ML23012A094

Non-Proprietary (Public), Enclosure 3: Safety Analysis Report for 10 CFR 72 Application of the Castor geo69 Dry Storage System (Chapter 4, Thermal Evaluation, Pg. 4.0-1, to Chapter 14, Quality Assurance, Pg. 14.0-2)

ML23012A094
Person / Time
Site:	07201052
Issue date:	12/21/2022
From:	Bussmann D GNS Gesellschaft fur Nuklear-Service mbH
To:	Office of Nuclear Material Safety and Safeguards
Shared Package
ML23012A102	List: ML23012A094 ML23012A096 ML23012A097
References
T1213-CO-00014 1014-SR-00002, Rev 1
Download: ML23012A094 (497)

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Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 4 Thermal Evaluation 4.0 Overview Name, Function Date Signature Prepared 10.11.2022 Reviewed 10.11.2022 4.0 Overview Section 4.0, Rev. 1 Page 4.0-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS In this chapter, it is shown that the CASTOR geo69 storage cask fulfils all requirements for stor-age with regard to thermal aspects. The thermal design for storage complies with the performance requirements of 10 CFR part 72 for normal conditions of storage (NCS), off-normal conditions and accident conditions of storage (ACS). During storage, the cask stands in vertical position on the storage pad of a free-field storage facility.

Furthermore, short-term operations inside the reactor facility are investigated. This comprise the fuel loading of the canister inside the transfer cask under water, the dewatering, vacuum drying and helium backfilling of the canister interior as well as the transfer of the transfer cask inside the reactor building. Additionally, the on~site transfer of the CASTOR geo69 storage cask to the stor-age pad is considered.

The temperature distribution within the cask and the maximum temperatures of the components are calculated by numerical methods applying the finite element method (FEM).

The proofs show that all calculated temperatures are below the corresponding maximum admissi-ble values with large safety margins and a save heat removal during storage is ensured.

The following verification objectives are considered for the thermal design under NCS and off-normal conditions:

1. The safe enclosure of the content has to be ensured. For this purpose, the temperatures of the design-relevant lid gaskets shall not exceed the admissible limit value of 111111°C accord-ing to Chapter 8 for continuous operation. Furthermore, the temperatures of the lid gaskets and the mean temperature of the canister filling gas are. used for the verification of the con-tainment in Chapter 7.

2. The integrity of the fuel rod cladding has to be ensured. For this purpose, it is shown that the maximum cladding temperature does not exceed the admissible limit value of 400 °C according to [1].

3. The effectiveness of the shielding has to be ensured. For this purpose, the maximum tem-peratures for the moderator components shall stay permanently below the maximum ad-missible temperature of - °C for . . . - according to Chapter 8. Furthermore, the mean temperatures of the moderator components are taken into account for the shielding analyses in Chapter 5.

The admissible component temperatures of the cask are not exceeded under NCS and meet suffi-ciently large safety margins.

4.0 Overview Section 4.0, Rev. 1 Page 4.0-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The following verification objectives are taken into account for the thermal design under ACS:

4. The safe enclosure of the content has to be ensured even for ACS. Therefore, the maxi-mum temporary temperatures of the cask and canister lid gaskets are limited to llllll'C and for the pressure switch gasket, the protection cap gasket and the tightening plug gasket to llllll'C. These are the maximum temporary temperatures for which failure of the metal gas-kets can be excluded according to Chapter 8.

5. The cladding temperatures are shown to remain below the admissible value of 570 °C for ACS according to [1].

6. The thermal degradation of the moderators shielding effectiveness under ACS is taken into account in the shielding analyses in Chapter 5, therefore the assessment of the moderator component temperatures is omitted.

The admissible component temperatures of the cask are not exceeded under ACS and meet suffi-ciently large safety margins.

List of References

[1] Cladding Considerations for the Transportation and Storage of Spent Fuel Interim Staff Guidance - 11, Rev. 3 Spent Fuel Project Office U.S. Nuclear Regulatory Commission (2003) 4.0 Overview Section 4.0, Rev. 1 Page 4.0-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 4.1 Discussion Name, Function Date Signature Prepared 10.11.2022 Reviewed 10.11.2022 4.1 Discussion Section 4.1, Rev. 1 Page 4.1-1 I

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 4.1.1 Design Features A detailed description of the CASTOR geo69 storage cask can be found in Section 1.2. The cask consists of a thick-walled cask body made of ductile cask iron (DCI) and an inner canister made of stainless steel. The canister accommodates up to 69 spent fuel assemblies (FA) from boiling water reactors (BWR). Their maximum decay heat amounts to 18.5 kW, while different homogeneous and heterogeneous loading patterns are possible. A description of the different loading patterns can be found in Section 4.1 .2.

During storage, the cask stands in vertical position on the storage pad of a free-field storage facili-ty. Deviating from the other chapters of this SAR, within the thermal evaluations it is assumed that the cask is fixed with a storage frame to the base plate of the storage pad. This conservatively leads to slightly higher temperatures compared to the case without storage frame. In storage con-figuration, a protection cover is mounted at the top of the cask.

The safe enclosure of the content is ensured by two independent barriers, the canister with re-openable lid and the cask with bolted lid. Both are sealed by metal gaskets. The temperature of the gaskets is examined in the thermal evaluation for NCS and ACS to ensure the long-term tightening function.

The FA are kept in position by the basket sheets and the outer sheets which ensure besides criti-cality safety a sufficient heat removal from the FA. The basket sheets are made of boronized alu-minium, the outer sheets are made of stainless steel. Additional components inside the canister are the round segments and the shielding elements which are both made of aluminium.

The heat removal within the cask is achieved by conduction, convection and radiation. Conduction takes place mainly inside the walls of the cask and the canister and within the basket sheets. Ra-diation takes place between all surfaces which border on gas atmosphere. Heat removal from the cask surface to the environment takes place by convection and radiation to the ambience without using active cooling mechanisms or additional coolants. In order to improve the heat dissipation on the outer surface, the cask lateral surface is equipped with cooling fins. Due to the high thermal resistance of the gaps in the lid system, the majority of the thermal energy is dissipated by the fin zone.

To enhance heat removal from the FA, the space between cask and canister and the interior of the canister are backfilled with helium, which has a comparably high thermal conductivity. Additionally, the temperature gradients that occur radially cause a convective flow, which further enhances the heat removal.

4.1 Discussion Section 4.1, Rev. 1 Page 4.1-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Inside the cask wall, - moderator rods are placed for radial neutron shielding. Furthermore, moderator components are made ol moderator plates are placed in the cask bottom and in the lid area for axial neutron shielding. All The containment relevant components of the storage cask are designed in accordance with Divi-sion 3 of the BPVC.

4.1.2 Description of the Content The cask can be loaded with 69 FA of six different types from BWR according to Section 2.1. The types of FA considered here are listed in Table 4.1-1. The active length of the FA and the axial heat power distribution depend on the type of the FA, the burn-up profile additionally depends on the irradiation time and the decay time. For the modelling of the FA, the most unfavourable charac-teristics are considered, see Section 4.4.2.4 for details.

The maximum summarized decay heat of the FA is limited to 18.5 kW per cask. Three bounding loading patterns are defined, all fitting into a certain scheme of six position groups of FA in the fuel basket. These loading patterns are identified by their decay heat per FA and are called thermal requirements. The three thermal requirements considered here are shown in Figure 4.1-1.

The total heat power of the cask loading amounts to I

I I

- - For the thermal evaluation of NCS, all thermal requirements are examined. The most unfa-vourable thermal requirement is investigated for the evaluation of ACS.

4.1 Discussion Section 4.1, Rev. 1 Page 4.1-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 4.1-1 Considered types of FA Figure 4.1-1 Thermal requirements 1-3 with maximum decay heat per FA position in W 4.1 Discussion Section 4.1, Rev. 1 Page 4.1-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 4.1.3 Summary Tables of Temperatures The component temperatures of the cask and the content are summarized and discussed in the following Sections:

• for NCS hot case in Section 4.4.3 and Table 4.4-5

• for NCS cold case in Section 4.4.4

• for off-normal conditions of storage in Section 4.5.4.2 and Table 4.5-1

• for ACS fire in Section 4.6.1.2 and Table 4.6-2

• for ACS burial in Section 4.6.2.2 and Table 4.6-4

• for ACS impact in Section 4.6.3.2, Table 4.6-8 and Table 4.6-9

• during fuel loading under water in Section 4.7.2

• for the water-filled cask in Section 4.7.3

• during vacuum drying in Section 4.7.4.4 and Table 4.7-5

• after helium backfilling in Section 4.7.5.4 and Table 4.7-6

• during on-site transfer for normal conditions in Section 4.7.6.2 and Table 4.7-7

• during on-site transfer for off-normal conditions in Section 4.7.7

• during on-site transfer for accident conditions fire in Section 4.7.8.2 and Table 4.7-8

• during on-site transfer for accident conditions impact in Section 4.7.9 The applicable temperature limits of the materials and components are given in Table 4.3-1.

4.1.4 Summary Tables of Maximum Pressures The calculation of the maximum pressures for NCS, off-normal conditions and ACS and a discus-sion on flammable gases can be found in the containment evaluation in Chapter 7.

4.1 Discussion Section 4.1, Rev. 1 Page 4.1-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 4.2 Summary of Thermal Properties of Materials Name, Function Date Prepared 10.11.2022 Reviewed 10.11.2022 4.2 Summary of Thermal Properties of Materials Section 4.2, Rev. 1 Page 4.2-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The relevant material data of the components of the storage cask and content used in the thermal calculations are

• the thermal conductivity k, W/(m*K),

• the density p, kg/m 3 ,

• the specific heat capacity c, J/(kg*K),

• the emissivity E and

• the solar absorption coefficient a.

The following Section 4.2.1 describes the material data of solid components and pure gases. Sec-tion 4.2.2 discusses the material data of gas mixtures in case of fuel rod failure.

4.2.1 Thermal Data of Solid Components and Pure Gases The material data used in the thermal calculations are taken from Chapter 8 ir:, accordance with Section 11, Part D (Metric) of the BPVC [1]. Some material properties base on manufacturers speci-fications and individual material qualifications. The material properties of gases and concrete as well as the surface emissivities are taken from appropriate technical literature.

Materials used for the storage cask are ductile cast iron for the cask body and stainless steel for the canister, the bottom closure plate, the retention ring and the lids. The moderator material, which is used for the moderator rods *and the moderator plates, is made of - The materi-al data of the cask and canister components are listed in Table 4.2-1.

The basket sheets are made of boronized aluminium. The round segments and the shielding ele-ments are made of aluminium. The outer sheets are made of steel. The material data of the basket components are listed in Table 4.2-2.

The fuel assemblies contain a various number of fuel rods filled with fuel pellets made of uranium oxide. The structural components such as cladding, water rods and fuel channels are made of zircaloy. The interior of the FA is backfilled with helium. The material data of the FA components are listed in Table 4.2-3.

Within the numerical mod~I of the cask, the FA are not modelled in detail but are replaced by ho-mogenized zones. The effective material data of these zones .are calculated in advance, which is described in Section 4.4.2.4. The calculated effective material data of the homogenized zones are listed in Table 4.2-4 for the case of pure helium as filling gas. Section 4.2.2 contains the effective 4.2 Summary of Thermal Properties of Materials Section 4.2, Rev. 1 Page 4.2-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 material data of the homogenized FA zones for the case of a filling with a gas mixture caused by fuel rod failure (Table 4.2-11 ).

Table 4.2-5 contains the coefficients of thermal expansion for some relevant components.

The interior of the canister and the space between cask and canister are backfilled with helium.

The gap around the bottom moderator plate is filled with dry air. The material data of the pure gas-es are listed in Table 4.2-6. Section 4.2.2 contains the material data of gas mixtures in case of fuel rod failure (Table 4.2-10).

Table 4.2-7 contains the material data of storage-specific components. This comprises the material data for the protection cover, the storage frame and the concrete base plate of the storage pad.

The enhancement of the thermal conductivity due to the reinforcement of the concrete base plate is neglected conservatively.

The surface emissivities of the components exchanging heat by thermal radiation are listed in Ta-ble 4.2-8.

Table 4.2-9 contains the solar absorption coefficients of surfaces exposed to insolation, which con-servatively are set to 1 for all surfaces. In reality, the solar absorption coefficient for the painted cask surfaces amounts to 0.4 (light colour) according to [2].

4.2 Summary of Thermal Properties of Materials Section 4.2, Rev. 1 Page 4.2-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS I Table 4.2-1 Material data of cask and canister components Component Tempera-. Heat Specific heat Density, Reference ture, conductivity, capacity, Material oc kg/m 3 W/(m*K) J/(kg*K) 20 37.5 455 50 38.5 461 Cask body 100 39.73 479 150 40.45 495 Ductile cast iron Chapter 8 7200 200 40.73 508 SA-874M 250 40.64 521 300 40.23 535 325 39.93 542 20 14.1 492 100 15.4 511 150 16.1 519 Canister body 200 16.8 526 Chapter 8 8030 SA-240M 316L 250 17.6 533 300 18.3 540 350 19.0 546 400 19.7 553 20 11.1 482 100 12.5 510 150 13.3 522 Canister lid 200 14.2 536 Chapter 8 7810 SA-965M FXM-19 250 15.0 544 300 15.9 556 350 16.7 564 400 17.5 570 Cask lid Retention ring Chapter 8 See SA-240M 316L SA-182M Gr F316 Closure plate SA-240M Chapter 8 See SA-240M 316L 22Cr-5Ni-3Mo-N Moderator rods Moderator plates Chapter 8 (bottom / lid) 4.2 Summary of Thermal Properties of Materials Section 4.2, Rev. 1 Page 4.2-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Table 4.2-2 Material data of basket components Component Tempera- Heat Specific heat Density; Reference ture, conductivity, capacity, Material oc kg/m 3 W/(m*K) J/(kg*K) 20 134.0 898 50 137.6 909 Round segments 75 140.8 923 100 143.6 934 Shielding elements Chapter 8 2690 125 145.9 942 SB-209 Alloy 5454 150 148.0 951 175 150.0 961 200 151.9 972

- Basket sheets Chapter 8 L ____

Outer sheets Chapter 8 See SA-240M 316L SA-240M 316 Table 4.2-3 Material data of fuel assembly components Component Tempera- Heat Specific heat Density, Reference ture, conductivity, capacity, Material oc kg/m 3 W/(m*K) J/(kg*K) 4.2 Summary of Thermal Properties of Materials Section 4.2, Rev. 1 Page 4.2-5

Non-Proprietary Version 1014-SR-00002 Rev. 1 Proprietary Information withheld per 10CFR 2.390 s

Table 4.2-4 Calculated material data of the homogenized FA zones applied in the 3D FE model (pure helium as filling gas)

Tempera- Radial heat Axial heat Specific heat Density, Material zone ture, conductivity, conductivity, capacity, oc kg/m 3 W/(m*K) W/(m*K) J/(kg*K)

Table 4.2-5 Coefficient of thermal expansion of relevant components Component Tempera- Coefficient of Reference ture, thermal expansion, Material oc 10-5 1/K 50 10.5 Cask body 100 10.9 Ductile cast iron Chapter 8 150 11.3 SA-874M 200 11.8 225 12.0 50 15.6 Canister body 100 16.2 Chapter 8 150 16.6 SA-240M 316L 200 17.0 225 17.2 Basket sheets Chapter 8 4.2 Summary of Thermal Properties of Materials Section 4.2, Rev. 1 Page 4.2-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 I Table 4.2-6 Material data of pure gases Tempera- Heat Specific heat Density, Fluid Reference ture, conductivity, capacity, oc kg/m 3 W/(m*K) J/(kg*K) 20 0.0257 100 0.0314 Air (dry) 200 0.0380 720

[5] 1.188 300 0.0441 (p =canst.)

400 0.0500 500 0.0556 25 0.1536 100 0.1793 3115 Helium [5] 200 0.2116 0.166 (p =canst.)

300 0.2420 400 0.2708 Table 4.2-7 Material data of storage-specific components Component Tempera- Heat Specific heat Density, Reference ture, conductivity, capacity, Material oc kg/m 3 W/(m*K) J/(kg*K) 20 60.4 431 50 59.8 453 100 58.0 480 Protection cover 150 55.9 500 Storage frame 1 Chapter 8 200 53.6 7750 516 SA-738M Gr. C 250 51.4 534 300 49.2 553 350 47.0 575 400 44.9 600 1

Material assumed for storage frame.

