ML22073A011

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1014-SR-00001, Revision 2, Safety Analyses Report, Type B(U)F Transport Package Castor geo69
ML22073A011
Person / Time
Site: 07109383
Issue date: 02/25/2022
From:
GNS Gesellschaft fur Nuklear-Service mbH
To:
Office of Nuclear Material Safety and Safeguards
References
T1213-CO-00014 1014-SR-00001, Rev 2
Download: ML22073A011 (69)
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Text

NS Safety Analyses Report Type B(U)F Transport Package CASTOR geo69 Docket No.: 71-9383 Non-Proprietary Version Document Type SR Document No. 1014-SR-00001 Revision 2 Name, Function Date Signature Prepared Reviewed Approved

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS 0.2 Revision Status of this Document Name, Function Date Signature Prepared 25.02.2022 Reviewed 25.02.2022 0.2 Revision Status of this Document Section 0.2, Rev. 2 Page 0.2-1

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS Revision Date Author Revised section 0 22.12.2020 T. Fischer-Wasels - First issue 1 30.08.2021 I. Eichmann Revised / new Sections:

- 0.2, 0.3, 0.4, 0.5

- 1.2

- 2.0, 2.1, 2.2, 2.12

- 3.1, 3.2, 3.4, 3.5, 3.6, 3. 7

- 5.1, 5.2, 5.3, 5.4

- 6.3

- 7.0, 7.1, 7.2

- 8.1, 8.2, 8.3 2 25.02.2021 I. Eichmann Revised / new Sections:

- 0.2, 0.3, 0.4

- 3.5

- 4.1, 4.2, 4.3, 4.5 Changes in relation to the previous revision are marked with a vertical bar on the left side 0.2 Revision Status of this Document Section 0.2, Rev. 2 Page 0.2-2

Non-Proprietary Vers ion 1014-SR-00001 Proprietary lnformatio n withheld Rev. 2 per 10CFR 2.390 @GNS 0.3 Revision Status of Sections Name, Function Date Signature Prepared 2 5.02.2022 Reviewed 2 5.02.2022 0.3 Revision Status of Sections Section 0.3, Rev. 2 Page 0.3-1

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @ GNS Section Rev. Date Summary description of change 0

0.1 0 18.12.2020 First issue 0.2 2 25.02 .2022 Addition of Revision 2 incl. list of revised sections 0.3 2 25.02.2022 Description of Revision of sections 0.4 2 25.02 .2022 Update of Table of Contents 0.5 1 25.08.2021 Addition of new Definitions 1

1.0 0 18.12.2020 First issue 1.1 0 18.12.2020 First issue 1.2 1 18.08 .2021 Correction of Table 1.2-11 1.3 0 18.12.2020 First issue 2

2.0 1 12.08 .2021 Editorial Changes 2.1 1 17.08.2021 Editorial Changes 2.2 1 13.08 .2021 Revision of Section 2.2.1 (Material Properties) 2.3 0 18.12.2020 First issue 2.4 0 21.12.2020 First issue 2.5 0 21 .12.2020 First issue 2.6 0 21 .12.2020 First issue 2.7 0 21 .12.2020 First issue 2.8 0 21.12.2020 First issue 2.9 0 21 .12.2020 First issue 2.10 0 21 .12.2020 First issue 2.11 0 21 .12.2020 First issue 2.12 1 18.08.2021 Revision of Appendix 2-2 Addition of Appendices 2-4 to 2-10 3

3.0 0 17.12.2020 First issue 3.1 1 18.08.2021 Editorial Changes 0.3 Revision Status of Sections Section 0.3, Rev. 2 Page 0.3-2

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS 3.2 1 18.08.2021 Editorial Changes Corrections Table 3.2-9 3.3 0 17.12.2020 First issue 3.4 1 18.08.2021 Editorial Changes 3.5 2 15.02.2022 Complete revision of chapter for thermal evaluation of fuel rod failure 3.6 1 25.08.2021 Implementation of chapter for thermal evaluation of short-term operations 3.7 0 18.08.2021 Relocation due to issue of new chapters 3.5 and 3.6 4

4.0 0 11 .12.2020 First issue 4 .1 1 23.02.2022 Editorial changes 4.2 1 23.02.2022 Consideration of changed temperatures and adjust-ments due to switch of regulations (NUREG2224 in-stead of NUREG/CR-6487) 4.3 1 23.02.2022 Consideration of changed temperatures and adjust-ments due to switch of regulations (NUREG2224 in-stead of NUREG/CR-6487) 4.4 0 11.12.2020 First issue 4.5 1 23.02.2022 Adjustments in Appendix 4-2 due to switch of regula-tions (NUREG2224 instead of NUREG/CR-6487) 5 5.0 0 10.12.2020 First issue 5.1 1 11.08.2021 Editorial changes 5.2 1 11 .08 .2021 Taking into account grid spacers and longer cooling times 5.3 1 11 .08 .2021 Editorial changes 5.4 1 11 .08 .2021 Plots for the axial ends of the package 5.5 0 10.12.2020 First issue 6

6.0 0 10.12.2020 First issue 6.1 0 10.12.2020 First issue 6.2 0 10.12.2020 First issue 0.3 Revision Status of Sections Section 0.3 , Rev. 2 Page 0.3-3

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS 6.3 1 11.08.2021 Editorial changes (Figure 6.3-3 , Table 6.3-5 and Figure 6.3-11)

Evaluation of fuel basket deformations 6.4 0 10.12.2020 First issue 6.5 0 10.12.2020 First issue 6.6 0 10.12.2020 First issue 6.7 0 10.12.2020 First issue 6.8 0 10.12.2020 First issue 6.9 0 10.12.2020 First issue 7

7.0 1 12.08.2021 Editorial changes 7.1 1 12.08.2021 Adjustments regarding time constraints and He-filling Deletion of usage of a transport hood Mention of canister drying criteria (vacuum pressure, residual moisture/pressure rise , hold time) 7.2 1 12.08.2021 Distinction between direct FA removal or FA removal via transfer cask Deletion of usage of a transport hood Discussion on drying procedure Thermal constraints on loading operation 7.3 0 18.12.2020 First issue 7.4 0 18.12.2020 First issue 7.5 1 12.08.2021 Editorial changes 8

8.0 0 18.12.2020 First issue 8.1 1 12.08.2021 Editorial changes More precise description of acceptance tests and crite-ria with regard to codes and standards Pneumatic test corrected into hydrostatic test Consideration of material tests on non-ASME BPVC materials 8.2 1 12.08.2021 Editorial changes More precise description of maintenance program 8.3 1 12.08.2021 Editorial changes 0.3 Revision Status of Sections Section 0.3, Rev. 2 Page 0.3-4

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS 0.4 Table of Contents Name, Function Date Signature Prepared 25.02.2022 Reviewed 25.02.2022 CASTOR is a registered trade mark.

0.4 Table of Contents Section 0.4, Rev. 2 Page 0,4-1

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS 0 GENERAL AND ORGANISATION 0.1 Document Organisation 0.2 Revision Status of this Document 0.3 Revision Status of Sections 0.4 Table of Contents 0.5 Glossary 1 GENERAL INFORMATION 1.0 Overview 1.1 Introduction 1.2 Package Description 1.3 Appendix 2 Structural Evaluation 2.0 Overview 2.1 Description of the Structural Design 2.2 Materials 2.3 Fabrication and Examination 2.4 General Requirements for all Packages 2.5 Lifting and Tie Standards for all Packages 2.6 Normal Conditions of Transport

2. 7 Hypothetical Accident Conditions 2.8 Accident Conditions for Air Transport of Plutonium 2.9 Accident Conditions for Fissile Material Packages for Air Transport 2.10 Special Form 2.11 Fuel Rods 2.12 Appendix 3 Thermal Evaluation 3.0 Overview 3.1 Description of Thermal Design 3.2 Material Properties and Component Specifications 3.3 Thermal Evaluation under Normal Conditions of Transport 3.4 Thermal Evaluation und Hypothetical Accident Conditions 3.5 Thermal Evaluation of Fuel Rod Failure 3.6 Thermal Evaluation of Short-Term Operations 0.4 Table of Contents Section 0.4, Rev. 2 Page 0.4-2

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 GNS

3. 7 Appendix 4 Containment 4.0 Overview 4.1 Description of the Containment System 4.2 Containment under Normal Conditions of Transport 4.3 Containment under Hypothetical Accident Conditions 4.4 Leakage Rate Tests for Type 8 Packages 4.5 Appendix 5 Shielding Evaluation 5.0 Overview 5.1 Description of Shielding Design 5.2 Source Specification 5.3 Shielding Model 5.4 Shielding Evaluation.

