
- 3. Appearance: no anomaly is observed.
- 4. Surface density: not more than 0.056 Bq/cm2 for radioactive materials emitting alpha ray Airtight Apply air pressure of 0.392 MPa The leakage rate does not exceed 1.09 x 10-2 MPa leakage [gauge] to the sealed parts of cm3/s.
inspection the inner lid for 30 minutes, and measure the pressure drop to determine the leakage rate.
Pressure The decay heat generated from the contents is minimal, and the vessels measurement / temperature will remain the same as the ambient temperature. Therefore, this inspection inspection shall not be conducted.
Temperature The decay heat generated from the contents is minimal, the pressure in the package measurement / will remain constant, and therefore the pressure from inside the package will remain inspection the same as the ambient pressure. Therefore, this inspection shall not be conducted.
()3
()-B Maintenance requirement (IV)-B. Maintenance requirements The transport packaging shall be stored indoor. Periodical self-controlled inspections shall be conducted in accordance with the following instructions at least once every year (at least once every 10 times of use for those used 10 times or more yearly).
B.1 Visual appearance inspection Perform a visual inspection to confirm that there is no cracking, abnormal flaw, deformation, etc.
in the inner and outer surfaces of the main body, fuel basket, inner lid, and outer lid.
B.2 Pressure durability inspection If a repair or the like that may affect the pressure durability performance has been conducted, install a provisional inner lid and inspect the leakage rate for the main body of the inner shell by pressurized leakage testing (initial inspection pressure: 0.392 MPa [gauge] or more; inspection time: 30 minutes or more) to confirm that the leakage rate is not more than 1.09 x 10-2 MPacm3/s.
Subsequently, perform a visual inspection to confirm that there is no cracking, abnormal flaw, deformation, etc. in the inner surface of the main body of the inner shell.
B.3 Airtight leakage inspection Conduct airtight leakage inspection for the O-ring of the inner lid by pressurized leakage testing (inspection pressure: 0.392 MPa [gauge] or more; inspection time: 30 minutes or more) to confirm that the leakage rate is not more than 1.09 x 10-2 MPacm3/s.
B.4 Shielding inspection This does not apply since no particular shield is used in this transport packaging.
B.5 Subcriticality inspection Perform visual inspection to confirm that there is no anomaly in the dimensions, shape, etc. of the fuel basket, such as cracking, abnormal flaw, and deformation.
B.6 Thermal inspection This does not apply since this transport packaging has no particular exothermic body.
B.7 Lifting inspection With the transport packaging lifted, inspect the appearance of the transport packaging to visually confirm that the eye-plates have no cracking, abnormal flaw, deformation, etc.
B.8 Actuation check/inspection This does not apply since this transport packaging has no special articles such as valves.
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B.9 Maintenance of auxiliary systems This does not apply since this transport packaging has no auxiliary system.
B.10 Maintenance of the valves, gaskets, etc. of sealing devices This transport packaging has no valve or the like.
Inspect the O-ring of the inner lid to confirm that it has no cracking, abnormal flaw, deformation, etc. If any anomaly is observed, replace the O-ring.
B.11 Storage of the transport packaging The transport packaging shall be stored indoor.
B.12 Retention of records While this transport packaging is in service, retain a record of inspection conducted during fabrication and a record of periodical self-controlled inspection.
B.13 Others Not Applicable
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() Important Notice about a safe design and the safe transportation Not Applicable
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