4.2 Summary of Thermal Properties of Materials Section 4.2, Rev. 1 Page 4.2-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Tabl~ 4.2-8 Surface emissivities Surface Surface Reference emissivity Cask body, outer surface (painted)

Cask body, cavity surface (coated)

Cask body (untreated)

Canister (stainless steel)

Cask lid (stainless steel)

Canister lid (stainless steel)

Closure plate (stainless steel)

Retention ring (stainless steel)

Moderator rods and discs (polyethylene)

Cladding and water rods (zircaloy, untreated)

Basket sheets (aluminium, anodized)

Outer sheets (stainless steel)

Fuel channels (zircaloy, untreated)

Round segments (aluminium, anodized)

Shielding elements (aluminium, anodized)

Protection cover (painted)

Storage frame (painted)

Base plate of storage pad (concrete)

Ambience (wide-ranging) .chosen 1 Emissivity of the fire 10 CFR 71 (§ 71.73(c)(4)) 0.9 Sooted surfaces (during fire and cooling phase) 10 CFR 71 (§*71.73(c)(4)) 0.8

- Table 4.2-9 Solar absorption coefficients of surfaces Surface Reference Absorption coefficient Cask outer surface (painted, light colour) chosen 1 Storage frame outer surface (painted, light colour) chosen 1 Base plate of storage pad (concrete) chosen 1 4.2 Summary of Thermal Properties of Materials Section 4.2, Rev. 1 Page 4.2-8

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 4.2.2 Thermal Data of Gas Mixtures for Fuel Rod Failure The thermal evaluations include the case of potential fuel rod failure according to NUREG-2224 [7]

for the different storage conditions

• NCS,

• off-normal conditions,

• ACS fire (ACS burial analogously) and

• ACS impact.

In the thermal investigations, fractions of fuel rod failure of 1 % for NCS, 10 % for off-normal condi-tions and 100 % for ACS are considered. The fraction of fission gas release from the fuel rods due to cladding breach is assumed to be 0.15 for NCS, off-normal conditions, ACS fire and ACS burial and 0.35 for ACS impact. The fission gas release leads to a correspondingly reduced heat conduc-tivity of the gas atmosphere inside the canister. For ACS impact, two scenarios are investigated.

For scenario I, no fuel release from the fuel rods is assumed. For scenario II, a complete fuel re-lease from the fuel rods and a reconfiguration of the released fuel particles inside the interior of the canister is considered.

The mechanical analyses in Chapter 3 show that the canister remains leak-tight for NCS, off-normal conditions and ACS. For that reason, no fission gas release from the canister interior to the cask cavity occurs.

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 7 and are listed in Table 4.2-10. The main fission gases are xenon (86.5 %), krypton (7.4 %) and helium (4.0 %). The thermal conductivity of the fission gas mixture in the canister is much lower compared to helium because the thermal conductivity de-creases with increasing molar mass. Xenon has the lowest thermal conductivity of all fission gases and is considered therefore exclusively for the following calculations.

4.2 Summary of Thermal Properties of Materials Section 4.2, Rev. 1 Page 4.2-9

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The fractions of failed fuel rods and of the fission gas release for NCS, off-normal conditions, ACS fire, ACS burial and ACS impact in Table 4.2-10 are taken from [7]. The gas mixture in the canister includes the canister filling gas (helium), the fuel rod filling gas (helium) and the fission gases.

The thermal conductivity of mixtures of helium and xenon are calculated according to [8] in de-pendence on the individual volume fraction. Table 4.2-10 contains the resulting thermal conductivi-ties of the different fractions of xenon for NCS, off-normal conditions, ACS fire, ACS burial and ACS impact.

- Table 4.2-10 Thermal conductivity of mixtures of helium and fission gases The FA are modelled as homogenized zones with effective material values analogous to Section 4.4.2.4. 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 the 4.2 Summary of Thermal Properties of Materials Section 4.2, Rev. 1 Page 4.2-10

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 (_@)GNS effective axial thermal conductivity, on the effective density and on the effective specific heat ca-pacity is very low. The calculation of the effective material values is described in Section 4.4.2.4 in detail.

The applied effective radial and axial thermal conductivities of the homogenized FA zones are listed in Table 4.2-11 for the different storage conditions with fuel rod failure and the corresponding gas mixtures of Table 4.2-10.

Table 4.2-11 Effective thermal conductivities of the FA for fuel rod failure List of References

[1] 2017 ASME Boiler and Pressure Vessel Code Sectio_n II Materials Part D Properties (Metric)

[2] R. C. Weast et al.

Handbook of Chemistry and Physics CRC Press, Boca Raton, Florida, 6ih Edition, 1986-1987

[3] SCALE 4.4a: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation for Workstations and Personal Computers Oak Ridge National Laboratory NUREG/CR-0200 ORNL/RSIC, CCC-545, 2000

[4] MATPRO-Version 11 (Revision 2): A Handbook of Materials Properties for Use in the Anal-ysis of Light Water Reactor Fuel Rod Behaviour US Nuclear Regulatory Commission, NUREG/CR-0497-REV-2, 1981 4.2 Summary of Thermal Properties of Materials Section 4.2, Rev. 1 Page 4.2-11

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1

[5] VDI Heat Atlas Calculation Sheets for the Heat Transfer Springer, 2010

[6] U. Grigul, U. Sandner Warmeleitung Springer-Verlag, Berlin, 1979 Heat Conduction

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

[8] 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 4.2 Summary of Thermal Properties of Materials Section 4.2, Rev. 1 Page 4.2-12

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 4.3 Specifications for Components Name, Function Date Prepared 10.11.2022 Reviewed 10.11.2022 4.3 Specifications for Components Section 4.3, Rev. 1 Page4.3-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The temperatures of the following components of the cask and content are limited to certain maxi-mum admissible values under NCS and off-normal conditions (see Section 4.0):

• the lid gaskets,

• the fuel rod cladding and

• the moderator material.

The applicable temperature limits of the components and materials are listed in Table 4.3-1.

The integrity of the fuel rod cladding has to be ensured. For this purpose, the maximum cladding temperature shall not exceed the admissible limit of 400 °C for NCS and off-normal conditions ac-cording to [1 ].

- The safe enclosure of the content has to be ensured. For this purpose, the temp~ratures of the design-relevant lid gaskets have to stay below the admissible limit value of ~ C according to Chapter 8 for continuous operation. Furthermore, the temperatures of the lid gaskets and the mean temperature of the filling gas in the cask cavity are used for the verification of the containment in Chapter 7.

The effectiveness of the shielding is ensured when the maximum temperatures of the moderator components are below the maximum admissible application temperatures of - °C for . . .

- according to Chapter 8. Furthermore, the mean temperatures of the moderator components are used for the shielding analyses in Chapter 5.

The shielding elements and the round segments in the fuel basket are made of aluminium (SB-209 Alloy 5454). For these components it is shown that the calculated temperatures are far below the melting temperature of 603 °C.

Under ACS, the safe enclosure of the content has to be ensured. Therefore, the maximum temper-atures of the cask and canister lid gaskets are limited to - °C. The maximum temperatures of the gaskets of the protection cap, the pressure switch and the tightening plug are limited to - °C.

These are the maximum temporary temperatures for which failure of the metal gaskets can be ex-cluded according to Chapter 8.

The cladding temperatures are shown to remain below the admissible value of 570 °C for ACS according to [1].

As the thermal degradation of the moderators shielding effectiveness under ACS is taken into ac-count in the shielding analyses in Chapter 5, assessment of the moderator component tempera-tures is omitted.

4.3 Specifications for Components Section 4.3, Rev. 1 Page 4.3-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Table 4.3-1 Temperature limits of components Maximum admissible temperature, 0 c Component Literature (Material) NCS and off-normal ACS conditions Cladding [1] 400 570 Gaskets Chapter 8 Moderator material Chapter 8 Cask body Chapter 8 343 (Ductile cast iron SA-874M)

Canister body Chapter 8 427 (SA-240M 316L)

Head ring Chapter 8 427 (SA-182M FXM-19)

Canister lid Chapter 8 427 (SA-965M FXM-19)

Cask lid, retention ring Chapter 8 427 (SA-182M Gr F316)

Closure plate Chapter 8 316 (SA-240M 22Cr-5Ni-3Mo-N)

Basket sheets Chapter 8 (AI-B4C-MMC)

Outer sheets Chapter 8 427 (SA-240M 316)

List of Reference

[1] Cladding Considerations for the Transportation and Storage of Spent Fuel Interim Staff Guidance - 11, Rev. 3 Spent Fuel Project Office U.S. Nuclear Regulatory Commission (2003) 4.3 Specifications for Components Section 4.3, Rev. 1 Page 4.3-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 (_@GNS 4.4 Thermal Evaluation for Normal Conditions of Storage Name, Function Date Prepared 10.11.2022 Reviewed 10.11.2022 4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 (@)GNS In this section, the thermal performance of the cask during storage under NCS is evaluated. The cask stands in vertical position on the storage pad of a free-field storage facility. The cask stands in an array with other casks of equal type.

4.4.1 Environmental Conditions Under NCS, the environmental conditions applied in the thermal calculations take into account the requirements of 10 CRF 71 (§ 71. 71) for the hot case and the cold case.

According to § 71. 71, the environmental temperature for the hot case amounts to 38 °C ( 100 °F).

Conservatively, this high value for the environmental temperature is used in the thermal calcula-tions as a steady-state temperature. In spite of the high thermal inertia of cask and content, varia-tions of the environmental temperature within several days and fluctuations during day and night are not taken into account.

According to § 71. 71, insolation over a period of 12 h per day has to be considered for the hot case with the following solar heat fluxes:

• Flat surfaces orientated horizontally

- Base none 2

- Other surfaces 800 W/m 2 > 800 g-cal / cm / 12 h

• Flat surfaces not orientated horizontally 200 W/m 2 > 200 g-cal / cm 2 / 12 h

• Curved surfaces 400 W/m 2 > 400 g-cal / cm 2 / 12 h Due to the high thermal inertia of cask and content, fluctuations of the insolation during day and night have a negligible influence on the temperatures of the design-relevant components of cask and content. Accordingly, the solar heat fluxes are applied in the thermal calculations as steady-state values averaged over 24 h per day.

The absorption coefficient of surfaces exposed to insolation is conservatively set to 1. In reality, the solar absorption coefficient for the painted cask surfaces amounts to 0.4 (light paint), see Section 4.2.1.

In the ground beneath the storage pad base plate at a depth of 8 m, a constant temperature of 10 °c (50 °F) occurs for continental climate. This temperature underlies neither daily nor seasonal fluctuations and is applied in the thermal calculations.

4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS For the cold case under NCS, an environmental temperature of -40 °C and no insolation are ap-

• plied in the thermal analyses in accordance with § 71. 71.

4.4.2 Thermal Model The calculation methods based on a numerical model following the FEM including the geometrical model, the boundary conditions and further assumptions under NCS are described in the following sections. For the evaluation under NCS, steady-state calculations are carried out for the cask standing in an array of equal casks on a storage pad.

4.4.2.1 Calculation Methods The calculation methods for the thermal analyses are validated in Section 4.8, Appendix 4-1 for the thermal design of transport and storage casks for steady-state and transient investigations. This includes especially the comparison of calculation results with experimental test results.

For the numerical calculations, the FEM software ANSYS Mechanical [1] is used. ANSYS, Inc.

extensively verifies and validates the ANSYS program as part of a quality assurance process.

Additionally, the used program version is verified and validated in Section 4.8, Appendix 4-2 espe-cially for the application to the thermal design of transport and storage casks.

4.4.2.2 Geometric Modelling For the numerical calculations, a three-dimensional fi~ite element (3D FE) model is used, repre-senting in circumferential direction one half of the cask and content taking into account its sym-metry. The FE model is shown in Figure 4.4-1 to Figure 4.4-3. The FE model consists of approxi-mately finite elements. The FE model contains all thermally relevant components of the design parts lists. Section 1.1 contains the design parts lists and the corresponding design draw-ings with the component dimensions. The material properties as described in Section 4.2 are used.

In the storage configuration, the cask itself and the content are identical to the transport configura-tion. In contrast to the transport configuration, impact limiters are not mounted. In storage configu-ration, a protection cover is mounted at the top of the cask (see Figure 4.4-1). The cask is fixed with a storage frame to the base plate of the storage pad. Additionally, the FE model contains a section of the base plate of the storage pad and 4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 The FA are not resolved in detail but are modelled as homogenized zones with effective material properties. The calculation of these effective material properties is described in Section 4.4.2.4.

The cooling fins on the cask cylindrical surface are not modelled explicitly, too. Instead, the im-proved convective heat transfer is taken into account by an effective surface enlargement factor, which is explained in Section 4.4.2.3.

- As heat transfer mechanisms thermal conduction in the solids and the gas atmospheres as well as heat radiation between all free surfaces are considered. Heat transfer by radiation is modelled ex-plicitly using the radiation matrix method in ANSYS [1]. All gases inside the cask are assumed as stagnant. The positive effects on heat transfer by convective flows are neglected.

Figure 4.4-1 3D FE model of the storage cask - overall view 4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 4.4-2 3D FE model of the storage cask - detailed view 4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS i

Figure 4.4-3 3D FE model of the storage cask - cross section All components are modeled with nominal dimensions from the design drawings, which are related to an assembly temperature of 20 °C. The manufacturing tolerances of the components are tight, so that their impact on the temperature distribution is neglected. Exceptions are the radial gaps between cask and canister and between canister and basket because they significantly influence the heat removal which mainly takes place in radial direction. This is discussed in the following paragraphs.

4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 (@)GNS Table 4.4-1 Thermal expansion for the outer and inner radial gap 4.4.2.3 Modelling of the Fins The cask cylindrical surface is equipped with radial fins in order to improve the convective heat removal. The cooling fins are not modelled explicitly in the FE model. Instead, the improved con-vective heat transfer is taken into account by an effective surface enlargement factor. At the smooth cylindrical cask surface in the FE model, where the fins are located, the convective heat transfer coefficient (see Section 4.4.2.8) is multiplied by this surface enlargement factor according-ly.

4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev: 1 @GNS

- --*----. I I

-*** - I

• (T I) I~

I I

I I

...---*________..Jj

- Free cask surface AFree

- Surface of fins AFin

- Base surface of fins A 8

! . * .,I

- -- --- ------ -- ___]

Figure 4.4-4 Schematic representation of a fin 4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-8

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @)GNS Table 4.4-2 Input values for the calculation of the effective surface enlargement factor The surface enlargement factor is applied to the convective amount of the heat transfer only. The enhancement of the radiative heat transfer by the fins is conservatively neglected. The reason is that the lateral surfaces of the fins are almost perpendicular to the ambient, facing towards neigh-boring fins (see Figure 4.4-4). As neighboring fins have a similar temperature distribution, almost*

no heat is transferred between the lateral fin surfaces. Furthermore, the radiative heat transfer be-tween the lateral surfaces of the fins and the ambient is small as the view factors are very small.

4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-9

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 4.4.2.4 Modelling of the Fuel Assemblies The FA are not modelled in detail in the 3D FE model. Instead, a simplified geometry is used as shown in Figure 4.4-1 to Figure 4.4-3.

4.4.2.4.1 Effective Axial Thermal Conductivity, Density and Heat Capacity

- I I I I I I I

4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-10

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per' 10CFR 2.390 Rev. 1 4.4.2.4.2 Effective Radial Thermal Conductivity I

I I

I 4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-11

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 (@)GNS I

I I

I I

4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-12

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 4.4-5 E model of a fuel assembly of type GE 8x8-1 4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-13

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 4.4-6 Temperature distribution 4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-14

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 4.4.2.4.3 Axial Heat Power Distribution The decay heat power is distributed heterogeneously along the active length of the FA with a peak-ing approximately at half height (see Figure 4.4-7). The detailed axial heat power distributions are calculated for all intended types of FA and for various decay times. The calculations are based on the burn-up calculations presented in Section 2.1. The resulting peaking factors are summarized in Table 4.1-1 in Section 4.1.2.

Figure 4.4-7 Detailed and simplified axial heat power profiles 4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-15

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 are shown in Figure 4.4-7.

4.4.2.4.4 Results and Summary As a result of the homogenization calculations, Table 4.4-3 lists the effective thermal conductivities in radial and axial direction for the different FA types according to Table 4.1-1 for the case of pure helium as filling gas.

Because the main portion of the heat power is removed radially from the FA, the type of FA with the lowest effective radial heat conductivity leads to the highest fuel rod temperatures. For that reason, the effective material properties of the FA type GE 8x8-1 are used in the 3D FE model.

Table 4.2-4 lists the effective radial and axial thermal conductivity, the effective density and effec-tive specific heat capacity used in the 3D FE model for the case of pure helium as filling gas. Sec-tion 4.2.2 contains the effective material data for the case of a filling with a gas mixture caused by fuel rod failure (Table 4.2-11 ).

4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-16

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 4.4-3 Effective radial and axial thermal conductivity of considered FA types (pure helium as filling gas) 4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-17

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @)GNS 4.4.2.5 Filling Gases The free space within the canister and the space between canister and cask are backfilled with the filling gas helium. The gap around the bottom moderator plate is filled with dry air.

During operation of the cask under NCS, a failure of 1 % of the fuel rods in conjunction with a frac-tion of fission gas release from the fuel rods of 0.15 is considered according to Section 4.2.2. The fission gas release leads to a correspondingly reduced heat conductivity of the gas atmosphere inside the canister. Accordingly, the thermal properties of the canister filling gas mixture and the homogenized FA zones are applied according to Table 4.2-10 and Table 4.2-11 under considera-tion of fuel rod failure.