5.5 Appendix 6 Criticality Evaluation 6.0 Overview 6.1 Description of Criticality Design

  • 6.2 Fissile Material Content 6.3 General Considerations 6.4 Single Package Evaluation 6.5 Evaluation of Package Arrays und Normal Conditions of Transport 6.6 Package Arrays und Hypothetical Accident Conditions
6. 7 Fissile Material Packages for Air Transport 6.8 Benchmark Evaluation 6.9 Appendix 7 Package Operations 7 .0 Overview 7.1 Package Loading 7.2 Package Unloading 7.3 Preparation of Empty Package for Transport 7.4 Other Operations 7.5 Appendix 8 Acceptance Tests and Maintenance Program 0.4 Table of Contents Section 0.4, Rev. 2 Page 0.4-3

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS 8.0 Overview 8.1 Acceptance Tests 8.2 Maintenance Program 8.3 Appendix 0.4 Table of Contents Section 0.4, Rev. 2 Page 0.4-4

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS 3.5 Thermal Evaluation for Fuel Rod Failure Name, Function Date Signature Prepared 15.02.2022 Reviewed 15.02.2022 3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-1

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev.2 per 10CFR 2.390 @GNS In this section, the thermal evaluation in case of potential fuel rod failure is documented for the different transport conditions according to NUREG-2224 [1]

  • NCT,
  • HAC fire and
  • HAC impact.

In the following investigations, fractions of fuel rod failure of 3 % for NCT and of 100 % for HAC fire and HAC impact are considered. The fraction of fission gas release from the fuel rods due to clad-ding breach is assumed to be 0.15 for NCT and HAC fire and 0.35 for HAC impact. The fission gas release leads to a correspondingly reduced heat conductivity of the gas atmosphere inside the canister. For HAC impact, two scenarios are investigated. For scenario I, no fuel release from the fuel rods is assumed. For scenario II, a complete fuel release from the fuel rods and a reconfigura-tion of the released fuel particles inside the interior of the canister is considered.

The mechanical analyses in chapter 2 show that-the canister remains leak tight for NCT, HAC fire and HAC impact. For that reason, no fission gas release from the canister interior to the cask cavi-ty occurs.

For the calculations, the maximum heat power of about 18.5 kW is applied. Thermal requirement 2 according to section 3.1.2 is considered only, as it leads to slightly higher temperatures than the thermal requirements 1 and 3.

3.5.1 Thermal Properties of the Canister Filling Gas Mixture and FA In case of fuel rod failure, the release of fission gases from the fuel rods leads to a reduction of the heat conductivity of the gas atmosphere inside the canister. The resulting thermal properties of the gas mixtures inside the canister and the homogenized FA zones are calculated in the following paragraphs.

The input parameters for the calculations concerning the amounts of filling gases (helium) and fis-sion gases are taken from chapter 4 and are listed in Table 3.5-1. The main fission gases are xen-on (86.5 %), krypton (7.4 %) and helium (4.0 %). The thermal conductivity of the fission gas mix-ture in the canister is much lower compared to helium because the thermal conductivity decreases with increasing molar mass.

3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-2

Non-Proprietary Version 1014-SR-0000 1 Proprietary Information withheld Rev.2 per 10CFR 2.390 @GNS The fractions offailed fuel rods and of the fission gas release for NCT, HAC fire and HAC impact in Table 3.5-1 are taken from [1]. The gas mixture in the canister includes the canister filling gas (he-lium), the fuel rod filling gas (helium) and the fission gases.

The thermal conductivity of mixtures of are calculated according to [2] in de-pendence on the individual volume fraction. Table 3.5-1 contains the resulting thermal conductivi-ties -for NCT, HAC fire and HAC impact.

Table 3.5-1 Thermal conducti,vity of mixtures of helium and fission gases 3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-3

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 GNS The FA are modelled as homogenized zones with effective material values analogous to section 3.3.1.4. The effective radial thermal conductivity of the homogenized FA zones depends on ther-mal radiation and conduction. The portion of thermal conduction is reduced by the lower thermal conductivity of the gas mixture in case of fuel rod failure. The influence of the gas mixture on effec-tive axial thermal conductivity, on effective density and on the effective specific heat capacity is very low. The calculation of the effective material values is described in section 3.3.1.4 in detail.

The applied effective radial thermal conductivities of the homogenized FA zones are listed in Table 3.5-2 for the different transport conditions with fuel rod failure and the corresponding gas mixtures of Table 3.5-1.

Table 3.5-2 Effective thermal conductivities of the FA 3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-4

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS 3.5.2 Fuel Rod Failure for NCT The temperatures for NCT are calculated for a canister gas atmosphere of to Table 3.5-1.

3.5.2.1 Numerical Model The same numerical model is used as described in section 3.3.1 for NCT. Exceptions are the thermal properties of the canister filling gas mixture and homogenized FA zones, which are adapted according to section 3.5.1 3.5.2.2 Maximum Temperatures The temperature distribution of the package for NCT with fuel rod failure is shown in Figure 3.5-1.

The resulting maximum temperatures of the package are summarized in Table 3.5-3.

Below, the design-relevant temperatures are compared to their maximum admissible values ac-cording to section 3.2.2:

  • The maximum temperature of the fuel rods is 219 °C and is therefore significantly lower than the maximum admissible temperature of 400 °C.
  • The maximum temperature for the moderator rods is 110 °C, for the bottom moderator plate 121 °C and 107 °C for the lid moderator plate. Therefore, the maximum temperatures of all moderator material are far below the maximum admissible temperature of
  • The highest gasket temperature of 107 °C occurs in the tightening plug gasket. The maxi-mum admissible temperature for continuous operation of the gaskets is Therefore, all gasket temperatures are far below the temperature limit.
  • Furthermore, temperature limits for structural components (e.g. fuel basket sheets) listed in Table 3.2-9, which are relevant for the mechanical integrity, are met.

The evaluation of the results shows that all calculated maximum temperatures of the package components and the content are far below the maximum admissible values with large safety mar-gins.

3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-5

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS Figure 3.5-1 Temperature distribution for NCT with fuel rod failure 3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-6

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev.2 per 10CFR 2.390 Table 3.5-3 Component temperatures for NCT with fuel rod failure Component - type of temperature Temperature, °C Fuel rods - maximum 219 Cask surface - maximum 85 Cask surface, hottest plane - circumferential average 85 Cask surface - longitudinal average 85 Cavity surface - maximum 96 Cavity surface - hottest plane - circumferential average 95 Cavity bottom - maximum 124 Cavity bottom - surface average 117 Bottom closure plate - maximum 106 Bottom closure plate - area average 104 Moderator rods (inner row) (MR-i) - maximum 110 MR-i - area-average hottest plane, hottest rod 107 MR-i-volume-averaged, hottest rod 94 Moderator rods (outer row) (MR-a)- maximum 106 MR-a - area-average hottest plane, hottest rod 104 MR-a - volume-averaged, hottest rod 91 Inner canister surface - maximum 118 Inner canister surface - hottest plane - circ. average 116 Outer canister surface - maximum 117 Outer canister surface - hottest plane - circ. average 115 Canister bottom - maximum 139 Moderator plate (bottom) - maximum 121 Moderator plate (bottom) - volume averaged 112 Moderator plate (lid) - maximum 107 Moderator plate (lid) - volume averaged 100 Retention ring - maximum 95 Closure plate - maximum 106 Closure plate - volume averaged 105 3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-7

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev.2 per 10CFR 2.390 @GNS Table 3.5-3 Component temperatures for NCT with fuel rod failure (continued)

Component - type of temperature Temperature, °C Trunnion - maximum 103 1

Trunnion - screws - maximum 110 Fuel channels - maximum 205 Basket sheets - maximum 203 Round segment - maximum 163 Outer sheets - maximum 170 Shielding element - maximum 169 Filling gas canister - volume-averaged 171 Filling gas cask - volume averaged 103 Canister lid - maximum 116 Canister lid - volume-averaged 109 Canister lid gasket - maximum 105 Canister lid - screws 1 - maximum 110 Cask lid - maximum 99 Cask lid - volume averaged 95 Cask lid gasket - maximum 93 Cask lid - screws 1 - maximum 110 Protection cap gasket - maximum 95 Blind flange gasket- maximum 96 Tightening plug gasket 107 1

for the screw temperature, the surface temperature plus a safety margin is used 3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-8

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS 3.5.3 Fuel Rod Failure for HAC Fire The temperatures for HAC fire are calculated for a canister gas atmosphere of according to Table 3.5-1.