The mechanical analyses in Chapter 3 show that the canister remains leak-tight for NCS. For that reason, no fission gas release from the canister interior to the cask cavity occurs.

4.4.2.6 Heat Removal Mechanisms The heat removal from the storage cask to the ambience is purely passive. An active heat removal system is not necessary.

The decay heat generated in the pellets is transferred by thermal conduction and radiation to the cladding. From there, it is transferred via thermal radiation, convection and conduction to the fuel channels, the fuel basket and the canister body and lid. Between the canister and the cask bodies and lids, the heat transfer also occurs by conduction, convection and radiation. Within the canister and cask bodies and lids, the heat is transferred by conduction only. Most of the heat is transferred radially to the cask surface.

Conservatively, the calculation does not take into account the convective heat transfer inside the canister and the cask. Therefore, in all gaps filled with gas atmosphere, only conduction and radia-tion is considered.

From the outer surface of the cask, the heat is mainly removed by convection to the environmental air. To enhance convective heat transfer, the cylindrical surface of the cask is equipped with fins.

Additionally, heat is transferred by thermal radiation to the ambience. The heat removal by radia-tion is restricted by the neighbouring casks having the same surface temperature because they are assumed to be of the same type and have the same heat power. A minor part of the heat is re-moved by conduction through the base plate of the storage pad.

4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-18

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 4.4.2.7 Modelling of the Neighbouring Casks Under NCS, the storage cask stands in an array with other casks of equal type on the base plate of a storage pad. The pitch of th~ array (distances between the cask axes) amounts to 3 m. Figure 4.4-8 (left) shows the cask standing in an array with other casks.

4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-19

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 4.4.2.8 Boundary Conditions Thi_s section describes the boundary conditions applied under NCS for the hot case. The storage cask stands in an array of equal casks on a storage pad. The heat removal from the cask surface occurs by free convection to the surrounding air and radiative heat transfer to the ambience. The heat removal by thermal radiation is reduced by neighbouring casks For the ambient temperature for convection T A,canv, an environmental temperature of 38 °C accord-ing to Section 4.4.1 is applied. Conservatively, this value is used as a steady-state temperature.

On the cylindrical surface of the cask, a convective heat transfer coefficient hcanv is applied which is derived from the following Nusselt law according to [2] for turbulent heat transfer by free convection at a vertical cylinder:

4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-20

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The Nusselt number Nu is defined as:

Nu= h

• L / k with: h - heat transfer coefficient, W/(m 2 *K)

L - characteristic length (cask height), m k - thermal conductivity, W/(m*K)

The Grashot number Gr is defined as:

Gr= g

• f3

• Lff

• L3 / v2 with: g - gravitational acceleration, g = 9.81 m/s 2 f3 - coefficient of thermal expansion, 1/K tiT - difference between cask surface and ambient temperature Tc - TA,conv, K L - characteristic length (cask height), m v - kinematic viscosity, m2/s The Prandtl number Pr is defined as:

Pr= v / a with: a - thermal diffusivity, m2/s v - kinematic viscosity, m2/s The temperature dependency of the material properties of air is considered according to Table 4.4-4. According to [5], for all material data the mean value over the cask surface temperature and the bulk temperature is taken as reference temperature with the exception of the coefficient of thermal expansion, for which the bulk temperature is used.

For the convective heat transfer within the fin zone of the cask surface, an effective surface en-largement factor- is applied which is explained in Section 4.4.2.3. The convective heat trans-fer coefficient hconv is multiplied by this factor.

At the lid-side front face of the cask, an effective heat transfer coefficient of is applied, which consists of a convective portion hconv and a radiative portion hrad*

The convective portion hconv is calculated using the following Nusselt law according to [2] for turbu-lent heat transfer by free convection at a horizontal plane in case of heat release at the upper side:

4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-21

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 4.4-4 Material properties of dry air according to [2]

Temperature, k, 13, v, Pr, oc 10*3 W/(m*K) 10*3 1/K 10*7 m2/s -

20 25.69 3.421 153.5 0.7148 30 26.43 3.307 162.9 0.7134 40 27.16 3.200 172.7 0.7122 60 28.60 3.007 192.7 0.7100 80 30.01 2.836 213.5 0.7083 100 31.39 2.683 235.2 0.7070 120 32.75 2.546 257.5 0.7060 140 34.08 2.422 280.7 0.7054 160 35.39 2.310 304.6 0.7050 180 36.68 2.208 329.3 0.7049 200 37.95 2.115 354.7 0.7051 250 41.06 1.912 421.2 0.7063 300 44.09 1.745 491.8 0.7083 350 47.05 1.605 566.5 0.7109 400 49.96 1.486 645.1 0.7137 800 71.54 0.932 1402 0.7342 4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-22

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 lnsolation is considered with the solar heat fluxes according to Section 4.4.1 averaged over 24 h per day. On the cylindrical surfaces of the cask and the storage frame, solar heat fluxes of 200 W/m 2 for curved surfaces are applied. At the lid-side front face of the cask and the free upper 2

surfaces of the storage pad base plate and the storage frame, solar heat fluxes of 400 W/m for upward-facing flat surfaces orientated horizontally are used.

4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-23

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 4.4.3 Maximum Temperatures The resulting maximum temperatures for various components of the storage cask and the content are summarized in Table 4.4-5 for NCS with fuel rod failure. The three thermal requirements, as defined in Section 4.1.2, only have a minor impact on the temperatures. For thermal requirement 3, the highest temperatures of the design-relevant components result. The temperature differences compared to thermal requirement 2 are below the rounding tolerance of 1 K. The temperatures of thermal requirement 1 are up to 3 K lower compared to thermal requirement 3.

Figure 4.4-9 shows the temperature distribution in the cask for thermal requirement 3. Figure 4.4-10 shows the temperature distribution in the hottest cross-section of the FA for thermal re-quirements 1 to 3. Figure 4.4-11 shows the axial temperature profile on the cask lateral surface and the cask cavity surface for thermal requirement 3. Figure 4.4-12 and Figure 4.4-13 show the temperature distributions in the lids of cask and canister for thermal requirement 3.

Below, the design-relevant temperatures are compared to their maximum admissible values ac-cording to Section 4.3:

• The maximum temperature of the fuel rods amounts to 227 °C and is therefore significantly lower than the maximum admissible temperature of 400 °C.

• The maximum temperature for the moderator rods is 118 °C, for the bottom moderator plate 130 °C and 103 °C for the lid moderator plate. Therefore, the maximum temperatures of all moderator material are far below the maximum admissible temperature of . , C .

• The highest gasket temperature of 108 °C occurs in the tightening plug gasket. The maxi-mum admissible temperature for continuous operation of the gaskets is . , C . Therefore, all gasket temperatures are far below the temperature limit.

• Furthermore, temperature limits for structural components (e.g. fuel basket sheets) listed in Table 4.3-1, which are relevant for the mechanical integrity, are met.

The evaluation of the results show that all calculated maximum temperatures of the cask compo-nents and the content are far below the maximum admissible values with large safety margins.

Large additional safety margins exist because of the conservative approaches for the thermal modelling. The main conservatisms are listed below:

• All components are centrically arranged in their respective mounting positions without con-tact. This applies especially to the FA, which are centrically arranged in the basket posi-4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-24

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 tions, as well as the basket, which is centrically arranged in the canister, and the canister, which is centrically arranged in the cask. In reality, local contacts between the components exist which enhance the heat removal.

• The individual basket sheets have no contact among each other. Instead, gaps are mod-elled between the individual basket sheets in radial and axial direction. This applies also to the connection to the round segments and the shielding elements.

• The enhanced heat transport by convection in the gas atmospheres inside cask and canis-ter cavities is neglected completely.

• Conservative boundary conditions and material properties are applied.

4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-25

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 I Table 4.4-5 Component temperatures for NCS with fuel rod failure for thermal require-ments 1-3 Temperature, °C Component - type of temperature Thermal requirement 1 2 3 Fuel rods - maximum 225 227 227 Cask lateral surface - maximum 104 105 105 Cask lateral surface - maximum circumferential average 102 102 103 Cask lateral surface - maximum longitudinal average 96 97 97 Cask cavity lateral surface - maximum 118 119 119 Cask cavity lateral surface - maximum circumferential average 117 118 118 Cask cavity bottom - maximum 131 133 133 Cask cavity bottom - area average 125 126 126 Closure plate, underside - maximum 114 115 115 Closure plate, underside - area average 112 113 113 Moderator rods, inner row (MR-i) - maximum 117 118 118 MR-i - maximum cross-sectional average, hottest r:od 114 115 116 MR-i - volume average, hottest rod 104 105 105 Moderator rods, outer row (MR-o) - maximum 113 114 114 MR-o - maximum cross-sectional average, hottest rod 111 112 112 MR-o - volume average, hottest rod 101 102 102 Canister inner lateral surface - maximum 128 130 129 Canister inner lateral surface - max. circumferential average 126 128 128 Canister outer lateral surface - maximum 127 128 128 Canister outer lateral surface - max. circumferential average 125 127 127 Canister bottom - maximum 147 148 148 Moderator plate, bottom - maximum 128 129 130 Moderator plate, bottom - volume average 120 121 121 Moderator plate, lid - maximum 102 103 103 Moderator plate, lid - volume average 95 95 95 Retention ring - maximum 94 94 94 Closure plate - maximum 114 115 115 Closure plate - volume average 113 113 114 4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-26

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 I Table 4.4-5 Component temperatures for NCS with fuel rod failure for thermal require-ments 1-3 (continued)

Temperature, °C Component - type of temperature Thermal requirement 1 2 3 Trunnion - maximum 109 110 110 1

Trunnion screws - maximum 118 Fuel channels - maximum 216 213 214 Basket sheets - maximum 212 211 212 Round segment - maximum 169 171 172 Outer sheets - maximum 176 178 178 Shielding element - maximum 174 177 177 Canister filling gas - volume average 176 179 179 Cask filling gas - volume average 107 108 108 Canister lid - maximum 116 117 117 Canister lid - volume average 109 110 110 Canister lid gasket- maximum 105 106 106 Canister lid screws 1 - maximum 106 Cask lid - maximum 91 92 92 Cask lid - volume average 90 90 90 Cask lid gasket - maximum 91 92 92 Cask lid screws 1 - maximum 100 Protection cap gasket - maximum 90 91 91 Pressure switch gasket - maximum 90 91 91 Tightening plug gasket - maximum 107 108 108 Protection cover - maximum 81 81 81 For the screw temperatures, the surface temperature of the corresponding component plus a safety margin is used.

4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-27

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 4.4-9 Temperature distribution in the cask for NCS with fuel rod failure for thermal requirement 3 4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-28

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 4.4-10 Temperature distribution in the hottest cross-section of the FA for NCS with fuel rod failure for thermal requirements 1-3 4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-29

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 4.4-11 Axial temperature profile on the cask lateral surface and the cask cavity sur-face for NCS with fuel rod failure for thermal requirement 3 4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-30

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Figure 4.4-12 Temperature distribution in the canister lid for NCS with fuel rod failure for thermal requirement 3 Figure 4.4-13 Temperature distribution in the cask lid for NCS with fuel rod failure for ther-mal requirement 3 4.4.4 Minimum Temperatures In this section, the temperatures of the storage cask are examined for a conservatively low ambient temperature of -40 °C and no insolation according to Section 4.4.1. The content heat power is set to the minimum admissible value of zero. Therefore, all components of cask and content have the ambient temperature of -40 °C.

4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-31

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 4.4.5 Maximum Internal Pressures The calculation of the MNOP and a discussion on the generation of flammable gases is document-ed in the containment evaluation in Chapter 7.

4.4.6 Maximum Thermal Stresses The occurrence of thermal stresses is minimized by sufficiently large gaps in axial and radial direc-tion, which allows for free thermal expansion of the different components without contact and re-straint. The discussion of thermal stresses due to temperature gradients within the components can be found in the structural evaluation in Chapter 3.

The dimensions of the cask components in the design drawings are related to an assembly tem-perature of 20 °C. In the following paragraphs, the radial and axial gap widths between cask and canister and between canister and basket are investigated under consideration of thermal expan-sion. For hot conditions, the gaps are smaller due to the higher thermal expansion of the inner components compared to the outer components. The results show that these gaps are sufficiently large to prevent contact and restraint forces for hot conditions.

4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-32

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Table 4.4-6 Thermal expansion for the outer and inner axial gap 4.4.7 Evaluation of Cask Performance for Normal Conditions of Storage It is demonstrated that the CASTOR geo69 storage cask fulfils all requirements for NCS with re-gard to thermal aspects. The following items summarize the results of the thermal design:

• The evaluation of the results in Section 4.4.3 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 demonstrated in Section 4.4.3 that additional safety margins exist because of the con-servative approaches for the thermal modelling.

• 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.

4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-33

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS

• 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.

• It is shown in Section 4.4.6 that the main gaps in axial and radial direction are sufficiently large to allow for free thermal expansion of the different components without contact and restraint.

• It is discussed in Section 4.4.2.2 that the heat removal capability is still ensured even under consideration of the most unfavourable manufacturing tolerances for the radial gaps be-tween cask and canister and between canister and basket and an additional extension of these gap widths by a factor of two.

• The evaluation of the maximum pressure and a discussion on the generation of gases is documented in the containment evaluation in Chapter 7.

• The influence of the calculated temperatures on the mechanical material properties and thermal stresses is evaluated in the structural evaluation in Chapter 3.

List of References

[1] ANSYS Release 17.2 UP20160718, © 2016 SAS IP Inc.

[2] VOi Heat Atlas Calculation Sheets for the Heat Transfer Springer, 2010

[3] J. P. Holman Heat Transfer McGraw Hill (2010)

[4] SCALE 4.4a: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation for Workstations and Personal Computers Oak Ridge National Laboratory NUREG/CR-0200 ORNL/RSIC, CCC-545, 2000

[5] F. J. Bayley, J. M. Owen, and A. B. Turner Heat Transfer William Clowes & Sons, Ltd., London, 1972 4.4 Thermal Evaluation for Normal Conditions of Storage Section 4.4, Rev. 1 Page 4.4-34

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 4.5 Thermal Evaluation for Off-Normal Conditions of Storage Name, Function Date Prepared 10.11.2022 Reviewed 10.11.2022 4.5 Thermal Evaluation for Off-Normal Conditions of Storage Section 4.5, Rev. 1 Page4.5-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 In the following sections the influence of off-normal conditions on the thermal design is discussed.

• 4.5.1 Loss of Power and Instrumentation Failures The CASTOR geo69 is a completely passive storage system. In particular, no active cooling sys-tem is required for heat removal. For that reason, loss of power and instrumentation failures are not relevant as off-normal conditions.

4.5.2 Blockage of Ventilation Openings .

The storage cask has no ventilation openings but is cooled passively at the outer surface. There-fore, a blockage of ventilation openings has not to be considered.

4.5.3 Off-Normal Environmental Conditions In order to take into account extreme seasonal variations of the environmental temperature at the storage facility site for off-normal conditions of storage, a maximum value of is ap-plied for the hot case. Conservatively, this peak value for the environmental temperature is used in the thermal calculations as a steady-state temperature. In spite of the high thermal inertia of cask and content, variations of the environmental temperature within several days and fluctuations dur-ing day and night are not taken into account.

The off-normal maximum environmental temperature o f - is used in conjunction with insolation according to 10 CFR 71 (§ 71. 71) with the solar heat fluxes listed in Section 4.4.1 averaged over a period of 24 h per day. These solar heat fluxes are applied in the thermal calculations as steady-state values. The absorption coefficient of surfaces exposed to insolation is conservatively set to 1.

In reality, the solar absorption coefficient for the painted cask surfaces amounts to 0.4 (light paint),

see Section 4.2.1.

A constant temperature of 10 °C (50 °F) in the ground beneath the storage pad base plate at a depth of 8 m according to Section 4.4.1 also applies for off-normal environmental conditions.

The investigation for an off-normal maximum environmental temperature o f - is combined with the assumption of a failure of 10 % of the fuel rods. The corresponding calculations are presented in the following Section 4.5.4.

The minimum environmental temperature of -40 °C taken into account for the cold case under NCS according to Section 4.4:1 is sufficiently low to consider off-normal seasonal variations,. Further-4.5 Thermal Evaluation for Off-Normal Conditions of Storage Section 4.5, Rev. 1 Page 4.5-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS more, a solar heat flux of zero is used for the cold case under NCS. The corresponding thermal analysis is presented in Section 4.4.4.

4.5.4 Off-Normal Fuel Rod Failure For off-normal conditions of storage, a failure of 10 % of the fuel rods in conjunction with a fraction-of fission gas release from the fuel rods of 0.15 is considered according to Section 4.2.2. Addition-ally, an off-normal maximum environmental temperature o f - according to Section 4.5.3 in con-junction with insolation is applied.