3.5.3.1 Numerical Model The same numerical model is used as described in section 3.4 for _HAC. Exceptions are the ther-mal properties of the canister filling gas mixture and homogenized FA zones, which are adapted according to section 3.5.1 3.5.3.2 Maximum Temperatures Figure 3.5-2 shows the maximum fuel rod temperature and maximum average temperatures of the gases in the package for HAC with fuel rod failure over time. Table 3.5-4 lists the maximum tem-peratures and their time of appearance (t = 0: beginning of fire) for various components of the package.

Below, the design-relevant temperatures are compared to their maximum admissible values ac-cording to section 3.2.2:

  • The hottest fuel rod reaches after 22 h its maximum temperature of 286 °C, which is signifi-cantly lower than the maximum admissible fuel rod temperature for HAC of 570 °C valid for intact fuel rods.
  • The temperatures of the gaskets are between 123 °C and 132 °C, which is considerably lower than the maximum admissible temperatures of I for the cask lid gasket and canister lid gasket and lforthe pressure switch gasket, the protection cap gasket and the tightening plug gasket.

The evaluation of the results shows that all calculated maximum temperatures of the package components and the content are far below the maximum admissible values with large safety mar-gins.

3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-9

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 300 280 260 240 0

220 cii

~ ~~

200

~

Cl) - Fuel rod (maximum) 0.

E 180 Cl) - Canister filling gas (volume average)

I-160 Cask filling gas (volume average) 140 120

~

I I I I  !

100 0 2 4 6 8 10 12 14 16 18 20 22 24 Time after beginning of the fire, h Figure 3.5-2 Temperatures of the hottest fuel rod and the filling gases over time for HAC fire with fuel rod failure 3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-10

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev.2 per 10CFR 2.390 @GNS I Table 3.5-4 Maximum component temperatures for HAC fire with fuel rod failure Maximum Time of Component - type of temperature temperature, °C appearance, h Fuel rods - maximum 286 22.0 Cask surface - maximum 379 0.5 Cavity surface - maximum 183 2.5 Moderator rods, inner row (MR-i) - maximum 201 1.1 MR-i - area average, hottest plane, hottest rod 182 2.2 MR-i - volume average, hottest rod 175 2.7 Moderator rods, outer row (MR-o) - maximum 310 0.5 MR-o - area average, hottest plane, hottest rod 186 0.6 MR-o - volume average, hottest rod 183 1.7 Moderator plate (bottom) - maximum 155 14.0 Moderator plate (bottom) - volume averaged 144 10.0 Moderator plate (lid) - maximum 132 29.0 Moderator plate (lid) - volume averaged 123 20.0 Canister wall - maximum 187 3.0 Basket sheets - maximum 261 23.0 Shielding elements - maximum 222 13.0 Canister filling gas - volume average 226 18.0 Cask filling gas - volume average 150 3.0 Canister lid gasket - maximum 131 11.0 Cask lid gasket - maximum 128 4.5 Protection cap gasket - maximum 123 8.5 Blind flange gasket - maximum 124 7.5.

Tightening plug gasket - maximum 132 19.0 3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-11

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev.2 per 10CFR 2.390 3.5.4 Fuel Rod Failure for HAC Impact The fuel rod failure for HAC impact considers a maximum amount of fission gas release and leads to maximum temperatures in the canister. The heating-up process needs a few days to get maxi-mum steady-state package temperatures due to the high thermal inertia. For HAC impact, two scenarios - without and with release of fuel particles from the fuel rods - are considered:

Scenario I:

The temperatures for the HAC impact are calculated without release of fuel particles for a canister gas atmosphere of I according to Table 3.5-1.

Scenario II:

This calculation is performed analogous to scenario I, but in combination with a massive fuel particle release. It is hypothetically assumed that the gaps between the components of basket, FA and canister are filled with a fuel particle packing. For the porosity of the fuel particle packing within gaps, a typical value for irregular particle sizes of - - is as-sumed.

The fuel particle packing and thereby the heat power is completely concentrated in the lid-side region of the canister interior in order to calculate maximum temperatures for the de-sign-relevant lid gaskets. The total mass of fuel in 69 FA of about 14100 kg concentrated in the lid-side region corresponds to an axial height of 1.593 m (reduced active length) where the total heat power of about 18.5 kW (18390 W) is dissipated in case of the hypothetical scenario II, see section 3.5.4.1.

3.5.4.1 Numerical Model For scenario I of HAC impact, the same numerical model is used as described in section 3.4 for HAC with the exception that the thermal properties of the filling gas mixture and homogenized FA zones are adapted according to section 3.5.1.

For scenario 11, the numerical model of scenario I is used with the exception that fuel particles are released from the fuel rods and are concentrated in the lid-side region of the canister interior. The gaps in the lid-side zones of the basket and the FA are filled with - f u e l particles and - gas mixture. The lid-side parts of the fuel rods are still filled with fuel without considering porosity. Fig-3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-12

Non-Proprietary Version 1014-SR-00001 Proprietary information withheld Rev. 2 per 10CFR 2.390 GNS ure 3.5-3 shows the 3D-FE model for scenario II with fuel reconfiguration. For better visibility, only the 3D-FE model of the canister is shown.

Figure 3.5-3 3D-FE model of canister for HAC impact with fuel rod failure - scenario II with fuel reconfiguration With a heavy metal mass of 180 kg per FA and a ratio of molar masses for U0 2 and uranium of 270/238, the total mass of the fuel in 69 FA amounts to 180 kg

  • 270/238
  • 69 = 14100 kg. The volume of the fuel in 69 FA without considering porosity is about 14100 kg / 10600 kg/m 3 = 1.33 m3 leading to a heat power density of 18390 W / 1.33 m3 = 13835 W/m 3 . The heat power density in the fuel particle packing amounts to The fractions of helium and fuel pellets in the homogenized active FA zones are 57 % and 31 %.

The former helium fraction of 57 % is now filled with the fuel particle packing. Table 3.5-5 shows the distribution of the heat power in all fuel filled zones. In the canister model (see Figure 3.5-3),

the chosen axial height of 1.593 m corresponds to a total fuel mass of about*** (fuel pellets and gaps with packing) instead of about 14100 kg for the fuel in 69 intact FA. Conservatively, this 3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-13

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev.2 per 10CFR 2.390 GNS leads to a higher total heat power of in the FE model (see Table 3.5-5) compared to the maximum value of 18390 W according to section 3.1.2.

Table 3.5-5 Heat power zones in the lid-side region of the canister interior for HAC impact with fuel rod failure - scenario II with fuel reconfiguration The thermal conductivity in the FA and gap zones inside the lid-side region of the canister interior is increased, because - of the gas mixture is replaced by the fuel particle packing. The effec-tive thermal conductivity in the fuel particle packing with a porosity of - is evaluated according to the model of [3, section - for packed beds. The following assumptions are made:

1. Temperatures of the fuel particle packing of 100 °C, 200 °C and 300 °C are considered.
2. Diameters of fuel particles between are investigated.
3. The gas atmosphere consists of All input parameters and results for a temperature of 200 °C and a particle diameter of - are summarized in Table 3.5-6. The calculations show, that particle diameters in the range o f - up to - have no effect. The effective thermal conductivity increases from at 100 °C up to I at 300 °C. For the FE calculations, a constant thermal conductivity of is chosen conservatively for gaps filled with fuel particles (see Table 3.5-7).

For the former active FA zone outside the height of 1.593 m, the thermal properties of the inactive FA zone without fuel are applied. For the reduced active zone of 1.593 m with fuel pellets, fuel packing and cladding, a constant thermal conductivity of is conservatively chosen for the radial direction in active and inactive FA zones (see Table 3.5-7). This value is only slightly higher compared to the fuel filled gaps and conservative because of the much higher conductivity of cladding and fuel pellets.

3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-14

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev.2 per 10CFR 2.390 @GNS Table 3.5-6 Calculation of effective thermal conductivity in fuel filled gaps for HAC impact with fuel rod failure - scenario II with fuel reconfiguration 3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-15

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev.2 per 10CFR 2.390 GNS The axial thermal conductivity values for the FA zones in Table 3.5-2 for HAC impact are slightly increased by the higher thermal conductivity of fuel filled zones (II, see Table 3.5-6) compared to gas filled zones (111 at 200 °C, see Table 3.5-1 ). The former free gas vol-ume in FA zones amounts to 57 %, which is now filled with the fuel particle packing.

Axial thermal conductivity of active FA zone at 200 °C:

Axial thermal conductivity of inactive FA zone at 200 °C:

The heat conductivities in fuel filled gaps and FA zones are summarized in Table 3.5-7.

Table 3.5-7 Thermal conductivity in fuel filled gaps and FA zones for HAC impact with fuel rod failure - scenario II with fuel reconfiguration 3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-16

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev.2 per 10CFR 2.390 3.5.4.2 Maximum Temperatures For HAC impact with fuel rod failure, the temperature distributions of the package are shown in Figure 3.5-4 and Figure 3.5-5 for scenario I (without fuel reconfiguration) and scenario II (with fuel reconfiguration). Table 3.5-8 and Table 3.5-9 list the maximum temperatures for various compo-nents of the package.

In comparison to scenario I without reconfiguration, scenario II with fuel reconfiguration leads to higher maximum temperatures of FA, basket and gaskets because of the higher heat power densi-ty in the lid-side region of the canister interior. In contrast, the heat resistance of fuel filled gaps and FA zones is lower compared to gas filled gaps and FA zones leading to lower temperatures in the lid-side region of the canister interior. The effect of the higher heat power density predominates over the effect of the lower heat resistance. In sum, this leads to higher maximum temperatures of FA, basket and gaskets for scenario II in comparison to scenario I.

For scenario II, the free gas volume in the lid-side region of the canister interior within the height of 1.593 m is reduced by the additional fuel particles and amounts to 0. 75 m3 (15 %). The free gas volume outside the height of 1.593 m without any fuel amounts to 4.25 m3 (85 %). The additional gas volume of the empty parts of the fuel rods (outside the height of 1.593 m) is conservatively not taken into account to get a maximum average gas temperature and pressure in the canister in case of scenario II. Due to the cold large gas volume outside the lid-side region without any heat power the averaged gas temperature for scenario II is much lower compared to scenario I.

Below, the design-relevant temperatures are compared to their maximum admissible values ac- .

cording to section 3.2.2:

  • The highest fuel rod temperatures are 293 °C (scenario I) and 321 °C (scenario II), which is far below the maximum admissible fuel rod temperature for HAC of 570 °C valid for intact fuel rods.
  • The temperatures of the gaskets are between 94 °C and 116 °C (scenario I) and between 125 °C and 175 °C (scenario II), which is considerably lower than the maximum admissible temperatures of I for the cask lid gasket and canister lid gasket and I for the pressure switch gasket, the protection cap gasket and the tightening plug gasket.

The evaluation of the results for HAC impact scenario I and scenario II shows that the calculated maximum temperatures of the package components and the content are far below the maximum admissible values with large safety margins.

3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-17

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev.2 per 10CFR 2.390 GNS Figure 3.5-4 Temperature distribution for HAC impact with fuel rod failure -scenario I without fuel reconfiguration 3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-18

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS Table 3.5-8 Maximum component temperatures for HAC impact with fuel rod failure - sce-nario I without fuel reconfiguration Maximum Component - type of temperature temperature, °C Fuel rods - maximum 293 Cask surface - maximum 83 Cavity surface - maximum 94 Moderator rods, inner row (MR-i) - maximum 117 MR-i - area average, hottest plane, hottest rod 114 MR-i - volume average, hottest rod 95 Moderator rods, outer row (MR-o)- maximum 112 MR-o - area average, hottest plane, hottest rod 110 MR-o - volume average, hottest rod 91 Moderator plate (bottom) - maximum 133 Moderator plate (bottom) - volume averaged 120 Moderator plate (lid) - maximum 117 Moderator plate (lid) - volume averaged 105 Canister wall - maximum 128 Basket sheets - maximum 264 Shielding elements - maximum 219 Canister filling gas - volume average 226 Cask filling gas - volume average 106 Canister lid gasket - maximum 111 Cask lid gasket - maximum 94 Protection cap gasket - maximum 98 Blind flange gasket - maximum 99 Tightening plug gasket - maximum 116 3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-19

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 GNS Figure 3.5-5 Temperature distribution for HAC impact with fuel rod failure - scenario II with fuel reconfiguration 3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-20

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev.2 per 10CFR 2.390 @GNS Table 3.5-9 Maximum component temperatures for HAC impact with fuel rod failure - sce-nario II with fuel reconfiguration Maximum Component - type of temperature temperature, °C Fuel rods -maximum 321 Cask surface - maximum 107 Cavity surface - maximum 126 Moderator rods, inner row (MR-i) - maximum 125 MR-i -'- area average, hottest plane, hottest rod 123 MR-i - volume average, hottest rod 92 Moderator rods, outer row (MR-o)- maximum 121 MR-o - area average, hottest plane, hottest rod 117 MR-o - volume average, hottest rod 88 Moderator plate (bottom) - maximum 94 Moderator plate (bottom) - volume averaged 90 Moderator plate (lid) - maximum 183 Moderator plate (lid) - volume averaged 152 Canister wall - maximum 162 Basket sheets - maximum 286 Shielding elements - maximum 210 Canister filling gas - volume averaged 159 Gas inside fuel particle packing (15 %, 243 °C) + gas outside packing (85 %, 147 °C)

1 / [0.15 / (243 °C + 273 K) + 0.85 / (147 °C + 273 K)] 432 K 159 °C Cask filling gas - volume averaged 127 Canister lid gasket - maximum 162 Cask lid gasket - maximum 125 Protection cap gasket - maximum 132 Blind flange gasket - maximum 134 Tightening plug gasket - maximum 175 3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-21

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev.2 per 10CFR 2.390 @GNS 3.5.5 Maximum Internal Pressures The calculation of the maximum internal pressures and a discussion on the generation of flamma-ble gases is documented in the containment evaluation in chapter 4.

3.5.6 Maximum Thermal Stresses The discussion of thermal stresses due to temperature gradients within the components can be found in the structural evaluation in chapter 2.

3.5. 7 Evaluation of Cask Performance for Fuel Rod Failure For NCT in case of fuel rod failure, it is demonstrated that the CASTOR geo69 package fulfils all requirements with regard to thermal aspects. The following items summarize the results of the thermal investigations:

  • The evaluation of the results in section 3.5.2.2 show that all calculated maximum tempera-tures of the cask components and the content are far below the maximum admissible val-ues with large safety margins.
  • It is proven that the calculated maximum temperatures of the gaskets do not lead to a deg-radation of the tightening function which is requirement for ensuring the safe enclosure of the content.
  • It is shown that the calculated maximum temperatures of the fuel rods do not lead to a deg-radation of the cladding material which is requirement for ensuring the integrity of the fuel rod cladding. The effects of potential fuel rod failure are incorporated.
  • It is demonstrated that the calculated maximum temperatures of the moderator components do not lead to a thermal degradation of the moderator material which is requirement for en-suring the effectiveness of the shielding.
  • The calculated maximum temperatures of all relevant structural components (e.g. fuel bas-ket sheets) are far below the maximum admissible values guaranteeing the mechanical in-tegrity which is requirement for ensuring heat removal performance, containment, activity retention and criticality safety.

3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-22

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 GNS

  • The evaluation of the maximum pressure and a discussion on the generation of gases is documented in the containment evaluation in chapter 4.
  • The influence of the calculated temperatures on the mechanical material properties and thermal stresses is evaluated in the structural evaluation in chapter 2.

For HAC fire and HAC impact in case of fuel rod failure, it is demonstrated that the package fulfils all requirements with regard to thermal aspects. The following items summarize the results of the thermal investigations:

  • The evaluation of the results in sections 3.5.3.2 and 3.5.4.2 show that all the calculated temperatures of cask components and the content are far below the maximum admissible values with large safety margins.
  • It is proven that the calculated maximum temperatures of the gaskets do not lead to a deg-radation of the tightening function which is requirement for ensuring the safe enclosure of the content.
  • The evaluation of the maximum pressure and a discussion on the generation of gases is documented in the containment evaluation in chapter 4.
  • The influence of the calculated temperatures on the mechanical material properties and thermal stresses is evaluated in the structural evaluation in chapter 2.