4.5.4.1 Thermal Model The same numerical model is used as described in Section 4.4.2 for NCS with the following excep-tions:

• The thermal properties of the canister filling gas mixture and the homogenized FA zones are adapted according to Table 4.2-10 and Table 4.2-11 under consideration of fuel rod failure. The mechanical analyses in Chapter 3 show that the canister remains leak-tight for off-normal conditions of storage. For that reason, no fission gas release from the canister interior to the cask cavity occurs.

• The ambient temperature for convection and thermal radiation is set to the off-normal value of - according to Section 4.5.3. This is respected in the boundary conditions of the nu-merical model accordingly as described in Section 4.4.2.8.

Thermal requirement 3 according to Section 4.1.2 is considered on!y, as it leads to slightly higher temperatures than thermal requirements 1 and 2.

4.5 Thermal Evaluation for Off-Normal Conditions of Storage Section 4.5, Rev. 1 Page 4.5-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 4.5.4.2 Maximum Temperatures The temperature distribution of the storage cask for o~-normal conditions with fuel rod failure and off-normal environmental temperature is shown in Figure 4.5-1. The resulting maximum tempera-tures of the storage cask are summarized in Table 4.5-1.

Below, the design-relevant temperatures are compared to their maximum admissible values ac-cording to Section 4.3: *

• The maximum temperature of the fuel rods is 250 °C and is therefore significantly lower than the maximum admissible temperature of 400 °C.

• The maximum temperature for the moderator rods is 133 °C, for the bottom moderator plate 145 °C and 117 °C for the lid moderator plate. Therefore, the maximum temperatures of all moderator material are far below the maximum admissible temperature o f - °C.

• The highest gasket temperature of 122 °C occurs in the tightening plug gasket. The maxi-mum admissible temperature for continuous operation of the gaskets is llllll'C. Therefore, all gasket temperatures are far below the temperature limit.

• Furthermore, temperature limits for structural components (e.g. fuel basket sheets) listed in Table 4.3-1, which are relevant for the mechanical integrity, are met.

The evaluation of the results shows that all calculated maximum temperatures of the cask compo-nents and the content are far below the maximum admissible values with large safety margins.

4.5 Thermal Evaluation for Off-Normal Conditions of Storage Section 4.5, Rev. 1 Page 4.5-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 4.5-1 Temperature distribution for off-normal conditions with fuel rod failure and off-normal environmental temperature 4.5 Thermal Evaluation for Off-Normal Conditions of Storage Section 4.5, Rev. 1 Page 4.5-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 4.5-1 Component temperatures for off-normal conditions with fuel rod failure and off-normal environmental temperature Component - type of temperature Temperature, °C Fuel rods - maximum 250 Cask lateral surface - maximum 119 Cask lateral surface - maximum circumferential average 117 Cask lateral surface - maximum longitudinal average 111 Cask cavity lateral surface - maximum 134 Cask cavity lateral surface - maximum circumferential average 133 Cask cavity bottom - maximum 148 Cask cavity bottom - area average 141 Closure plate, underside - maximum 130 Closure plate, underside - area average 128 Moderator rods, inner row (MR-i) - maximum 133 MR-i - maximum cross-sectional average, hottest rod 130 MR-i - volume average, hottest rod 119 Moderator rods, outer row (MR-o) - maximum 129 MR-o - maximum cross-sectional average, hottest rod 126 MR-o - volume average, hottest rod 116 Canister inner lateral surface - maximum 144 Canister inner lateral surface - max. circumferential average 143 Canister outer lateral surface - maximum 143 Canister outer lateral surface - max. circumferential average 142 Canister bottom - maximum 164 Moderator plate, bottom - maximum 145 Moderator plate, bottom - volume average 136 Moderator plate, lid - maximum 117 Moderator plate, lid - volume average 109 Retention ring - maximum 108 Closure plate - maximum 130 Closure plate - volume average 128 4.5 Thermal Evaluation for Off-Normal Conditions of Storage Section 4.5, Rev. 1 Page 4.5-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 4.5-1 Component temperatures for off-normal conditions with fuel rod failure and off-normal environmental temperature (continued)

Component - type of temperature Temperature, °C Trunnion - maximum 124 1

Trunnion screws - maximum 132 Fuel channels - maximum 236 Basket sheets - maximum 233 Round segment - maximum 190 Outer sheets - maximum 198 Shielding element - maximum 197 Canister filling gas - volume average 199 Cask filling gas - volume average 122 Canister lid - maximum 132 Canister lid - volume average 124 Canister lid gasket - maximum 120 Canister lid screws 1 - maximum 120 Cask lid - maximum 105 Cask lid - volume average 103 Cask lid gasket - maximum 105 Cask lid screws 1 - maximum 113 Protection cap gasket - maximum 104 Pressure switch gasket - maximum 104 Tightening plug gasket - maximum 122 Protection cover - maximum 94 For the screw temperatures, the surface temperature of the corresponding component plus a safety margin is used.

4.5.5 Maximum Internal Pressures The calculation of the maximum internal pressures under off-normal conditions of storage and a discussion on the generation of flammable gases is documented in the containment evaluation in Chapter 7. The storage cask is designed to withstand loads due to off-normal pressure what is shown in the structural evaluation in Chapter 3.

4.5.6 Maximum Thermal Stresses The occurrence of thermal stresses is minimized by sufficiently large gaps in axial and radial direc-tion, which allows for free thermal expansion of the different components without contact and re-4.5 Thermal Evaluation for Off-Normal Conditions of Storage Section 4.5, Rev. 1 Page 4.5-7

Non-Proprietary Version 1014-SR-00002 Rev. 1 Proprietary Information withheld per 10CFR 2.390

(@)GNS straint. The discussion of thermal stresses due to temperature gradients within the components can be found in the structural evaluation in Chapter 3.

4.5.7 Evaluation of Cask Performance for Off-Normal Conditions of Storage For off-normal conditions of storage, it is demonstrated that the CASTOR geo69 storage cask fulfils all requirements with regard to thermal aspects. The following items summarize the results of the ~hermal investigations:

• The evaluation of the results in Section 4.5.4.2 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 ensuring 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.

• The evaluation of the maximum pressure and a discussion on the generation of gases is documented in the containment evaluation in Chapter 7.

• The influence of the calculated temperatures on the mechanical material properties and thermal stresses is evaluated in the structural evaluation in Chapter 3.

4.5 Thermal Evaluation for Off-Normal Conditions of Storage Section 4.5, Rev. 1 Page 4.5-8

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 4.6 Thermal Evaluation for Accident Conditions of Storage Name, Function Date Signature Prepared 10.11.2022 Reviewed 10.11.2022 4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS In the scope of the them1al evaluation for ACS, the following conditions are discussed:

• ACS fire,

• ACS burial and

• ACS impact.

4.6.1 ACS Fire Under ACS fire, a fire accident during storage is considered. The evaluation under ACS fire com-prises an initial steady state, a transient fire phase, when the cask is exposed to a fully-engulfing pool fire with an average flame temperature of 800 °C, and a subsequent transient cooling phase.

For ACS fire, a failure of 100 % of the fuel rods in conjunction with a fraction of fission gas release from the fuel rods of 0.15 is considered according to Section 4.2.2.

4.6.1.1 Thermal Model The them1al analyses for ACS fire are based on the same calculation methods as described in Section 4.4.2 for NCS. The differing boundary conditions and assumptions under ACS fire are de-scribed in the following sections.

4.6.1.1.1 Geometric Modelling For ACS fire, the same geometrical model is used as for NCS (see Section 4.4.2.2 and Figure 4.4-1 to Figure 4.4-3).

For the transient calculation under ACS fire, the heat storage capacity of the fins, which are not modelled in the FE model explicitly, is taken into account.

I I

4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 4.6.1.1.2 Filling Gases The thermal properties of the canister filling gas mixture and the homogenized FA zones are ap-plied according to Table 4.2-10 and Table 4.2-11 under consideration of fuel rod failure. The adapted thermal properties are applied to the initial steady state as well as to the fire and cooling phase. The mechanical analyses in Chapter 3 show that the canister remains leak-tight for ACS.

For that reason, no fission gas release from the canister interior to the cask cavity occurs.

4.6.1.1.3 Initial State before the Fire Accident The initial conditions before the fire accident correspond to the steady state for NCS described in Section 4.4. The cask dissipates heat by natural convection and thermal radiation according to the boundary conditions described in Section 4.4.2.8.

For ACS fire, only thermal requirement 3 according to Section 4.1.2 is considered, as it leads to slightly higher temperatures than thermal requirements 1 and 2.

4.6.1.1.4 Boundary Conditions during the Fire Phase As fire accident for storage, the spillage and ignition of 200 I liquid fuel of a transporter vehicle is considered. The following fire conditions are applied according to 10 CFR 71 (§ 71.73(c)(4)):

• The cask is exposed to a fully-engulfing pool fire

• with a constant flame temperature of 800 °C,

• an emissivity of the fire of 0.9 and

• an emissivity of sooted surfaces of 0.8.

• The fuel source extends horizontally at least 1 m but not more than 3 m beyond the external surface of the cask.

The heat impact by the fire takes place at the external surface of the cask and the upper side of the storage pad base plate by natural convection and thermal radiation. The cask is assumed to still standing in a vertical position so that there is no direct impact of the fire at the bottom-side front face.

An effective heat transfer coefficient hett is defined on the surfaces exposed to the fire which con-sists of a convective portion hcanv and a radiative portion hrad:

heff = hconv + hrad 4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS For the ambient temperature for convection T A,conv and for radiation T A,rad, the flame temperature of 800 °C is applied.

For the fire phase, a conservatively high heat transfer coefficient for forced convection hconv is de-rived from the following Nusselt law according to [1]:

The Nusselt number Nu is defined as:

Nu= h

• D / k with: h heat transfer coefficient, W/(m 2 *K)

D cask diameter, m k thermal conductivity, W/(m*K)

The Prandtl number Pr is defined as:

Pr= v / a with: a - thermal diffusivity, m 2/s v - kinematic viscosity, m 2/s Re denotes the Reynolds number:

Re= v

• D / v with: v - velocity of the fluid, m/s D - cask diameter, m v - kinematic viscosity, m 2/s The parameters C and m depend on the Reynolds number Re. The heat transfer coefficient is cal-culated for the flame temperature of 800 °C and additionally for a sufficiently low temperature of 400°c.

The gas velocity inside the fire can be estimated by - according to [2]. For the kinematic vis-cosity v, the Prandtl Number Pr and the thermal conductivity k, the temperature dependent materi-al properties of dry air from Table 4.4-4 according to [3] are used. The diameter of the cask amounts to 2.54 m. In Table 4.6-1, the input values and the results for the Reynolds number Re, the Nusselt number Nu and the heat transfer coefficient hconv during the fire phase are summarized.

4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 4.6-1 Input values and results for the heat transfer coefficient during the fire phase Input values Unit Literature 800 °C 400 °C 2 7 V m /s [2] 1396. 7 . 10- 643.5. 10-7 Pr - [2] 0.7333 0.7081 k W/(m*K) [2] 71.35

• 10-3 50.24

• 10-3 V mis [2] 5.0 5.0 D m 2.54 2.54 C - [1] 0.027 0.027 m - [1] 0.805 0.805 Calculated values Re - 90929 197358 Nu - 239 441 2

hconv W/(m *K) 6.71 8.72 For the flame temperature of 800 °C, a heat transfer coefficient for forced convection hconv of is calculated. For the lower temperature of 400 °C, the heat transfer coefficient amounts to J ) due to the higher Reynolds number. Conservatively, a considerably in-creased convective heat transfer coefficient hconv of is applied during the fire phase.

The increased heat transfer coefficient of takes into account that the gas atmosphere under pool fire conditions is a complex composition of mainly air, CO 2 , water vapour and additional components such as soot particles. The exact gas mixture is varying locally and over time. As the real material properties of the gas cannot be evaluated, the heat transfer coefficient is calculated using the material data of air, which represents the major part of the flue gas.

The convective heat transfer coefficient is applied on the surface exposed to the fire and set con-stant during the fire phase. Within the fin zone of the cask surface, an effective surface enlarge-ment factor - is applied, which is explained in Section 4.4.2.3. The convective heat transfer coefficient hconv is multiplied by this factor.

The radiative portion hrad of the effective heat transfer coefficient is calculated using the Stefan Boltzmann law according to [3]:

4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS For the calculation of the effective emission coefficient, the fire is treated as a cylinder surrounding the cask at a distance of - from the external surface of the cask. According to [3], the effective emissivity between two concentric cylinders is given by:

The described boundary conditions lead to a radiative heat flux and a total heat flux including con-vection shown in Figure 4.6-1. The effective surface enlargement factor of the fin zone is not in-cluded. The heat fluxes are in the range recommended in [2].

Figure 4.6-1 Heat flux at the cask surface during fire phase For the fire accident, the spillage and ignition of 200 I of transporter fuel is assumed conservatively.

The fire duration is determined from the fuel volume of 200 I, which is assumed to be spread around the cask in an annular spillage. It is assumed that the liquid fuel has the minimum horizon-tal extension of 1 m beyond the external surface of the cask. Using the minimum ring width yields a deeper pool for a fixed quantity of liquid fuel and thereby a conservatively long fire duration. This leads to an outer diameter of the fuel ring of 2.54 m + 2

• 1 m = 4.54 m and to a maximum fuel depth of 18 mm. The fuel consumption rate is assumed to be 0.15 in/min ( corresponding to 3.8 mm/min), which is the minimum fuel consumption rate given in the Sandia report [4] on large 4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS pool fire testing. This leads to a fire duration of 4.7 min. Conservatively, a fire duration of 5 min is assumed for the calculations.

Conservatively, insolation according to Section 4.4.1 is considered during the fire phase, too. The solar heat fluxes are applied as described for the boundary conditions under NCS in Section 4.4.2.8.

4.6.1.1.5 Boundary Conditions during the Cooling Phase Due to the high thermal inertia of cask and content, the maximum temperatures inside the cask are reached several hours after the end of the fire phase. In the calculations, the cooling phase com-prises a sufficiently long time period of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> so that all temperatures of the components of cask and content reach their peak values and decrease again afterwards.

Basically, the boundary conditions during the cooling phase are similar to those for NCS (see Sec-tion 4.4.2.8). Differing boundary conditions are described below.

At the external surface of the cask and the upper side of the base plate, an effective heat transfer coefficient is applied, which consists of a convective portion hconv and a radiative portion hrad:

heff = hconv + hrad The convective portion hconv is calculated using the following Nusselt law according to [5] for turbu-lent heat transfer by free convection at vertical planes and vertical cylinders:

The Nusselt number Nu, the Grashot number Gr and the Prandtl number Pr are defined in Section 4.4.2.8.

For the first 12 h of the cooling phase, heat removal by thermal radiation is neglected completely to take into account conservatively that the neighbouring ca_sks and the base plate of the storage pad are heated up to same temperature level as the cask itself:

hrad = 0 Afterwards, the radiative portion hrad is calculated using the Stefan Boltzmann law according to [3]:

4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 4.6.1.2 Maximum Temperatures

- The basic calculations are carried out with the protection cover mounted at the top of the cask.

Within comparative calculations the case without protection cover is considered. Table 4.6-2 lists the maximum temperatures and their time of appearance (t = 0: beginning of fire) for various com-ponents of cask and content for the case with mounted protection cover.

Figure 4.6-2 shows the temperature distribution in the cask at the end of the 5 min fire phase when the maximum cask surface temperature is reached. Figure 4.6-3 shows the temperature distribu-tion in the cask 23 h after the beginning of the fire when the hottest fuel rod reaches its maximum temperature. Figure 4.6-4 shows the temperature distribution of the lid system for the times when maximum temperatures of the cask and the canister lid gaskets are reached. Figure 4.6-5 to Figure 4.6-7 show the temperature courses over time for various components of cask and content.

Below, the design-relevant temperatures are compared to their maximum admissible values ac-cording to Section 4.3:

• The hottest fuel rod reaches after 38 hits maximum temperature of 281 °C, which is signifi-cantly lower than the maximum admissible fuel rod temperature for ACS of 570 °C valid for intact fuel rods.

• The temperatures of the gaskets are between 110 °C and 125 °C, 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 show that all calculated maximum temperatures of the cask compo-nents and the content are far below the maximum admissible values with large safety margins.

4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-8

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 4.6-3 shows the results for the case without mounted protection cover. Compared to the re-sults for the case with protection cover in Table 4.6-2, the maximum temperatures of the gaskets in the cask lid are by 42-57 K higher because of the direct impact of the fire whereas the maximum temperatures of the gaskets in the canister lid remain approximately the same (f1 Ts 2 K).

The temperatures for the post-fire steady-state conditions are similar to the temperatures for the results of the initial state. Exemplarily, the post-fire steady-state temperature for the cask surface at half height amounts to approximately 100 °C, which is 3 K above the temperature of the results for the initial state (see Figure 4.6-7). This is mainly due to the fact that the emissivity after the fire of the sooted cask surface (111111) is lower than for the initial state (1111111). All other component tempera-tures shift accordingly.

4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-9

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 I Table 4.6-2 Maximum component temperatures for ACS fire with fuel rod failure with pro-tection cover and time of appearance after beginning of the fire Maximum Time of Component - type of temperature temperature, °C appearance, h Fuel rods - maximum 281 38.0 Cask lateral surface - half height 218 0.1 Cask cavity lateral surface - half height 123 2.8 Moderator rods, inner row (MR-i) - maximum 135 5.5 MR-i - maximum cross-sectional average, hottest rod 133 4.0 MR-i - volume average, hottest rod 122 2.8 Moderator rods, outer row (MR-o)...:. maximum 151 0.1 MR-o - maximum cross-sectional average, hottest rod 132 2.8 MR-o - volume average, hottest rod 121 1.7 Moderator plate, bottom - maximum 147 17.0 Moderator plate, bottom - volume average 135 15.0 Moderator plate, lid - maximum 121 22.0 Moderator plate, lid - volume average 114 14.0 Canister wall - maximum 145 7.0 Basket sheets - maximum 256 38.0 Shielding elements - maximum 215 30.0 Canister filling gas - volume average 220 31.0 Cask filling gas - volume average 122 5.5 Canister lid gasket - maximum 121 21.0 Cask lid gasket - maximum 111 4.5 Protection cap gasket - maximum 110 9.0 Pressure switch gasket - maximum 110 8.0 Tightening plug gasket - maximum 125 29.0 4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-10

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Figure 4.6-2 Temperature distribution at the time of the end of the fire.for ACS fire with fuel rod failure 4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-11

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 4.6-3 Temperature distribution in the cask at the time of maximum fuel rod tempera-ture for ACS fire with fuel rod failure 4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-12

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 4.6-4 Temperature distribution in the lid system at the time of maximum gasket temperatures for ACS fire with fuel rod failure 4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-13

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 4.6-5 Temperatures of the hottest fuel rod and the filling gases over time for ACS fire with fuel rod failure Figure 4.6-6 Temperatures of the lid gaskets over time for ACS fire with fuel rod failure 4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-14

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 G S Figure 4.6-7 Temperatures on the surfaces of cask, cavity and canister at half cask height over time for ACS fire with fuel rod failure 4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-15

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS I Table 4.6-3 Maximum component temperatures for ACS fire with fuel rod failure without protection cover and time of appearance after beginning of the fire Maximum Time of Component - type of temperature temperature, °C appearance, h Fuel rods - maximum 281 34.0 Cask lateral surface - half height 218 0.1 Cask cavity lateral surface - half height 123 2.8 Moderator rods, inner row (MR-i) - maximum 135 5.5 MR-i - maximum cross-sectional average, hottest rod 133 4.0 MR-i - volume average, hottest rod 123 3.0 Moderator rods, outer row (MR-a) - maximum 151 0.1 MR-a - maximum cross-sectional average, hottest rod 132 2.8 MR-a - volume average, hottest rod 121 1.7 Moderator plate, bottom - maximum 147 17.0 Moderator plate, bottom - volume average 135 15.0 Moderator plate, lid - maximum 134 0.7 Moderator plate, lid - volume average 124 3.0 Canister wall - maximum 145 7.0 Basket sheets - maximum 256 34.0 Shielding elements - maximum 215 27.0 Canister filling gas - volume average 220 26.0 Cask filling gas - volume average 125 3.5 Canister lid gasket - maximum 123 12.0 Cask lid gasket - maximum 154 0.2 Protection cap gasket- maximum 152 0.2 Pressure switch gasket - maximum 167 0.1 Tightening plug gasket - maximum 126 15.0 4.6.2 ACS Burial Under ACS burial, a complete burial of the storage cask for example by debris or sludge is consid-ered. The evaluation under ACS burial comprises an initial steady state and a subsequent transient burial phase, when the heat removal at the entire outer cask surface conservatively is completely eliminated. For ACS burial, a failure of 100 % of the fuel rods in conjunction with a fraction of fis-sion gas release from the fuel rods of 0.15 is assumed according to Section 4.2.2.

4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-16

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 4.6.2.1 Thermal Model The thermal analyses for ACS burial are based on the same calculation methods as described in Section 4.4.2 for NCS. The differing boundary conditions and assumptions under ACS burial are described in the following sections.

4.6.2.1.1 Geometric Modelling For ACS burial, the same geometrical model is used as for NCS see Section 4.4.2.2 and Figure 4.4-1 to Figure 4.4-3).

4.6.2.1.2 Filling Gases The thermal properties of the canister filling gas mixture and the homogenized FA zones are ap-plied according to Table 4.2-10 and Table 4.2-11 under consideration of fuel rod failure. The adapted thermal properties are applied to the initial steady state as well as to the burial phase. The mechanical analyses in Chapter 3 show that the canister remains leak-tight for ACS. For that rea-son, no fission gas release from the canister interior to the cask cavity occurs.

4.6.2.1.3 Initial State before the Burial Accident The initial conditions before the burial accident correspond to the steady state for NCS described in Section 4.4. The cask dissipates heat by natural convection and thermal radiation according to the boundary conditions described in Section 4.4.2.8.

For ACS burial, only thermal requirement 3 according to Section 4.1.2 is conside_red, as it leads to slightly higher temperatures than thermal requirements 1 and 2.

4.6.2.1.4 Boundary Conditions during the Burial Phase Under ACS burial, a complete burial of the storage cask for example by debris or sludge is consid-ered. As a hypothetic limit case, it is conservatively assumed that the heat removal is completely eliminated at the entire outer cask surface. In the numerical model, the entire outer cask surface is set adiabatic.

4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-17

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 In reality, the burial is only partly. Additionally, at the buried parts of the cask outer surface still a reduced heat removal takes place. The heat is transferred by conduction through the sludge or the particles of the debris and by convection and thermal radiation in cavities.

4.6.2.2 Maximum Temperatures As a result of the calculations for ACS burial, Figure 4.6-8 and Figure 4.6-9 show the time curves for the design-relevant temperatures of the hottest fuel rod and the lid gaskets. Because of the high thermal inertia of cask and content, during the burial the temperatures only rise very slowly at a nearly constant rate of about For the design-relevant components, Table 4.6-4 lists the times after the beginning of the burial until the temperatures reach the individual maximum admissible values for ACS according to Sec-tion 4.3. Below, the time limits for countermeasures are summarized:

• The maximum temperature of the hottest fuel rod reaches after its maximum ad-missible value for ACS of 570 °C valid for intact fuel rods.

• The temperatures of the canister lid gaskets first reach the lowest maximum admissible value of

• The temperatures of the cask lid gaskets first reach the lowest maximum admissible value of The evaluation of the results for ACS burial show that sufficient time for countermeasures remains until the maximum admissible temperatures of the design-relevant components of cask and con-tent are reached. In reality, even a considerably longer time for countermeasures is available be-cause of the very pessimistic assumption of a completely adiabatic outer cask surface (see Section 4.6.2.1.4 ).

4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-18

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Table 4.6-4 Time until reaching the maximum admissible temperatures for ACS burial with fuel rod failure Maximum admissible Time until reaching the Component temperature for ACS maximum admissible according to Section 4.3 temperature Fuel rods Canister lid gasket Tightening plug gasket Cask lid gasket Protection cap gasket Pressure switch gasket Figure 4.6-8 Temperature of the hottest fuel rod over time for ACS burial with fuel rod fail-ure 4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-19

Non-Proprietary Version 1014-SR-00002 Proprietary information withheld per 10CFR 2.390 Rev. 1 Figure 4.6-9 Temperatures of the lid gaskets over time for ACS burial with fuel rod failure 4.6.3 ACS Impact The fuel rod failure for ACS 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 cask temperatures due to the high thermal inertia of cask and content. For ACS impact, two scenarios - without and with release of fuel particles from the fuel rods - are consid-ered:

Scenario I:

The temperatures for the ACS impact are calculated without release of fuel particles for a failure of 100 % of the fuel rods in conjunction with a fraction of fission gas release from the fuel rods of 0.35 is considered according to Section 4.2.2.

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 4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-20

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 particle packing within gaps, a typical value for irregular particle sizes of * - is as-sumed.

Conservatively, the fuel particle packing and thereby the heat power is completely concen-trated in the lid-side region of the canister interior in order to calculate maximum tempera-tures for the design-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 ac-tive length), where the total heat power of about 18.5 kW (18.43 kW, thermal requirement 3) is dissipated in case of the hypothetical scenario II, see Section 4.6.3.1.

4.6.3.1 Thermal Model For scenario I of ACS impact, the same numerical model is used as described in Section 4.4.2 for NCS with the exception that the thermal properties of the canister filling gas mixture and the ho-mogenized FA zones are adapted according to Table 4.2-10 and Table 4.2-11 under consideration of fuel rod failure. The mechanical analyses in Chapter 3 show that the canister remains leak-tight for ACS. For that reason, no fission gas release from the canister interior to the cask cavity occurs.

Thermal requirement 3 according to Section 4.1.2 is considered only, as it leads to slightly higher temperatures than thermal requirements 1 and 2.

For scenario II, 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-fuel particles and - gas mixture. The lid-side parts of the fuel rods are still filled with fuel without considering porosity. Fig-ure 4.6-10 shows the 3D FE model for scenario II with fuel reconfiguration. For better visibility, only the 3D FE model of the canister is shown.

4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-21

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Figure 4.6-10 3D FE model of canister for ACS 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 18430 W / 1.33 m3 = 13865 W/m 3 . The heat power density in the fuel particle packing amounts t The fractions of helium and fuel pellets in the homogenized active FA zones are The former helium fraction of - is now filled with the fuel particle packing. Table 4.6-5 shows the distribution of the heat power in all fuel filled zones. In the canister model (see Figure 4.6-10),

the chosen axial height of 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 leads to a higher total heat power o f * *

• in the FE model (see Table 4.6-5) compared to the maximum value of 18430 W according to Section 4.1.2.

4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1' Page 4.6-22

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 4.6-5 Heat power zones in lid-side region of the canister interior for ACS 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 - i s evaluated according to the model of I [3] for packed beds. The following assumptions are made:

1. Temperatures of the fuel particle packing: 100 °C, 200 °C and 300 °Care 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 4.6-6. The calculations show that particle diameters in the range of 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 4.6-7).

For the former active FA zone outside the height of , the thermal properties of the inactive FA zone without fuel are applied. For the reduced active zone of with fuel pellets, fuel packing and cladding, a constant thermal conductivity of I is conservatively chosen for the radial direction in active and inactive FA zones (see Table 4.6-7). 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.

4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-23

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Table 4.6-6 Calculation of effective thermal conductivity in fuel filled gaps for ACS impact with fuel rod failure - scenario II with fuel reconfiguration The axial thermal conductivity values in Table 4.2-11 for ACS impact are slightly increased by the higher thermal conductivity of fuel filled gap zones (II* Table 4.6-6) compared to gas filled zones (0.038 W/(m*K) at 200 °C, see Table 4.2-10). The former free gas volume in FA zones amounts to - which is now filled with the fuel particle packing.

4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-24

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 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 4.6-7.

Table 4.6-7 Thermal conductivity in fuel filled gaps and FA zones for ACS impact with fuel rod failure - scenario II with fuel reconfiguration 4.6.3.2 Maximum Temperatures For ACS impact with fuel rod failure, the temperature distributions of the cask are shown in Figure 4.6-11 and Figure 4.6-12 for scenario I (without fuel reconfiguration) and scenario II (with fuel re-configuration). Table 4.6-8 and Table 4.6-9 list the maximum temperatures for various components of cask and content.

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 zon*es 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

• I- The free gas volume outside the height 01*1 without any fuel amounts to

• I- The additional gas volume of the empty parts of the fuel rods (outside the height of II) is conservatively not taken into account to get a maximum average gas temperature and pressure in the canister in 4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-25

Non-Proprietary Version 1014-SR-00002 Rev. 1 Proprietary Information withheld per 10CFR 2.390

(@)G S 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 4.3:

• The highest fuel rod temperatures are 298 °C (scenario I} and 327 °C (scenario II), which is far below the maximum admissible fuel rod temperature for ACS of 570 °C valid for intact fuel rods.

• The temperatures of the gaskets are between 92 °C and 117 °C (scenario I) and between 117 °C and 174 °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 ACS impact scenario I and scenario II shows that the calculated maximum temperatures of cask components and content are far below the maximum admissible values with large safety margins.

Figure 4.6-11 Temperature distribution for ACS impact with fuel rod failure - scenario I without fuel reconfiguration 4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-26

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Table 4.6-8 Maximum component temperatures for ACS impact with fuel rod failure - sce-nario I without fuel reconfiguration Component - type of temperature Temperature, 0 c

Fuel rods - maximum 298 Cask lateral surface - maximum 108 Cask lateral surface - maximum circumferential average 105 Cask lateral surface - maximum longitudinal average 97 Cask cavity lateral surface - maximum 125 Cask cavity lateral surface - maximum circumferential average 125 Cask cavity bottom - maximum 145 Cask cavity bottom - area average 136 Closure plate, underside - maximum 122 Closure plate, underside - area average 119 Moderator rods, inner row (MR-i) - maximum 125 MR-i - maximum cross-sectional average, hottest rod 122 MR-i - volume average, hottest rod 105 Moderator rods, outer row (MR-o)- maximum 120 MR-o - maximum cross-sectional average, hottest rod 117 MR-o - volume average, hottest rod 102 Canister inner lateral surface - maximum 137 Canister inner lateral surface - max. circumferential average 135 Canister outer lateral surface - maximum 135 Canister outer lateral surface - max. circumferential average 134' Canister bottom - maximum 165 Moderator plate, bottom - maximum 141 Moderator plate, bottom - volume average 129 Moderator plate, lid - maximum 112 Moderator plate, lid - volume average 100 Retention ring - maximum 96 Closure plate - maximum 122 Closure plate - volume average 120 4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-27

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 4.6-8 Maximum component temperatures for ACS impact with fuel rod failure - sce-nario I without fuel reconfiguration (continued)

Component - type of temperature Temperature, °C Trunnion - maximum 115 1

Trunnion screws - maximum 123 Fuel channels - maximum 280 Basket sheets - maximum 272 Round segment - maximum 209 Outer sheets - maximum 222 Shielding element - maximum 225 Canister filling gas - volume average 233 Cask filling gas - volume average 110 Canister lid - maximum 131 Canister lid - volume average 119 Canister lid gasket - maximum 111 Canister lid screws 1 - maximum 111 Cask lid - maximum 93 Cask lid - volume average 92 Cask lid gasket - maximum 92 Cask lid screws 1 - maximum 100 Protection cap gasket - maximum 92 Pressure switch gasket - maximum 92 Tightening plug gasket - maximum 117 Protection cover - maximum 81 1

For the screw temperatures, the surface temperature of the corresponding component plus a safety margin is used.

4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-28

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Figure 4.6-12 Temperature distribution for ACS impact with fuel rod failure - scenario II with fuel reconfiguration 4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-29

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390

• Rev. 1 @GNS Table 4.6-9 Maximum component temperatures for ACS impact with fuel rod failure - sce-nario II with fuel reconfiguration Component - type of temperature Temperature, °C Fuel rods - maximum 327 Cask lateral surface - maximum 119 Cask lateral surface - maximum circumferential average 118 Cask lateral surface - maximum longitudinal average 96 Cask cavity lateral surface - maximum 131 Cask cavity lateral surface - maximum circumferential average 130 Cask cavity bottom - maximum 101 Cask cavity bottom - area average 98 Closure plate, underside - maximum 93 Closure plate, underside - area average 93 Moderator rods, inner row (MR-i) - maximum 130 MR-i - maximum cross-sectional average, hottest rod 126 MR-i - volume average, hottest rod 101 Moderator rods, outer row (MR-o) - maximum 124 MR-o '- maximum cross-sectional average, hottest rod 120 MR-o - volume average, hottest rod 98 Canister inner lateral surface - maximum 166 Canister inner lateral surface - max. circumferential average 164 Canister outer lateral surface - maximum 164 Canister outer lateral surface - max. circumferential average 162 Canister bottom - maximum 109 Moderator plate, bottom - maximum 100 Moderator plate, bottom - volume average 96 Moderator plate, lid - maximum 172 Moderator plate, lid - volume average 137 Retention ring - maximum 128 Closure plate - maximum 94 Closure plate - volume average 93 4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-30

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Table 4.6-9 Maximum component temperatures for ACS impact with fuel rod failure - sce-nario II with fuel reconfiguration (continued)

Component - type of temperature Temperature, °C Trunnion - maximum 117 1

Trunnion screws - maximum 125 Fuel channels - maximum 308 Basket sheets - maximum 292 Round segment - maximum 212 Outer sheets - maximum 225 Shielding element - maximum 217 Canister filling gas - volume average 167 Cask filling gas - volume average 130 Canister lid - maximum 212 Canister lid - volume average 181 Canister lid gasket - maximum 162 Canister lid screws 1 - maximum 162 Cask lid - maximum 120 Cask lid - volume average 117 Cask lid gasket - maximum 119 1

Cask lid screws - maximum 127 Protection cap gasket - maximum 117 Pressure switch gasket - maximum 117 Tightening plug gasket - maximum 174 Protection cover - maximum 94 1

For the screw temperatures, the surface temperature of the corresponding component plus a safety margin is used.