3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-23

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev.2 per 10CFR 2.390 @GNS List of References

[1] U.S.NRC NUREG-2224 Dry Storage and Transportation of High Burnup Spent Nuclear Fuel Final Report

[2] J. Kestin, K. Knierim, E. A. Mason, B. Najafi, S. T. Ro and M. Waldman Equilibrium and Transport Properties of the Noble Gases and Their Mixtures at Low Density J. Phys. Chem. Ref. Data, Vol.13, No.1, 1984

[3] VOi Heat Atlas Calculation sheets for the Heat Transfer VOi-Veriag, Dusseldorf, 9. Edition, 2002 3.5 Thermal Evaluation for Fuel Rod Failure Section 3.5, Rev. 2 Page 3.5-24

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2

@GNS per 10CFR 2.390 4.1 Description of the Containment System Name, Function Date Prepared 23 .02 .2or2..

Reviewed 4.1 Description of the Containment System Section 4.1, Rev. 1 Page4.1-1

Non-Proprietary Version 1014-SR-00001. Proprietary Information withheld Rev.2 per 10CFR 2.390 @GNS 4.1.1 Containment Boundary The containment system of the CASTOR geo69 is constituted by the following subassemblies (item numbers acc. the Design Parts Lists and *Drawings listed in section 1.3):

a) Inner containment (canister) canister body (Item 2),

canister lid (Item 3) and clamping elements (Item 4), thread bolts (Item 6) and metal gasket (Item 16 (Ag) with a torus diameter of , as well as tightening plug (Item 10) and a pressure nut (Item 11) in the canister lid and metal gasket (Item 13 (Ag) with a torus diameter of b) Outer containment (cask) cask body (Item 2),

cask lid (Item 55) and hexagonal screws (Item 62), hexagonal head screws for sealing (Item 63) and metal gasket (Item 69 (Ag) with a torus diameter of protection cap (Item 113) in the cask lid, cap screws (Item 37) and metal gasket (Item 44 (Ag) with a torus diameter of **I), as well as blind flange (Item 89) in the cask lid, cap screws (Item 37) and metal gasket (Item 71 (Ag) with a torus diameter of As listed above, the outer jackets of all metal gaskets of the containment system are made of silver (Ag). The canister body (Item 2) is designed by welding Items 2-2 to 2-5 together.

The cask CASTOR geo69 is designed for FA with moderate burn-up as well as high burn-up fuel with an averaged burn-up above 45 GWd/MgHM and therefore a double (inner and outer) containment is required. In ~ conservative way, only the outer containment (cask lid) is taken into account for the containment analysis.

This implies considering the failure of the inner containment. When considering the double containment, the calculated activity release decreases significantly. In addition, the design pressure values are separately calculated inside the inner (canister) and the outer containment (cask).

The monolithic cask body and the lids can be considered as leak-tight due to the fabrication testing which is regularly performed, so the containment analysis can be reduced to the gasket sealing system.

4.1 Description of the Containment System Section 4.1, Rev. 1 Page 4.1-2

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 The potential leakage paths of the outer containment are shown in Figure 4.1-1.

Environment T

I I I Metal gasket (Item 44, A(J) Metal gasket (Item 69, A(J) Metal gasket (Item 71 , A(J) in the protection cap (Item 113) in the cask lid (Item 55) in the blind flange (Item 89) i t i I

Welded daddings i

Fuel matrix Figure 4.1-1 Leakage paths of the outer containment The sealing effect is the result of the sealing function of the metal gaskets employed . Each metal gasket consists of a helical spring made of nickel alloy surrounded by an inner jacket of stainless steel and an outer jacket of silver.

The sealing effect of a metal gasket is based on the plastic deformation of the outer jacket, which is the result of the pretension force induced by the screwed connection of the lid . The ductility is larger for the outer jacket of the metal gasket than for the inner jacket so that the gasket will adapt to the surface structure of the sealing surface. The function of the inner jacket is to distribute uniformly the force due to pressure that is generated during the compression of the helical spring over the outer jacket. For metal gaskets, capillary leakage is the only relevant potential leakage mechanism and continuous venting is precluded .

For the gasket sealing system of each containment, a maximum reference standard helium leakage rate o f * * * * * * (leak test criterion) is proven by measurement after loading (see chapter 7). This corresponds to a reference air leakage rate of******* For the outer containment, this maximum reference air leakage rate is proven again before transportation.

The minimum support width of the metal gaskets are estimated from the torus diameter (listed in the Design Parts Lists and Drawings acc. to section 1. 3), resulting in a leakage hole length of 3.5 mm for Item 16 and Item 69 resp. 2.0 mm for Item 13, Item 44 and Item 71 , as given in Table 4.1-1 .

4.1 Description of the Containment System Section 4.1, Rev. 1 Page 4.1-3

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 GNS Table 4.1-1 Gasket parameters Estimated Torus minimum Component Item no. diameter, support width, mm mm 13 Canister 16 Cask 44 69 71 -- --

The outer containment boundary is shown in Figure 4.1-2 and in more detail at the leakage paths in Figure 4.1-3 and Figure 4.1-4. The inner containment boundary is shown in Figure 4.1-5 and in more detail at the leakage path in Figure 4.1-6.

There are no valves or pressure relief systems in the containment.

Due to pretension of the screws (see chapter 2), unintended opening and opening due to internal pressure are excluded.

Due to the inner containment system, which constitutes the second containment, the transportation of high burn-up fuel with an averaged burn-up above 45 GWd/MgHM and of a plutonium content in solid form exceeding 0.74 TBq is permissible (cf. Appendix4-1).

Due to the materials used for the containment system, significant chemical, galvanic or other reactions are excluded (see section 2.2.2).

4.1 Description of the Containment System Section 4.1, Rev. 1 Page 4.1-4

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 Figure 4.1-2 Outer containment (cask) 4.1 Description of the Containment System Section 4.1, Rev. 1 Page 4.1-5

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev.2 per 10CFR 2.390 @GNS Figure 4.1-3 Containment boundary detail at the cask lid with protection cap Figure 4.1-4 Containment boundary detail at the cask lid with blind flange 4.1 Description of the Containment System Section 4.1, Rev. 1 Page 4.1-6

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev.2 per 10CFR 2.390 @GNS Figure 4.1-5 Inner containment (canister)

Figure 4.1-6 Containment boundary detail at the canister lid with tightening plug 4.1 Description of the Containment System Section 4.1, Rev. 1 Page4.1-7

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS 4.1.2 Codes and Standards For calculating activity release from the cask to define allowable leakage rates, analytical calculations in case of overpressure compared to the ambient atmosphere according to ANSI N14.5 [1] by means of the Knudsen equation are performed. The volume leakage rates through the containment are consistently calculated in accordance with ANSI N14.5 [1] and ISO 12807 [2].

The compliance of the CASTOR geo69 containment system with the permitted activity release limits specified in 10 CFR 71, § 71.51 (a)(1) for normal conditions of transport (NCT) and

§ 71.52 (a)(2) for hypothetical accident conditions (HAC) is demonstrated.

4.1.3 Special Requirements for Damaged Spent Nuclear Fuel The content of the CASTOR geo69 cask specified in section 1.2.2 does not include damaged SNF, so there are no special requirements regarding damaged SNF necessary.

List of References

[1] ANSI N14.5-2014, American National Standard For Radioactive Materials - Leakage Tests on Packages for Shipment

[2] ISO 12807:201 B(E)

Safe transport of radioactive materials - Leakage testing on packages 4.1 Description of the Containment System Section 4.1, Rev. 1 Page 4.1-8

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 GNS per 10CFR 2.390 4.2 Containment under Normal Conditions of Transport Name, Function Date Prepared 2J.. 0'2. 2b2.1.

Reviewed 4.2 Containment under Normal Conditions of Transport Section 4.2, Rev. 1 Page 4.2-1

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS 4.2.1 Pressurization The canister inside the CASTOR geo69 cask contains SNF during NCT. The interior space inside the canister is drained, dried, evacuated and backfilled with helium gas prior to final closure of the canister. The dry interior space inside the cask with a loaded canister is evacuated and backfilled with helium gas prior to final closure of the cask. Therefore, no vapors or gases are present which could cause a reaction or explosion inside the canister and the cask. Procedural steps ensure a maximum total pressure of PHe,o (cf. Appendix 4-2) inside the canister as well as inside the cask and prevent over-pressurization during dispatching of the cask.