4.6.4 Maximum Internal Pressures The calculation of the maximum internal pressures under ACS and a discussion on the generation of flammable gases is documented in the containment evaluation in Chapter 7.

4.6.5 Maximum Thermal Stresses The occurrence of thermal stresses is minimized by sufficiently large gaps in axial and radial direc-tion, which allows for free thermal expansion of the different components without contact and re-4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-31

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS straint. The discussion of thermal stresses due to temperature gradients within the components can be found in the structural evaluation in Chapter 3.

4.6.6 Evaluation of Cask Performance for Accident Conditions of Storage It is demonstrated that the CASTOR geo69 storage cask fulfils all requirements under ACS with regard to thermal aspects. The following items summarize the results of the thermal investigations for ACS fire and ACS impact:

• The evaluation of the results in Sections 4.6.1.2 and 4.6.3.2 show that all calculated maxi-mum temperatures of the 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.

• 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 shown in Section 4.6.5 that the main gaps in axial and radial direction are sufficiently large to allow for free thermal expansion of the different components without contact and restraint.

• The evaluation of the maximum pressure and a discussion on the generation of gases is

. documented in the containment evaluation in Chapter 7.

• The influence of the calculated temperatures on the mechanical material properties and thermal stresses is evaluated in the structural evaluation in Chapter 3.

The evaluation of the results for ACS burial in Section 4.6.2.2 show that sufficient time for coun-termeasures remains until the maximum admissible temperatures- of the design-relevant compo-nents of cask and content are reached.

4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-32

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS List of References

[1] Theodore L. Bergman Fundamentals of Heat and Mass Transfer (7 th edition) 2011

[2] A Guide For Thermal Testing Transport Packages For Radioactive Material

- Hypothetical Accident Conditions -

Division of Transportation and Packaging Safety, Office of Risk Analysis and Technology 1993

[3] VDI Heat Atlas Calculation Sheets for the Heat Transfer Springer, 2010 *

[4] Gregory, J. J., et. al.

Thermal Measurements in a Series of Large Pool Fires SAND85-1096, Sandia National Laboratories, Albuquerque, NM, August 1987

[5] W. H. McAdams Heat Transmission McGraw-Hill, New York, 3rd Edition, 1985 4.6 Thermal Evaluation for Accident Conditions of Storage Section 4.6, Rev. 1 Page 4.6-33

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 4.7 Thermal Evaluation for Short-Term Operations Name, Function Date Prepared 10.11.2022 Reviewed 10.11.2022 4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-1

Non-Proprietary Version 1014-SR-00002 Rev. 1 Proprietary Information withheld per 10CFR 2.390 s

In this section, short-term operations inside the reactor facility are investigated. The different op-eration phases comprise the fuel loading of the canister inside the transfer cask under water, the dewatering, vacuum drying and helium backfilling of the canister interior as well as the transfer of the transfer cask inside the reactor building.

After loading the canister into the storage cask, the storage cask is transferred from the reactor building on-site to the storage pad. Thermal investigations of the on-site transfer comprise normal conditions, off-normal conditions, accident conditions fire and accident conditions impact.

The short-term operations are described in Section 1.2 and Chapter 9 in detail.

4. 7.1 Thermal Properties of Materials The relevant material data of the components of the cask used in the thermal calculations are

• the thermal conductivity k, W/(m*K),

• the density p, kg/m 3 ,

• the specific heat capacity c, J/(kg*K) and

• the emissivity E.

The following tables contain the material data of components of the transfer cask. The values are in accordance with Section II, Part D (Metric) of the BPVC [1] and are listed in Table 4.7-1 and Ta-ble 4.7-2. The material data of the canister, the fuel basket and the content are the same as de-scribed in Section 4.2.1. Table 4.7-3 contains the thermal properties of helium, air and water va-pour which are considered in the calculations as gas atmospheres in the interior of the transfer cask and canister.

Table 4.7-3 also contains the thermal properties of liquid water filled in in the inner and outer water chamber in the transfer cask wall. To take into account the enhanced heat transfer by free convec-tion, an effective heat conductivity kett is defined for the water in the chambers:

Liquid water has a high heat transfer coefficient for free convection of I 4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 G S Table 4.7-1 Material data of the transfer cask Tempera- Heat con- Specific heat Component Density, Reference ture, ductivity, capacity, (Material) oc kg/m 3 W/(m*K) J/(kg*K)

Liner, 20 14.1 492 bottom ring, 100 15.4 511 enclosure lead shield, 150 16.1 519 enclosure inner/outer 200 16.8 526 Chapter 8 8030 water chamber, 250 17.6 533 lid 300 18.3 540 350 19.0 546 (SA-240M 316L) 400 19.7 553 Head ring, bottom lid Chapter 8 See SA-240M 316L (SA-182M FXM-19) 0 35.5 -

25 - 129 127 - 132 Lead shield 150 33.0 -

(DIN EN 12659, Chapter 8 200 - 11340 -

PB940R) 227 - 137 250 31.8 -

300 - -

326 31.0 142 I Table 4.7-2 Surface emissivities of the transfer cask 4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-3

\.

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Table 4.7-3 Material data of fluids Tempera- Heat con- Specific heat Density, Fluid Reference ture, ductivity, capacity, oc kg/m3 W/(m*K) J/{kg*K) 25 0.1536 100 0.1793 3115 Helium [3] 200 0.2116 0.166 300 0.2420 (p =canst.)

400. 0.2708 20 0.0257 100 0.0314 720 Air [3] 200 0.0380 1.188 300 0.0441 (p = canst.)

400 0.0500 i 25 0.0186 100 0.0251 0.0046 1870 Water vapour [3] 200 0.0332 300 0.0433 (0.01 bar) (p =canst.)

400 0.0555 20 998 4185 Liquid water (in inner/outer water chamber)

[3]

30 40 50 60 70 80 (inner/outer, including free convection) 996 992 988 983 978 972 4180 4179 4180 4183 4188 4196 90 965 4205 100 959 4216 4.7.2 Fuel Loading under Water In the operation phase considered here, the transfer cask is loaded with fuel in the pool under wa-ter. The interior of the transfer cask and canister is filled with water and the lid of the canister is not yet mounted. This allows for a free interchange of water by convection between the interior of the canister and the pool. For the pool water a conservatively high temperature of 50 °C is assumed.

The temperature of the water in the interior of the transfer cask has almost the pool water tempera-ture of 50 °C because of the free interchange by convection with the pool water. The temperatures of the transfer cask components also have almost the pool water temperature of 50 °C because of the good heat transfer conditions at the surfaces bordering on water. Liquid water has a high heat

' transfer coefficient for free convection of *** according to [2].

4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Consecutively, the cladding temperatures of the FA are estimated. The surface of the fuel rods of a single FA of type GE 8x8-1 in the area of the active zone amounts to 63

• 3.708 m

• n

• 12.52 mm=

9.2 m2 (see Table 4.1-1). With a maximum heat power of 450 W per FA, the heat power density amounts to 49 W/m 2

• With a conservatively low heat transfer coefficient for free convection in water of , the temperature difference between the fuel rod surface and the pool water is llT

=

Conservatively, a constant temperature of 50 °C is chosen for the components of the water-filled transfer cask and the content for the time under water until the canister lid is mounted.

As long as the water-filled transfer cask is under water and the canister lid is not yet mo,unted, no temporal restrictions are required for this operation phase.

4.7.3 Water-Filled Cask In the operation phase considered here, the interior of the transfer cask is filled with water and the lid of the canister is mounted. This operation phase starts with mounting the canister lid under wa-ter and ends with the beginning of the dewatering. As estimated in the previous Section 4.7.2, the components of the transfer cask and the content have a constant temperature of 5_0 °C at the be-ginning of the operation phase when the canister lid is mounted.

After mounting the canister lid, the components of the transfer cask and the content continuously heat up because the interchange of water by convection between the interior of the canister and the pool is interrupted. Furthermore, the heat removal at the outer surface of the transfer cask de-grades after partly lifting out of the pool water because the heat is now partly transferred to air with

- a significantly lower convective heat transfer coefficient than water.

During heating, the heat power of the FA is stored partly in the transfer cask. Furthermore, a por-tion of the heat power is dissipated over the transfer cask surface which conservatively is neglect-ed completely.

With the following analytical calculation the duration within the transfer cask heats up from 50 °C to 100 °C, which is the boiling temperature of water at a pressure of 1 bar, is estimated. It is assumed that all components of the transfer cask and the content heat up uniformly because of the good heat transfer conditions inside the water-filled transfer cask. For the heat power the maximum val-ue of 18.5 kW is considered.

4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The time period is calculated using the following equation for the internal energy according to [3]:

fit = I(m;

• Cp,i) * (TE - Ts) / 0 with: m; - massofthecomponenti(seeTable4.7-4),kg Cp,i - specific heat capacity of the component i (see Table 4.7-4), J/(kg*K)

Ts - start temperature of the transfer cask and content, Ts = 50 °C TE - end temperature of the transfer cask and content, TE = 100 °C Q - heat power of the content, Q = 18500 W The time period until the boiling temperature of 100 °C at a pressure of 1 bar is reached inside the transfer cask amounts to fit = - In reality, it will take longer because the portion of the heat power dissipated over the transfer cask surface is neglected completely, although most of the transfer cask surface is still under water.

For the operation phase with the water-filled transfer cask, the time period after mounting of the canister lid until the beginning of the dewatering has to be restricted to -

Table 4.7-4 Heat capacity of the components of transfer cask and content 4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @)GNS 4.7.4 Vacuum Drying The vacuum drying follows the dewatering of the canister. During the vacuum drying, the interior of the canister is filled with pure water vapour. Because of the low thermal conductivity of water va-pour in comparison to liquid water or helium, the most unfavourable conditions for heat removal of all operation phases exist during vacuum drying. Conservatively, it is assumed that the space be-tween canister and transfer cask is dewatered at the same time and is afterwards also filled with water vapour.

In reality, the time period of vacuum drying amounts to several hours. Conservatively, in the follow-ing investigations the final steady state after heating up is considered, which corresponds to an unlimited time period of vacuum drying.

4.7.4.1 Thermal Model The calculation methods for short-term operations inside the reactor building in principle corre-spond to the approach for NCS described in Section 4.4.2. In the following sections, mainly the differences in the thermal modelling are described. Thermal requirement 3 according to Section 4.1.2 is considered only, as it leads to slightly higher temperatures than thermal requirements 1 and 2.

4.7.4.2 Geometric Modelling For the numerical calculations, a 3D FE model is used, representing in circumferential direction one half of the transfer cask and content taking into account its symmetry. The FE model is shown

- in Figure 4.7-1 to Figure 4.7-3. The FE model consists of approximately** finite elements.

The FE model contains all thermally relevant components of the design parts lists. Section 1.1 con-tains the design parts lists and the corresponding design drawings with the component dimensions.

The material properties as described in Section 4.7.1 are used.

The geometrical model corresponds to the FE model described in Section 4.4.2 for NCS with the only exception that the canister during operation is loaded in the transfer cask instead of the stor-age cask. The modelling of the canister, the fuel basket and the FA is exactly the same and is de-scribed in Section 4.4.2 in detail. The transfer cask is not equipped with cooling fins. The interior of the canister and the space between canister and transfer cask are filled with pure water vapour. It is assumed that the transfer cask lid (optional) is mounted leading to higher temperatures conser-vatively.

4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 4.7-1 3D FE model of the transfer cask- overall view 4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-8

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 4.7-2 3D FE model of the transfer cask- detailed view Figure 4.7-3 3D FE model of the transfer cask- cross section 4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-9

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 (@)GNS 4.7.4.3 Boundary Conditions This section describes the boundary conditions applied to the transfer cask during vacuum drying.

The transfer cask stands inside the reactor facility. The heat removal from the cask surface occurs by free convection to the surrounding air and radiative heat transfer to the colder structures of the reactor facility.

For the ambient temperature for convection T A,canv, a conservatively high air temperature of 35 °C is applied. The ambient temperature for radiation TA.rad represents the temperature of the inner structures of the reactor facility and is set also to 35 °C. The transfer cask surface is not exposed to insolation.

At the external surface of the transfer cask, an effective heat transfer coefficient is applied, which consists of a convective portion hcanv and a radiative portion hrad:

heft = hconv + hrad On the cylindrical surface of the transfer cask, a convective heat transfer coefficient hcanv is applied which is derived from the following Nusselt law according to [3] for turbulent heat transfer by free convection at a vertical cylinder:

with: Nu - Nusselt number, -

Gr - Grashof number, -

Pr Prandtl number, -

h transfer cask height, m d - transfer cask diameter, m The Nusselt number Nu, the Grashof number Gr and the Prandtl number Pr are defined in Section 4.4.2.8.

In order to take into account partial disturbance of the convective heat transfer by a scaffold or po-tential additional shielding mounted to the transfer cask, the convective heat transfer coefficient hcanv is reduced by a factor of * .

At the lid-side front face of the transfer cask, the convective portion hcanv is calculated using the following Nusselt law according to [3] for turbulent heat transfer by free convection at a horizontal plane in case of heat release at the upper side:

4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-10

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 The radiative portion hrad is calculated using the Stefan Boltzmann law according to [3]:

The bottom-side front face of the cask is set adiabatic conservatively.

4.7.4.4 Maximum Temperatures After dewatering, the interior of the canister and the space between canister and transfer cask are filled with pure water vapour. Accordingly, the transfer cask and the content heat up during vacuum drying. The following results apply for an unlimited time period of vacuum drying in final steady state after heating up to thermal equilibrium.

Figure 4.7-4 shows the temperature distribution in the transfer cask in final steady state. Table 4.7-5 summarizes the resulting maximum temperatures for various components of the transfer cask and the content in final steady state.

Below, the design-relevant temperatures are compared to their maximum admissible values ac-cording to Section 4.3:

• The maximum temperature of the fuel rods amounts to 322 °C and is therefore significantly lower than the maximum admissible temperature of 400 °C.

• The maximum temperature for the lead shield is 103 °C which is far below the melting tem-perature of 327 °c.

• The highest gasket temperature of 128 °C occurs in the tightening plug gasket. The maxi-mum admissible temperature for continuous operation of the gaskets is ..re. Therefore, all gasket temperatures are far below the temperature limit.

• The maximum temperature inside the water chambers amounts to 100 °C, which not ex-ceeds the boiling temperature of 100 °C at a pressure of 1 bar. In reality, a safety margin exits because the water chambers are designed for a gauge pressure of 3 bar in the struc-tural evaluation in Chapter 3. This corresponds to an absolute pressure of 4 bar leading to a boiling temperature of 143 °C:

4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-11

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1

• Furthermore, temperature limits for structural components (e.g. fuel basket sheets) listed in Table 4.3-1, which are relevant for the mechanical integrity, are met.

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

Figure 4.7-4 Temperature distribution in the transfer cask during vacuum drying in final steady state 4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-12

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Table 4.7-5 Component temperatures for the transfer cask during vacuum drying in final steady state Component - type of temperature Temperature, °C Fuel rods - maximum 322 Cask lateral surface - maximum 92 Cask lateral surface - maximum circumferential average 92 Cask lateral surface - maximum longitudinal average 78 Water chamber, outer - maximum 96 Water chamber, inner - maximum 100 Lead shield - maximum 103 Cask cavity lateral surface - maximum 116 Cask cavity lateral surface - maximum circumferential average 116 Bottom-lid, upper side - maximum 180 Bottom lid, upper side - area average 153 Bottom lid, underside - maximum 178 Bottom lid, underside - area average 136 Canister inner lateral surface - maximum 170 Canister inner lateral surface - maximum circumferential average 165 Canister outer lateral surface - maximum 169 Canister outer lateral surface - maximum circumferential average 164 Canister bottom - maximum 215 Trunnion - maximum 69 Fuel channels - maximum 306 Basket sheets - maximum 299 Round segment - maximum 239 Outer sheets - maximum 251 Shielding element - maximum 254 Canister filling gas - volume average 260 Cask filling gas - volume average 110 Canister lid - maximum 143 Canister lid - volume average 131 Canister lid gasket - maximum 124 Cask lid - maximum 65 Cask lid - volume average 61 Tightening plug gasket - maximum 128 4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-13

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 4.7.5 Helium Backfilling After dewatering and vacuum drying, the interior of the canister is backfilled with helium. The space between canister and transfer cask is filled with dry air. In this section, the maximum temperatures are calculated after heating up to steady state in thermal equilibrium.

4.7.5.1 Thermal Model The calculation methods for short-term operations inside the reactor building in principle corre-spond to the approach for NCS described in Section 4.4.2. In the following sections mainly the dif-ferences in the thermal modelling are described. Thermal requirement 3 according to Section 4.1.2 is considered only, as it leads to slightly higher temperatures than thermal requirements 1 and 2.