With the procedure described in Appendix 4-2 the maximum internal pressure for NCT is calculated for the two different containments assumed in section 4.1.1. There are no combustible gases inside the containment.

Assuming the inner containment, the maximum absolute pressure Pu = inside the canister results with the boundary conditions given in Table 4.2-1.

Assuming the outer containment, the maximum absolute pressure Pu = inside the cask results with the boundary conditions given in Table 4.2-2. For this calculation the two different areas filled with helium (canister and cask) are combined to one summed volume. The amount of filling gas helium in each area is added to the amount of gas inside the cask (n = nHe,canister +

nHe,cask + nFG + nFR) substituting the summand PHe,o

  • T gas / T He,o in the equation for Pu (Appendix 4-2).

Without implying the failure of the inner containment while regarding the outer containment, the maximum absolute pressure Pu = ** inside the cask without impacts from the content is obtained with Pu =PHe,o

  • Tgas/ T He,o (Appendix 4-2) and T gas =- (covering NCT value for the filling gas of the cask in chapter 3).

The maximum normal operating pressure (MNOP) is the value of the upstream absolute pressure Pu for NCT, reduced by the atmospheric pressure at mean sea level, i. e .. 101.3 kPa. With a maximum absolute pressure Pu 4.2 Containment under Normal Conditions of Transport Section 4.2, Rev. 1 Page4.2-2

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS Table 4.2-1 Boundary conditions for canister pressure calculation (NCT)

Table 4.2-2: Boundary conditions for cask pressure calculation (NCT)

With the procedure described in Appendix 4-2, the minimum internal pressure Pu = resp.

Pu= results for the inner resp. outer containment boundary.

4.2 Containment under Normal Conditions of Transport Section 4.2, Rev. 1 Page4.2-3

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390' Structural integrity and containment of the CASTOR geo69 package are not impaired in NCT (cf.

chapter 2).

4.2.2 Containment Criteria The basic requirements for licensing the casks for safe transport of radioactive materials are laid out in the 10 CFR 71. According to § 71.51 the following requirements have to be fulfilled for NCT.

Under the tests to demonstrate the package's ability to withstand NCT (1 O CFR 71. 71 ),

- the loss or dispersal of radioactive contents must be restricted to max. 1o-s A2 per hour,

- no significant increase in external surface radiation levels may result and

- no substantial reduction of the effectiveness of the CASTOR geo69 packaging may result.

The A2 value of mixtures of radionuclides is calculated as stipulated in appendix A, paragraph IV in 85 10 CFR71. This includes the use of 10 A2 for Kr also for NCT, as it is present in a mixture of nuclides. In line with 10 CFR71, the ambient pressure is assumed as 25 kPa for NCT.

4.2.3 Compliance with the Containment Criteria As the containments are not impaired in NCT (cf. chapter 2), the design leakage rate of the considered containment is not greater than 10-7 ref* cm 3/s (leak-tight according to chapter 7) and no significant increase in external surface radiation levels caused by potentially released fines or crud is ensured. Therefore, no dedicated activity release calculations are required.

Nevertheless, the reference air leakage rate corresponding to the allowable leakage hole diameter for NCT which would lead to 100 % utilization of the NCT limit value is calculated for the outer containment as follows.

With the procedure described in Appendix 4-2 and using values for fs, fo from Appendix 4-2 and for V from Table 4.2-2, the activity mobilization inside the cask is shown in Table 4.2-3 for gases and volatiles, in Table 4.2-4 for fines and crud and in Table 4.2-5 summed up for the nuclide mixture for NCT. With the A2 values from 10 CFR71, the mobilized activity and activity concentration are calculated for each nuclide as well as for each category of nuclides depending on their mobility type (effective values). The listed activity fraction fN(i) is the ratio between the mobilized content AN(i) for nuclide i and the sum of mobilized contents. The relative values of fNU)/A2U) in Table 4.2-5 show that crud contributes most to activity release (in terms of utilization of limit value).

4.2 Containment under Normal Conditions of Transport Section 4.2, Rev. 1 Page4.2-4

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 NS Table 4.2-3 Mobilized activity and activity concentration for gases and volatiles for NCT 4.2 Containment under Normal Conditions of Transport Section 4.2, Rev. 1 Page4.2-5

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS Table 4.2-4: Mobilized activity and activity concentration for fines and crud for NCT 4.2 Containment under Normal Conditions of Transport Section 4.2, Rev. 1 Page4.2-6

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS Table 4.2-5: Mobilized activity and activity concentration summed up for the nuclide mixture for NCT The allowable leakage rate NCT is obtained from the allowable activity release rate RN via LN(Pu) = RN /CN, where CN denotes the sum of mobilized activity concentration in Table 4.2-5. With the A2 value of of the nuclide mixture (see Table 4.2-5) and the containment criterion that the loss or dispersal of radioactive contents must be restricted to max. 1o-s A2 per hour (cf.

section 4.2.2; RN , the following allowable leakage rate is obtained:

for NCT.

The maximum allowable leakage hole diameter for which the allowable leakage rate LN yields is calculated by solving the Knudsen equation (cf. Appendix 4-2) for NCT. For the calculations, the cask atmosphere is considered as helium gas. This is conservative, because it yields a lower viscosity and smaller molecular mass, and thus results in a smaller allowable leakage hole diameter and smaller reference air leakage rate compared to the mixture with relevant contributions of fission gas (e.g. Xenon). As described in section 4.2.2, a reduced ambient pressure of Pd = 25 kPa is considered. The gasket temperature valid for the gaskets of the outer containment is taken from chapter 3. The relevant parameters as well as the allowable leakage hole diameter which result from solving the Knudsen equation for NCT are shown in Table 4.2-6.

The leakage hole diameter calculated in Table 4.2-6 is used to determine a reference leakage rate of dry air leaking from Pu = 1 atm to Pd= 0.01 atm at a temperature of 298 K. The relevant parameters as well as the resulting reference air leakage rate at standard conditions are shown in Table 4.2-7.

This assessment is independent of filters or a mechanical cooling system.

4.2 Containment under Normal Conditions of Transport Section 4.2, Rev. 1 Page4.2-7

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev.2 per 10CFR 2.390 G S Table 4.2-6: Allowable leakage hole diameter corresponding to the allowable leakage rate for NCT Table 4.2-7: Reference air leakage rate corresponding to the allowable leakage hole diameter for NCT List of References

[1] NUREG-2224, November 2020 Dry Storage and Transportation of High Burnup Spent Nuclear Fuel 4.2 Containment under Normal Conditions of Transport Section 4.2, Rev. 1 Page4.2-8

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 4.3 Containment under Hypothetical Accident Conditions Name, Function Date Prepared 2i. 02. 2022 Reviewed 4.3 Containment under Hypothetical Accident Conditions Section 4.3, Rev. 1 Page 4.3-1

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS 4.3.1 Pressurization The canister inside the CASTOR geo69 cask contains SNF during HAC. The interior space inside the canister is drained, dried, evacuated and backfilled with helium gas prior to final closure of the canister. The interior space inside the cask with a loaded canister is evacuated and backfilled with helium gas prior to final closure of the cask. Therefore, no vapors or gases are present which could cause a reaction or explosion inside the canister and the cask. Procedural steps ensure a maximum total pressure of PHe,o (cf. Appendix 4-2) inside the canister as well as inside the cask and prevent over-pressurization during dispatching of the cask.

According to [1], there are two separate cases to be analyzed for hypothetical accident conditions:

HAG-fire and HAG-impact.

With the procedure described in Appendix 4-2 the maximum internal pressures for HAG-fire and HAG-impact are calculated for the two different containments assumed in section 4.1.1. There are no combustible gases inside the containment.

HAG-fire:

Assuming the inner containment, the maximum absolute pressure Pu = inside the canister results with the boundary conditions given in Table 4.3-1.

Assuming the outer containment, the maximum absolute pressure Pu = inside the cask results with the boundary conditions given in Table 4.3-2. For this calculation the two different areas filled with helium (canister and cask) are combined to one summed volume. The amount of filling gas helium in each area is added to the amount of gas inside the cask (n = nHe,canister +

nHe,cask + nFG + nFR) substituting the summand PHe,o

  • Tgas / T He,o in the equation for Pu (cf. Appendix 4-2).

Without implying the failure of the inner containment while regarding the outer containment, the maximum absolute pressure Pu = inside the cask without contributions from the content is obtained by Pu = PHe,o

  • Tgas / T He,o (cf. Appendix 4-2) and T gas =- (covering HAG-fire value for the filling gas of the cask in chapter 3).

Structural integrity and containment of the CASTOR geo69 package are not impaired under HAG-fire ( cf. chapter 2).

4.3 Containment under Hypothetical Accident Conditions Section 4.3, Rev. 1 Page4.3-2

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS Table 4.3-1 Boundary conditions for canister pressure calculation (HAC-fire)

Table 4.3-2: Boundary conditions for cask pressure calculation (HAC-fire) 4.3 Containment under Hypothetical Accident Conditions Section 4.3, Rev. 1 Page4.3-3

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS HAC-im pact:

Chapter 3 provides two scenarios for HAG-impact regarding release and reconfiguration of spent nuclear fuel from the inside of the fuel rod into the free volume of the canister after fuel rod failure:

- HAG-impact I: without release and reconfiguration of spent fuel

- HAG-impact II: with release and reconfiguration of spent fuel Assuming the inner containment, the maximum absolute pressure Pu = (HAG-impact I) resp. Pu = * * * <HAG-impact II) inside the canister results with the boundary conditions given in Table 4.3-3.

Assuming the outer containment, the maximum absolute pressure Pu = (HAG-impact I) resp. Pu = * * * (HAG-impact II) inside the cask results with the boundary conditions given in Table 4.3-4. For this calculation the two different areas filled with helium (canister and cask) are combined to one summed volume. The amount of filling gas helium in each area is added to the amount of gas inside the cask (n = nHe,Canister + nHe,Cask ,+ nFG + nFR) substituting the summand PHe,o

  • T gas/ T He,o in the equation for Pu (cf. Appendix 4-2).

Without implying the failure of the inner containment while regarding the outer containment, the maximum absolute pressure Pu = ** (HAG-impact I) resp. Pu = ** (HAG-impact II) inside the cask without contributions from the content is obtained by Pu= PHe,o

  • T gas/ T He,o (cf. Appendix 4-2) and Tgas = - resp. T gas = - (covering HAG-impact values for the filling gas of the cask in chapter 3).

Structural integrity and containment of the CASTOR geo69 package are not impaired under HAC-impact (cf. chapter 2).

4.3 Containment under Hypothetical Accident Conditions Section 4.3, Rev. 1 Page4.3-4

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS Table 4.3-3 Boundary conditions for canister pressure calculation (HAC-impact)

Table 4.3-4: Boundary conditions for cask pressure calculation (HAC-impact) 4.3 Containment under Hypothetical Accident Conditions Section 4.3, Rev. 1 Page4.3-5

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS 4.3.2 Containment Criteria The basic requirements for licensing the casks for safe transport of radioactive materials are laid out in the 10 GFR 71. According to§ 71.51 the following requirements have to be fulfilled for HAG.

Under the tests to demonstrate the package's ability to withstand HAG (10 GFR 71. 73),

- there would be no escape of 85 Kr exceeding 10 A2 per week and

- no escape of other radioactive material exceeding a total amount A2 per week.

The A2 value of mixtures of radionuclides is calculated as stipulated in appendix A, paragraph IV in 10 GFR71. In line with 10 GFR71, the ambient pressure is assumed as 100 kPa for HAG.

4.3.3 Compliance with Containment Criteria As the containments are not impaired in HAG (cf. chapter 2), the design leakage rate of the considered containment is not greater than 10-7 ref*cm 3/s (leak-tight according to chapter 7).

Therefore, no dedicated activity release calculations are required.

Nevertheless, the reference air leakage rate corresponding to the allowable leakage hole diameter for HAG which would lead to 100 % utilization of the HAG limit value is calculated for the outer containment separately for HAG-fire and HAG-impact as follows.

With the procedure described in Appendix 4-2 and using values for fs, fo from Appendix 4-2 and for V from Table 4.3-2 and Table 4.3-4, the activity mobilization inside the cask is shown in Table 4.3-5 and Table 4.3-8 for gases and volatiles, in Table 4.3-6 and Table 4.3-9 for fines and crud as well as in Table 4.3-7 and Table 4.3-10 summed up for the nuclide mixture for HAG-fire and HAG-impact, respectively. Because the release fractions described for HAG-impact in

. Appendix 4-2 do not differ for the two scenarios described in section 4.3.1, the described activity mobilization from Table 4.3-8 to Table 4.3-1 O for HAG-impact is valid for both scenarios.

With the A2 values from 10 GFR71, the mobilized activity and activity concentration are calculated for each nuclide as well as for each category of nuclides depending on their mobility type (effective values). The listed activity fraction fA(i) is the ratio between the mobilized content AA(i) for nuclide i and the sum of mobilized contents. The relative values of fAU)/A2U) in Table 4.3-7 and Table 4.3-10 show that fines contribute most to activity release (in terms of utilization of limit value) for HAG-fire and crud contributes most for HAG-impact.

4.3 Containment under Hypothetical Accident Conditions Section 4.3, Rev. 1 Page4.3-6

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 (@)GNS Table 4.3-5: Mobilized activity and activity concentration for gases and volatiles for HAC-fire 4.3 Containment under Hypothetical Accident Conditions Section 4.3, Rev. 1 Page4.3-7

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 Table 4.3-6: Mobilized activity and activity concentration for fines and crud for HAC-fire 4.3 Containment under Hypothetical Accident Conditions Section 4.3, Rev. 1 Page4.3-8

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS Table 4.3-7: Mobilized activity and activity concentration summed up for the nuclide mixture for HAC-fire Table 4.3-8: Mobilized activity and activity concentration for gases and volatiles for HAC-impact 4.3 Containment under Hypothetical Accident Conditions Section 4.3, Rev. 1 Page4.3-9

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 @GNS Table 4.3-9: Mobilized activity and activity concentration for fines and crud for HAC-impact 4.3 Containment under Hypothetical Accident Conditions Section 4.3, Rev. 1 Page4.3-10

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 G S Table 4.3-10: Mobilized activity and activity concentration summed up for the nuclide mixture for HAC-impact The allowable leakage rates for HAG-fire and HAG-impact are obtained from the allowable activity release rate RA via LA(Pu) = RA /CA, where CA denotes the sum of mobilized activity concentration in Table 4.3-7 and Table 4.3-10, respectively. With the A2 values of *** and respectively, of the nuclide mixture (see Table 4.3-7 and Table 4.3-10) and the containment criteria to section 4.3.2 (here: RA =

, the following allowable leakage rates are obtained:

for HAG-fire and for HAG-impact.

The maximum allowable leakage hole diameter for which the allowable leakage rate LA yields is calculated by solving the Knudsen equation (cf. Appendix 4-2) for HAC. Similar to NCT, the cask atmosphere is considered as helium gas. As described in section 4.3.2, an ambient pressure of Pd = 100 kPa is considered. The gasket temperature valid for the gaskets of the outer containment is taken from chapter 3. The relevant parameters as well as the allowable leakage hole diameters for HAG-fire, HAG-impact I and HAG-impact II are shown in Table 4.3-11.

The leakage hole diameters calculated in Table 4.3-11 are used to determine reference leakage rates of dry air leaking from Pu= 1 atm to Pd= 0.01 atm at a temperature of 298 K. The relevant parameters as well as the resulting reference air leakage rates at standard conditions are shown in Table 4.3-12 for HAG-fire, HAG-impact I and HAG-impact II.

This assessment is independent of filters or a mechanical cooling system.

4.3 Containment under Hypothetical Accident Conditions Section 4.3, Rev. 1 Page 4.3-11

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2 per 10CFR 2.390 GNS Table 4.3-11: Allowable leakage hole diameter corresponding to the allowable leakage rate for HAC Table 4.3-12: Reference air leakage rate corresponding to the allowable leakage hole diameter for HAC List of References

[1] NUREG-2224, November 2020 Dry Storage and Transportation of High Burn up Spent Nuclear Fuel 4.3 Containment under Hypothetical Accident Conditions Section 4.3, Rev. 1 Page 4.3-12

Non-Proprietary Version 1014-SR-00001 Proprietary Information withheld Rev. 2

@GNS per 10CFR 2.390 4.5 Appendix Name, Function Prepared Reviewed Appendix 4-1 Content Appendix 4-2 Calculation Concepts 4.5 Appendix Section 4.5, Rev. 1 Page 4.5-1

Non-Proprietary Version APPENDIX 4-2 to 1014-SR-f B~f6~;~ ~~~o;~ation withheld Calculation Concepts Assumptions

- The crud activity of is estimated for a cask loading.