4.7.5.2 Geometric Modelling The same numerical model of the transfer cask is used as described in Section 4.7.4.2 with the only exceptions that the interior of the canister is backfilled with helium and the space between canister and transfer cask is filled with dry air.

4.7.5.3 Boundary Conditions The same boundary conditions are applied as described in Section 4.7.4.3 with the only exception that the convective heat transfer coefficient hconv is not reduced by a factor o f - but applied to the full degree. That is because maximum temperatures occur after several days when the transfer cask is completely heated up to steady state. At this time, a scaffold or potential additional shield-ing, which partially disturb the convective heat transfer, are no longer mounted to the transfer cask.

4.7.5.4 Maximum Temperatures After dewatering and vacuum drying, the interior of the canister is backfilled with helium. For the steady state after heating up to thermal equilibrium, Table 4.7-6 summarizes the resulting maxi-mum temperatures for various components of the transfer cask and the content. Figure 4.7-5 shows the temperature distribution in the transfer cask after helium backfilling in final steady state.

4.7 Thermal Evaluation for Short-Term _Operations Section 4.7, Rev. 1 Page 4.7-14

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Below, the design-relevant temperatures are compared to their maximum admissible values ac-cording to Section 4.3:

• The maximum temperature of the fuel rods amounts to 248 °C and is therefore significantly lower than the maximum admissible temperature of 400 °C.

• The maximum temperature for the lead shield is 96 °c which is far below the melting tem-perature of 327 °C.

• The highest gasket temperature of 116 °C occurs in the tightening plug gasket. The maxi-mum admissible temperature for continuous operation of the gaskets is - °C. Therefore, all gasket temperatures are far below the temperature limit.

• The maximum temperature inside the water chambers amounts to 93 °C which is below the boiling temperature of 100 °C at a pressure of 1 bar. In reality, a higher safety margin exits because the water chambers are designed for a gauge pressure of 3 bar in the structural evaluation in Chapter 3. This corresponds to an absolute pressure of 4 bar leading to a boil-ing temperature of 143 °C.

• Furthermore, temperature limits for structural components (e.g. fuel basket sheets) listed in Table 4.3-1, which are relevant for the mechanical integrity, are met.

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

4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-15

Non-Proprietary Version .

1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 4.7-6 Component temperatures for the transfer cask after helium backfilling in final steady state Component - type of temperature Temperature, °C Fuel rods - maximum 248 Cask lateral surface - maximum 86 Cask lateral surface - maximum circumferential average 86 Cask lateral surface - maximum longitudinal average 77 Water chamber, outer - maximum 90 Water chamber, inner - maximum 93 Lead shield - maximum 96 Cask cavity lateral surface - maximum 108 Cask cavity lateral surface - maximum circumferential average

• 107 Bottom lid, upper side - maximum 156 Bottom lid, upper side - area average 136 Bottom lid, underside - maximum 155 Bottom lid, underside - area average 123 Canister inner lateral surface - maximum 163 Canister inner lateral surface - maximum circumferential average 161 Canister outer lateral surface - maximum 162 Canister outer lateral surface - maximum circumferential average 160 Canister bottom - maximum 181 Trunnion - maximum 69 Fuel channels - maximum 238 Basket sheets - maximum 236 Round segment - maximum 199 Outer sheets - maximum 205 Shielding element - maximum 203 Canister filling gas - volume average 202 Cask filling gas - volume average 105 Canister lid - maximum 126 Canister lid - volume average 118 Canister lid gasket - maximum 115 Cask lid - maximum 65 Cask lid -volume average 59 Tightening plug gasket - maximum 116 4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-16

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Figure 4.7-5 Temperature distribution in the transfer cask after helium backfilling in final steady state 4.7.6 On-Site Transfer for Normal Conditions After loading the canister into the storage cask, the storage cask is transferred on-site to the stor-age pad. The storage cask is transferred in a horizontal position lying on a transport vehicle. The protection cover and the storage frame are not yet mounted. In this section, thermal calculations concerning the on-site transfer for normal conditions are presented. Corresponding to NCS, a fail-ure of 1 % of the fuel rods in conjunction with a fraction of fission gas release from the fuel rods of 0.15 is considered according to Section 4.2.2.

4.7.6.1 Thermal Model The thermal analyses for the on-site transfer under normal conditions are based on the same cal-culation methods as described in Section 4.4.2 for NCS. The changes to the geometrical model and the differing boundary conditions are described in the following sections. Thermal require-4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-17

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS ment 3 according to Section 4.1.2 is considered only, as it leads to slightly higher temperatures than thermal requirements 1 and 2.

4.7.6.1.1 Geometric Modelling For the on-site transfer under normal conditions, the same geometrical model is used as for NCS (see Section 4.4.2.2 and Figure 4.4-1 to Figure 4.4-3) wit_h the following exceptions:

• The protection cover, the storage frame and the base plate of the storage pad are not part of the FE model.

• Likewise, the base plate of the storage pad is not modelled.

4.7.6.1.2 Filling Gases The thermal properties of the canister filling gas mixture and the homogenized FA zones are ap-plied according to Table 4.2-10 and Table 4.2-11 under consideration of fuel rod failure corre-sponding to NCS. The mechanical analyses in Chapter 3 show that the canister remains leak-tight during on-site transfer for normal conditions. For that reason, no fission gas release from the canis-ter interior to the cask cavity occurs.

4.7.6.1.3 Boundary Conditions This section describes the boundary conditions applied for the on-site transfer under normal condi-tions. The storage cask is transferred in a horizontal position lying on a transport vehicle. The heat removal from the cask surface occurs by free convection to the surrounding air and radiative heat transfer to the ambience.

For the ambient temperature for convection T A,conv, an environmental temperature of 38 °C accord-ing to Section 4.4.1 is applied. Conservatively, this value is used as a steady-state temperature.

The ambient temperature for radiation TA.rad is set to 38 °C, too.

At the external surface of the FE model, an effective heat transfer coefficient is applied, which con-sists of a convective portion hconv and a radiative portion hrad:

heff = hconv + hrad 4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-18

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 The convective portion hconv is calculated using the following Nusselt law according to [4] for turbu-lent heat transfer by free convection at vertical planes and horizontal cylinders:

The Nusselt number Nu, the Grashot number Gr and the Prandtl number Pr are defined in Section 4.4.2.8.*

The radiative portion hrad is calculated using the Stefan Boltzmann law according to [3]:

lnsolation is considered with the solar heat fluxes according to Section 4.4.1 averaged over 24 h per day. On the cylindrical surface of the cask, solar heat fluxes of 200 W/m 2 for curved surfaces are applied. At the lid-side and bottom-side front faces of the cask, solar heat fluxes of 100 W/m 2 for flat surfaces orientated vertically are used.

Heat transfer to the transport vehicle through the contact surfaces between cask and support bear-ings is neglected conservatively.

4.7.6.2 Maximum Temperatures The temperature distribution of the storage cask for the on-site transfer under normal conditions with fuel rod failure is shown in Figure 4.7-6. Table 4.7-7 lists the resulting maximum temperatures of the storage cask components. The comparison with the results for NCS in Table 4.4-5 shows that during on-site transfer the temperatures are 20-39 K lower because of the more favourable boundary conditions.

Below, the design-relevant temperatures are compared to their maximum admissible values ac-cording to Section 4.3:

• The maximum temperature of the fuel rods is 206 °C and is therefore significantly lower than the maximum admissible temperature of 400 °C.

4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-19

Non-Proprietary Version 1014-SR~00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 G S

• The maximum temperature for the moderator rods is 88 °C, for the bottom moderator plate 100 °C and 80 °C for the lid moderator plate. Therefore, the maximum temperatures of all moderator material are far below the maximum admissible temperature of ~ C .

• The highest gasket temperature of 87 °C occurs in the tightening plug gasket. The maxi-mum admissible temperature for continuous operation of the gaskets is ~ C . Therefore, all gasket temperatures are far below the temperature limit.

• Furthermore, temperature limits for structural components (e.g. fuel basket sheets) listed in Table 4.3-1, which are relevant for the mechanical integrity, are met.

The evaluation of the results shows that all calculated maximum temperatures of the cask compo-nents and the content are far below the maximum admissible values with large safety margins.

Figure 4.7-6 Temperature distribution during on-site transfer for normal conditions with fuel rod failure 4.7 Thermal Evaluation for Short-Term Operations Section 4. 7, Rev. 1 Page 4.7-20

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Table 4.7-7 Component temperatures during on-site transfer for normal conditions with fuel rod failure Component - type of temperature Temperature, 0 c

Fuel rods - maximum 206 Cask lateral surface - maximum 79 Cask lateral surface - maximum circumferential average 79 Cask lateral surface - maximum longitudinal average 74 Cask cavity lateral surface - maximum 90 Cask cavity lateral surface - maximum circumferential average 89 Cask cavity bottom - maximum 104 Cask cavity bottom - area average 97 Closure plate, underside - maximum 82 Closure plate, underside - area average 76 Moderator rods, inner row (MR-i) - maximum 88 MR-i - maximum cross-sectional average, hottest rod 85 MR-i -:-- volume average, hottest rod 81 Moderator rods, outer row (MR-o) - maximum 84 MR-o - maximum cross-sectional average, hottest rod 81 MR-o - volume average, hottest rod 78 Canister inner lateral surface - maximum 108 Canister inner lateral surface - max. circumferential average 106 Canister outer lateral surface - maximum 107 Canister outer lateral surface - max. circumferential average 105 Canister bottom - maximum 121 Moderator plate, bottom - maximum 100 Moderator plate, bottom - volume average 87 Moderator plate, lid - maximum 80 Moderator plate, lid - volume average 72 Retention ring - maximum 72 Closure plate - maximum 83 Closure plate - volume average 75 4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-21

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 4.7-7 Component temperatures during on-site transfer for normal conditions with fuel rod failure (continued)

Component - type of temperature Temperature, °C Trunnion - maximum 79 1

Trunnion screws - maximum 87 Fuel channels - maximum 193 Basket sheets - maximum 191 Round segment - maximum 151 Outer sheets - maximum 158 Shielding element - maximum 156 Canister filling gas - volume average 157 Cask filling gas - volume average 86 Canister lid - maximum 96 Canister lid - volume average 89 Canister lid gasket - maximum 85 Canister lid screws 1 - maximum 85 Cask lid - maximum 69 Cask lid - volume average 67 Cask lid gasket - maximum 69 Cask lid screws 1 - maximum 77 Protection cap gasket - maximum 68 Pressure switch gasket - maximum 68 Tightening plug gasket - maximum 87 1

For the screw temperatures, the surface temperature of the corresponding component plus a safety margin is used.

4.7.7 On-Site Transfer for Off-Normal Conditions According to the results in Section 4.7.6.2, the temperatures for normal conditions during on-site transfer are 20-39 K lower compared to NCS. Correspondingly lower temperatures result for off-normal conditions during on-site. transfer compared to off-normal conditions of storage because the same changes are applied to both thermal models. For that reason, no additional calculations are required. The temperatures calculated for off-normal conditions of storage in Table 4.5-1 and the evaluation in Section 4.5.4.2 cover also the case of off-normal conditions during on-site transfer.

4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-22

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 4.7.8 On-Site Transfer for Accident Conditions Fire During on-site transfer under accident conditions fire, a fire accident is considered. The evaluation comprises an initial steady state, a transient fire phase, when the cask is exposed to a fully-engulfing pool fire with an average flame temperature of 800 °C, and the subsequent transient cooling phase. Corresponding to ACS fire, a failure of 100 % of the fuel rods in conjunction with a fraction of fission gas release from the fuel rods of 0.15 is considered according to Section 4.2.2.

4.7.8.1 Thermal Model The thermal analyses for accident conditions fire during on-site transfer are based on the same calculation methods as described in Section 4. 7 .6.1 for normal conditions during on-site transfer.

The differing boundary conditions and assumptions under accident conditions fire are described in the following sections.

4. 7.8.1.1 Geometric Modelling For accident conditions fire during on-site transfer, the same geometrical model is used as for normal conditions during on-site transfer (see Section 4. 7.6.1.1 ).

For the transient calculation under accident conditions fire, the heat storage capacity of the fins, which are not modelled in the FE model explicitly, is taken into account as described in Section 4.6.1.1.1.

4.7.8.1.2 Filling Gases The thermal properties of the canister filling gas mixture and the homogenized FA zones are ap-plied according to Table 4.2-10 and Table 4.2-11 under consideration *of fuel rod failure corre-sponding to ACS fire. The adapted thermal properties are applied to the initial steady state as well as to the fire and cooling phase. The mechanical analyses in Chapter 3 show that the canister re-mains leak-tight during on-site transfer for accident conditions. For that reason, no fission gas re-lease from the canister interior to the cask cavity occurs.

4.7.8.1.3 Initial State before the Fire Accident The initial conditions before the fire accident correspond to the steady state for normal conditions during on-site transfer described in Section 4.7.6. The cask dissipates heat by natural convection and thermal radiation according to the boundary conditions described in Section 4.7.6.1.3.

4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-23

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS For accident conditions fire during on-site transfer, only thermal requirement 3 according to Section 4.1.2 is considered, as it leads to slightly higher temperatures than thermal requirements 1 and 2.

4.7.8.1.4 Boundary Conditions during the Fire Phase The cask is exposed to a fully-engulfirig pool fire with a constant flame temperature of 800 °C. The heat impact by the fire takes place at the entire external surface of the cask, namely at the cylindri-cal surface as well as the lid-side and bottom-side front face.

For the fire accident during on-site transfer, the same boundary c*onditions are applied for the fire phase as described in Section 4.6.1.1.4 for ACS fire. The only exception is that insolation is ap-plied with the solar heat fluxes according to Section 4. 7.6.1.3.

4.7.8.1.5 Boundary Conditions during the Cooling Phase For the cooling phase of the fire accident during on-site transfer, the same boundary conditions are applied as described in Section 4.7.6.1.3 for normal conditions during on-site transfer. The only exception is that for the cooling phase an emission coefficient of - is used for the radiative heat transfer at the sooted cask surface.

Due to the high thermal inertia of cask and content, the maximum temperatures inside the cask are reached several hours after the end of the fire phase. In the calculations, the cooling phase com-prises a sufficiently long time period of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> so that all temperatures of the components of cask and content reach their peak values and decrease again afterwards.

4.7.8.2 Maximum Temperatures As a result of the calculations for accident conditions fire during on-site transfer, Table 4.7-8 lists the maximum temperatures and their time of appearance (t = 0: beginning of fire) for various com-ponents of cask and content. Figure 4.7-7 to Figure 4.7-9 show the temperature courses over time for various components of cask and content.

Below, the design-relevant temperatures are compared to their maximum admissible values ac-cording to Section 4.3:

• The hottest fuel rod reaches after 26 h its maximum temperature of 266 °C, which is signifi-cantly lower than the maximum admissible fuel rod temperature for accident conditions of 570 °C valid for intact fuel rods.

4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-24

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS

• The temperatures of the gaskets are between 98 °C and 145 °C, which is considerably lower than the maximum admissible temperatures of - °C for the cask lid gasket and canister lid gasket and - °C for the pressure switch gasket, the protection cap gasket and the tightening plug gasket.

The evaluation of the results show that all calculated maximum temperatures of the cask compo-nents and the content are far below the maximum admissible values with large safety margins.

Table 4.7-8 Maximum component temperatures during on-site transfer for accident condi-tions fire with fuel rod failure and time of appearance after beginning of the fire Maximum Time of Component - type of temperature temperature, 0 c appearance, h Fuel rods - maximum 266 26.0 Cask lateral surface - half height 198 0.1 Cask cavity lateral surface - half height 101 2.5 Moderator rods, inner row (MR-i) - maximum 195 0.1 MR-i - maximum cross-sectional average, hottest rod 147 0.1 MR-i - volume average, hottest rod 100 2.3 Moderator rods, outer row (MR-o) - maximum 205 0.1 MR-o - maximum cross-sectional average, hottest rod 155 0.1 MR-o - volume average, hottest rod 99 1.3 Moderator plate, bottom - maximum 126 0.8 Moderator plate, bottom - volume average 116 1.3 Moderator plate, lid - maximum 109 0.6 Moderator plate, lid - volume average 97 1.8 Canister wall - maximum 121 3.5 Basket sheets - maximum 239 26.0 Shielding elements - maximum 198 19.0 Canister filling gas - volume average 202 19.0 Cask filling gas - volume average 102 2.7 Canister lid gasket- maximum 98 6.0 Cask lid gasket - maximum 133 0.2 Protection cap gasket - maximum 130 0.1 Pressure switch gasket - maximum 145 0.1 Tightening plug gasket - maximum 100 10.5 4.7 Thermal Evaluation for Short-Term Operations Section 4:7, Rev. 1 Page 4.7-25

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 4.7-7 Temperatures of the hottest fuel rod and the filling gases over time during on-site transfer for accident conditions fire with fuel rod failure Figure 4.7-8 Temperatures of the lid gaskets over time during on-site transfer for accident conditions fire with fuel rod failure 4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-26

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 4.7-9 Temperatures on the surfaces of cask, cavity and canister at half cask height over time during on-site transfer for accident conditions fire with fuel rod fail-ure 4.7.9 On-Site Transfer for Accident Conditions Impact According to the results in Section 4.7.6.2, the temperatures for normal conditions during on-site transfer are 20-39 K lower compared to NCS. Correspondingly lower temperatures result for acci-dent conditions impact during on-site transfer compared to ACS impact because the same chang-es are applied to both thermal models. For that reason, no additional calculations are required. The temperatures calculated for ACS impact in Table 4.6-8 and Table 4.6-9 and the evaluation in Sec-tion 4.6.3.2 cover also the case of accident conditions impact during on-site transfer.