- An initial fuel rod filling gas pressure of is assumed for all fuel rods of all FA in the cask loading.

Residual water vapor is excluded in the calculations regarding design pressure values.

Determination of design pressure values Information on the minimum and maximum pressure Pu inside the canister or the cask is required for various analyses.

The absolute pressure is obtained by using the ideal gas law Pu = n

  • Runiv
  • T gas / V, where n is the amount of gas, Runiv is the universal gas constant 8.314 J/mol/K, Tgas is the absolute gas temperature (volume average) and Vis the free gas volume inside the canister or the cask. In addition, the assumed maximum helium filling partial pressure of the canister is temperature-corrected by the gas temperature T g~s under test conditions. Therefore, Pu is calculated by:

Pu = PHe,O

  • T gas/ T He,O + n
  • Runiv
  • T gas/ V.

Procedural steps ensure a maximum total pressure of PHe,o =

inside the canister as well as inside the cask.

To determine the maximum absolute pressure Pu, the values of n and T gas are maximized while V is minimized. Therefore, the influence parameters are estimated as follows:

maximum amount of gas n:

All relevant gas contributions have to be added. This includes the maximum amount of gas released from the content, i. e. fission gas and filling gas of the fuel rods. How this value is deduced is explained below.

maximum absolute gas temperature T gas:

In the context of the thermal design calculations, the maximum volume averaged gas temperature is calculated for various test conditions.

minimum free gas volume V:

The free gas volume is calculated based on the canister and cask design. From the canister cavity volume, the displacement volumes of the basket and the fuel assemblies are subtracted. The free gas volume inside fuel rods that are considered to have failed is not included in the total free gas volume.

Page 1 of 5

Non-Proprietary Version APPENDIX 4-2 to 1014-SR-fB~r6~f~ h~~o;~ation withheld For maximum pressure considerations, the gas release from the content is calculated as follows:

- The fraction of failed fuel rods fs is assumed as 0.03 (3 %) for NGT and as 1 (100 %)

for HAG (see Table 1, according to [1]).

- The maximum total amount of filling gas of the fuel rods is determined based on the information provided for the fuel rods. For each fuel rod that is assumed to have failed, the full amount of filling gas is assumed to be released into the canister resp. the cask.

- The produced amount of fission gas is determined via burn-up calculations.

- According to [1], the fraction of fission gas release fo is used as 0.15 (15 %) for NGT as well as for HAG-fire and as 0.35 (35 %) for HAG-impact which conservatively includes extra 20 % of the fission gases retained within the pellet grain boundaries that might be released during a drop impact (see Table 1). For each fuel rod that is assumed to have failed, the fraction fo of the produced fission gas is assumed to be released into the cavity of the canister resp. cask.

The moderator disc between the canister lid and the cask lid might cause an additional amount of radiolysis gas from irradiation. The energy dose from gamma irradiation of the lid-end moderator disc, which is made of the ultrahigh molecular weight polyethylene - is given as about*** in the shielding evaluation. Taking a G-value for hydrogen of 4 molecules per 100 eV for polyethylene (with ultrahigh molecular weight) into account ( cf. [2]), a negligible amount of results in a year.

After the cask drying process, no residual water has to be assumed to be present as vapor after dispatch. Further gases are not formed during operation of the package, either.

The maximum pressure which can occur is the value of the absolute pressure Pu for HAG with either the combination of high internal temperatures due to thermal test conditions and the covering value for fuel rod failure (HAG-fire) or the combination of high internal equilibrium temperatures due to the thermal impact of the 20 % extra fission gases and the covering value for fuel rod failure (HAG-impact).

The minimum internal cask pressure that can occur is determined by considering the minimum amount of filling gas of the canister resp. the cask PHe,o = ** and a minimum assumed gas temperature (set to the ambient temperature of -40 °G, see test condition 10 GFR 71. 71 (c)(2)).

Furthermore, gas released from the content is neglected.

Page 2 of 5

Non-Proprietary Version th APPENDIX 4-2 to 1014-SR-cw~~r6~~~ ~~o;~ation wi held Activity mobilization The activity content is classified in four categories, in line with the approach in [1]. The categories are gaseous substances, volatile substances, particulate substances from fuel and particulate substances from crud. Based on [1], the nuclides are classified as follows: 3 H and 85 Kr 106 134 137 Cs as gases, Ru, Cs and as volatiles, all nuclides from Appendix 4-1 as particulate 60 substances from fuel (fines) and Co as particulate substance from crud.

The fraction of failed fuel rods fs and the fraction of fission gas release fo are introduced above.

According to [1],

- the fraction of volatiles that are released due to a cladding breach is fv = 3

  • 10-5 for NGT, HAG (including HAG-fire and HAG-impact),

- the mass fraction of fuel that is released as fines due to a cladding breach is fF = 3

  • 10-5 for NGT as well as for HAG-impact and fr = 3
  • 10-3 for HAG-fire and

- the fraction of crud that spalls-off rods is fc =0.15 for NCT and f c =1.0 for HAG.

The release fractions are compiled in Table 1.

Table 1: Release fractions of radioactive materials Variable Symbol NCT HAC-fire HAC-impact Fraction of Fuel Rods fs 0.03 1.0 1.0 Assumed To Fail Fraction of Fission Gases Released fG 0.15 0.15 0.35 Due to a Cladding Breach Fraction of Volatiles Released fv 3E-05 3E-05 3E-05 Due to a Cladding Breach Mass Fraction of Fuel Released fF 3E-05 3E-03 3E-05 as Fines Due to a Cladding Breach Fraction of Crud fc 0.15 1.0 1.0 Spalling off Cladding The activity concentration of the gases 3 H and 85 Kr is obtained by multiplying the total activity of the concerned nuclides from Appendix 4-1 with the fraction of failed fuel rods fa and the fraction of fission gas release fo and dividing the result by the free gas volume V. In the same way but using the fraction of failed fuel rods fa and the fraction fv resp. fF the activity concentrations for volatiles resp. particulate substances from fuel (fines) are defined. For the activity concentration of crud only the fraction f c is taken into account.

Page 3 of 5

Non-Proprietary Version APPENDIX 4 2 t0 1014 SR doo9firietary Information withheld

- - - pe'r1 OCFR 2.390 Volumetric leakage rate The volumetric leakage rate L from the cask atmosphere is calculated by means of the Knudsen equation [3]:

with Lc(Pa) viscous volumetric flow rate, m3/s Lm(Pa) : molecular volumetric flow rate, m3/s

µ dynamic viscosity of the gas, Pa

  • s M molar mass of the gas molecules, kg/mol R universal gas constant (R = 8.314 J
  • mo1-1
  • K-1),

T gas temperature within the leakage hole (gasket temperature), K Pu upstream pressure, Pa Pd downstream pressure, Pa Pa average pressure (Pa = (Pu + Pd)/2), Pa a leakage hole length, m D leakage hole diameter, m.

The Knudsen equation describes the combined viscous (first term) and molecular (second term) flow of a gaseous substance with temperature T, molar mass M, and dynamic viscosity

µ. The flow through a straight leakage hole (with length a and diameter D) is driven by the pressure difference between the absolute upstream pressure Pu and the absolute downstream pressure Pd- The Knudsen equation is valid for the average pressure Pa= (pu + pd)/2 and is renormalized to the operation conditions of the cask via L(pJ= L(pa)* Pa/Pu to allow multiplica-tion with the upstream activity concentration for calculating the activity release rates.

Page 4 of 5

Non-Proprietary Version APPENDIX 4-2 to 1014-SR-CWB~r6~f~ ~~o;~ation withheld List of References

[1] NUREG-2224, November 2020 Dry Storage and Transportation of High Burnup Spent Nuclear Fuel

[2] AMEC/200615/001 Issue 3, Determination of G-values for use in SMOGG gas generation calculations

[3] ANSI N14.5-2014, American National Standard For Radioactive Materials - Leakage Tests on Packages for Shipment Page 5 of 5

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