4.7.10 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 7.

4.7.11 Maximum Thermal Stresses The discussion of thermal stresses due to temperature gradients within the components can be found in the structural evaluation in Chapter 3.

4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-27

Non-Proprietary Version 1014-SR-00002 Proprietary Information Withheld per 10CFR 2.390 Rev. 1 @GNS 4.7.12 Evaluation of Cask Performance for Short-term Operations For the short-term operations of the transfer cask considered above, the following temporal re-strictions apply:

• According to Section 4.7.2, no temporal restrictions are required for the operation phase of the fuel loading in the pool as long as the water-filled transfer cask is still under water and the canister lid is not yet mounted.

• For the operation phase of the water-filled transfer cask with mounted lid, the time period after mounting of the canister lid until the beginning of the dewatering has to be restricted to

-according to Section 4.7.3.

• For the operation phase of vacuum drying of the transfer cask interior, no temporal re-strictions are required according to Section 4.7.4.4.

• After the beginning of the backfilling with helium, the times for further operations and transport of the transfer cask inside the reactor building as well as during on-site transfer to the storage pad do not have to be restricted according to Section 4.7.5.4.

If the above mentioned temporal restrictions are respected, it is demonstrated that the CASTOR geo69 transfer cask fulfils all requirements for the short-term operations considered above with regard to thermal aspects. The following items summarize the results of the thermal investigations:

• The evaluation of the results in Sections 4.7.2, 4.7.3, 4.7.4.4, 4.7.5.4, 4.7.6.2, 4.7.7, 4.7.8.2 and 4.7.9 show that all calculated maximum temperatures of the cask components and the content are far below the maximum admissible values with large safety margins.

• It is demonstrated in Section 4.4.3 that additional safety margins exist because of the con-servative approaches for the thermal modelling.

• 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 during on-site transfer are incorpo-rated.

4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-28

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1

• During operation of the transfer cask inside the reactor building, it is demonstrated that the calculated maximum temperature for the lead shield is below the melting temperature which is requirement for ensuring the effectiveness of the shielding.

• During on-site transfer of the storage cask to the storage pad, it is shown that the calculated maximum temperatures of the moderator components do not lead to a thermal degradation of the moderator material which is requirement for ensuring the effectiveness of the shield-ing.

• 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. *

• During operation of the transfer cask inside the reactor building, the maximum temperatures inside the water chambers are far below the boiling temperature.

• The evaluation of the maximum pressure and a discussion on the generation of gases is documented in the containment evaluation in Chapter 7.

• The influence of the calculated temperatures on the mechanical material properties and thermal stresses is evaluated in the structural evaluation in Chapter 3.

List of References

[1] 2017 ASME Boiler and Pressure Vessel Code Section 11 Materials Part D Properties (Metric)

[2] Peter Grassmann Physikalische Grundlagen der Verfahrenstechnik

3. Auflage, Otto Salle Verlag, Frankfurt am Main, 1983 Physical Fundamentals of Process Engineering

[3] VDI Heat Atlas Calculation Sheets for the Heat Transfer Springer, 2010

[4] W. H. McAdams Heat Transmission McGraw-Hill, New York, 3rd Edition, 1985 4.7 Thermal Evaluation for Short-Term Operations Section 4.7, Rev. 1 Page 4.7-29

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 4.8 Appendix Name, Function Date Prepared 10.11.2022 Reviewed 10.11.2022 4.8 Appendix Section 4.8, Rev. 1 Page 4.8-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Appendix 4-1 GNS B 078/2018 E, Rev. 0 Validation of Computational Methods for the Thermal Design of Trans;port and Storage Casks Appendix 4-2 WTl/20/17 E, Rev. 0 Validation of the ANSYS 17.2 Finite Element Program for Thermal Ca1lcula-tions 4.8 Appendix Section 4.8, Rev. 1 Pa£1e 4.8-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev.1 5 Shielding Evaluation 5.0 Overview.

Prepared Reviewed 5.0 Overview Section 5.0, Rev. 1 Page 5.0-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS In this chapter, the shielding analysis of the CASTOR geo69 dual purpose cask (transport and storage cask) is presented in its storage configuration. The CASTOR geo69 storage cask as a part of DSS, in the following text also as (simply) storage cask identified, is designed to accommodate 69 spent nuclear fuel assemblies (SNF) from boiling water reactors (BWR). There a r e * *

• of SNF described in the contents description (see Section 1.2.3) authorised for storage in the cask.

In order to offer more flexibility in the fuel storage, three loading patterns TR1 to TR3 defined in Section 4.1 are used to distribute the SNF into ****** of fuel with identical characteristics (see Figure 5.0-1). These patterns are either uniform (TR1) or regionalised (TR2 and TR3) loadings, mainly identified by corresponding decay heats per SNF. It is shown in the following sections that the homogeneous pattern TR1 with its bounding source terms covers the other two patterns in a

-I sense of external dose rate.

The DSS comprises the storage cask and the protection cover. The storage configuration of the cask, analysed in this chapter, coincides with its transport version besides the absence of the impact limiters in storage. The protection cover being an integral part of the DSS is conservatively not mod-elled and is not discussed any further in this chapter.

This chapter will demonstrate that the design of the CASTOR geo69 storage cask fulfils the follow-ing acceptance criteria outlined in the Standard Review Plan NUREG-2215 [1]:

• The radiation shielding features of the proposed DSS must be sufficient for it to meet the radiation dose requirements in 10 CFR 72.104. This is demonstrated by providing a shielding analysis of the surrounding dose rates that contribute to off-site doses at appropriate dis-tances for a cask (typical array of casks in the most bounding site configuration) with bound-

- I

• ing source terms for normal conditions of storage and anticipated occurrences (off-normal conditions).

DSS contents and design features important for shielding are adequately described for eval-uating shielding effects and dose rates.

• Radiation shielding features must be sufficient for the design to meet the requirements in 10 CFR 72.106. This is demonstrated by calculating dose rates and doses at appropriate dis-tances for relevant ACS for appropriate configurations of the cask and assumptions regarding accidents.

• Dose rates from the cask must be consistent with a well-established "as low as reasonably possible" (ALARA) programme for activities in and around the storage site.

• The proposed shielding features should enable a general licensee that uses the dry storage system to meet the regulatory requirements prescribed in 10 CFR Part 20.

5.0 Overview Section 5.0, Rev. 1 Page 5.0-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1

• Appropriate distances for the foregoing criteria are distances that are consistent with, or bounding for, the distances to the controlled area boundaries of potential cask users. The minimum distance to the controlled area boundary is 100 meters.

Figure 5.0-1 Position groups in the basket This chapter provides the following information demonstrating full compliance with the aforemen-tioned criteria:

• A description of the shielding features of the storage cask

• A description of the bounding source term

• A description of the shielding analysis methodology as well of the shielding model and ma-terials

• Analyses of the external dose rates in the vicinity of the storage cask important for the han-dling operation in view of ALARA practices. These analyses include the configuration under normal, off-normal and accident conditions of storage

• Analyses of the dose rates and doses at the controlled area boundary performed with the covering configuration of the storage cask array For flexibility regarding configuration of the storage site, an analysis aiming at the determination of the minimum distance to the site boundary, where a most penalising array of the DSS complies with the radiation dose limit from 10 CFR 72.104, is extended. In addition to this evaluation, a minimum 5.0 Overview Section 5.0, Rev. 1 Page 5.0-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 concrete wall thickness of a storage building is determined so that the radiation dose limit from 10 CFR 72.104 is met at a distance of 100 m. Although this information is not required for the free-field storage, it might be useful for future modifications of the storage site.

The use of the CLU (see Section 1.2) including the transfer cask, the transfer lock and further equip-ment its contribution to the dose at the storage site boundary can be neglected (see Section 5.1 ). A discussion of the estimated occupational exposures for the CLU is considered in chapter 11 together with the exposures of the CASTOR geo69 DSS. A corresponding shielding model of the CLU is presented in Appendix 5-3 in Section 5.5.

In this chapter, the dose rates and doses from the direct neutron and gamma radiation stemming

- from the storage cask are calculated. The discussion of the possible release of the radioactive ma-terials from the storage cask is presented in chapter 7.

List of References

[1] NUREG-2215, Standard Review Plan for Spent Fuel Dry Storage Systems and Facilities, Office of Nuclear Material Safety and Safeguards, April 2020 5.0 Overview Section 5.0, Rev. 1 Page 5.0-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev.1 5.1 Discussion and Results Prepared Reviewed 5.1 Discussion and Results Section 5.1, Rev. 1 Page 5.1-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The principal sources of radiation from SNF are:

• Gamma radiation originating from a decay of actinides and radioactive fission products, of fuel and hardware activation products generated during reactor operation, as* well as sec-ondary gamma particles from neutron capture,

• Neutron radiation from spontaneous fission, from (a,n)-reactions in fuel materials, from sec-ondary neutrons produced by fission via subcritical multiplication, and from (y,n)-reactions.

The latter source is however negligible.

The major parts of the storage cask relevant for the shielding of radiation sources are:

• the basket with 69 positions to accommodate SNF,

• the canister with its lid,

-

• the cask body (with its lid) incorporating moderator rods and plates.

Shielding from gamma radiation is provided by the steel structure of the canister and the lid system and by the ductile cast iron (DCI) of the cask body. In order to make neutron shielding effective, the neutrons have to be thermalised and then absorbed. For this purpose, the moderator rods and plates made of unborated ultra high-molecular weight polyethylene (UHMW PE) are incorporated into the cask body. Together with relatively high carbon contents in the DCI, they provide an eff~ctive way to thermalise neutrons. Sufficient DCI behind the polyethylene rods towards the external surface of the cask not only allows for an efficient absorption of neutrons, but greatly supresses the high energetic secondary gamma radiation.

The borated structures of the basket ( are not primarily aimed to improve the shield-ing performance of the cask, but nevertheless diminish the thermal part of the neutron spectrum around the SNF to some extent. This helps to reduce the dose rate contributions (mostly) from the

- inner fuel assemblies.

Additional basket elements (round segments and shielding elements) made of aluminium are added to the basket not only to stabilise it and support heat dissipation, but also to provide some additional gamma shielding.

The cross-sectional top view of the storage cask (quarter cut) is presented in Fig_ure 5.1-1, while the elevation section of the cask is displayed in Figure 5.1-2. These views are directly generated from the calculation inputs at normal conditions. The colour code corresponds to the materials used in the calculations.

In Section 5.3 the shielding model of the storage cask is presented in greater detail.

5.1 Discussion and Results Section 5.1, Rev. 1 Page 5.1-2

Non-Proprietary Version

• 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Figure 5.1-1 Cross sectional view of the cask model (2020 mm above cask bottom)

Figure 5.1-2 Elevation cut through the cask model 5.1 Discussion and Results Section 5.1, Rev. 1 Page 5.1-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The most important dimensions of the storage cask for the shielding calculations in its central plane are presented in Figure 5.1-3 (standard MCNP colours are desaturated). The dimensions according to the technical drawings in Section 1.5 are highlighted as blue text, the implementation into the shielding model as green text. The thicknesses of the materials relevant for the shielding analysis are set to their minimum.

The occurring gaps are filled with air.

Figure 5.1-3 Cross sectional view of the cask model with major dimension (in mm) 5.1 Discussion and Results Section 5.1, Rev. 1 Page 5.1-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Axially the main shielding is provided by:

• a cask bottom part consisting of thick DCI part ( bottom modera-tor plate, and closure plate;

• the lid system consisting of the canister lid ( * * *

• moderator plate ( ), and cask lid No credit has been taken of the protection cover (see Section 1.2).

The dose rates on the lid side of the storage cask are much smaller than on the shell side, which is decisive for the design, I The densities of the materials are reduced relative to their nominal values as discussed in Section

- A unique feature of the CASTOR geo69 are the moderator rods placed directly in the cask body.

The rods are made of UHMW PE without neutron absorbing additives and serve the neutron mod-eration purpose only. From point of view of storage, few standard situations are to be considered:

Two shielding models, cold and hot, are hence analysed. One of the two delivering the highest ex-ternal dose rates is considered as representative for the cask design. Figure 5.1-4 and Figure 5.1-5 illustrate these two shielding models for the inner and outer moderator rods, respectively. In both 5.1 Discussion and Results Section 5.1, Rev. 1 Page 5.1-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS cases, the geometry of the setup is implemented according to Section 3.3, and the tolerances are chosen such that the air gaps are maximised.

Figure 5.1-4 Inner moderator rods under different operating conditions (dimensions in mm)

Figure 5.1-5 Outer moderator rods under different operating conditions (dimensions in mm) 5.1 Discussion and Results Section 5.1, Rev. 1 Page 5.1-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The information about particular dose rates is gained from detectors positioned all around the cask.

Besides this geometry-independent mesh of detectors ( , separate volumet-ric detectors are modelled in order to control the calculation process. The maximum dose rates are always searched for in different dose rate locations.

High burn-up spent fuel is intended to be stored in the DSS beyond 20 years of dry storage, therefore, the impact of the fuel failure under normal, off-normal and accident conditions of storage is evaluated according to NUREG-2224 [1 ]. It is assumed that under normal and off-normal conditions of storage the DSS remains vertically oriented, thus the source occurring due to the fuel failure is relocated towards the bottom of the canister into the region with potentially lower shielding performance due to flattenings in the reg ions of the tilting studs and also due to the finite size of the moderator rods.

The corresponding dose rate ~valuations are performed for normal and off-normal conditions with the source terms after 20 years of dry storage.

Besides the fuel failure there are no factors influencing the shielding performance of the storage cask under off-normal storage conditions. Contrary to this, under ACS there is a certain probability to completely lose the neutron moderator as a result of fire. It is shown, however, in chapters 3 and 12 that the basket, canister, and the cask remain largely unaltered in different accident scenarios and that the complete loss of the neutron moderator is excluded. Nevertheless, shielding analysis is performed with this highly conservative assumption. A fuel failure of 100 % according to [1] with fuel relocation is also considered. For this scenario, no additional cooling down of the contents is as-sumed.

The code system MCNP6 in version 2.0 [2] is used to calculate the dose rates. An overview of the computer codes used in the entire calculation chain is summarised in Appendix 5-1.

The fuel assemblies are generally classified by their type (SNF number, see Section 2.1) and by their maximum allowed decay heat (see Section 4.1 ). These two quantities, geometry and heat, complemented by the minimum initial enrichment and maximum burn-up including burn-up uncer-tainty unambiguously define the bounding source terms described in detail in Section 5.2.

In the shielding model, the design of the SNF No 6, is used.

With this choice, the conservative configuration is realised.

As discussed in Section 5.4, the dose rates for the uniform loading pattern (TR1) bound the dose rates for both regionalised loading patterns (TR2 and TR3), therefore, the fuel failure scenarios are 5.1 Discussion and Results Section 5.1, Rev. 1 Page 5.1-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS only performed for the former loading pattern. For the same reason, the dose rates and doses at the far distances from the storage cask are calculated for the uniform loading pattern only.

In order to calculate the dose to public and to understand, where the boundary of the restricted area should be established, a bounding array of storage casks is investigated. I Figure 5.1-6 Bounding array of storage casks

- I Keeping in mind potential modifications of the storage site with respect to the free-field storage, it is separately investigated, what the minimum distance of compliance with the requirements of 10 CFR 72.104 is, which minimum thickness of the storage building is needed to comply with the require-ments of 10 CFR 72.104 already at a distance of 100 m, and whether the requirements of 10 CFR 72.106 are fulfilled.

Apparently, none of the off-normal conditions has any impact on the shielding analysis. The only significant difference is that under off-normal conditions 10 % fuel failure is assumed instead of 3 %

for the storage period beyond 20 years. Other boundary conditions remain identical for the purpose of the shielding evaluation.

The 10 CFR 72.104 criteria for radioactive materials in effluents and direct radiation during normal and off-normal operations are:

• During normal operations and anticipated occurrences, the annual dose equivalent to any real individual who is located beyond the controlled area must not exceed 0.25 mSv (25 mrem) to the whole body, 0.75 mSv (75 mrem) to the thyroid and 0.25 mSv (25 mrem) to any other critical organ.

• Operational restrictions must be established to meet as low as reasonably achievable (ALARA) objectives for radioactive materials in effluents and direct radiation.

10 CFR 20 [3] Subpart C and D specify additional requirements for occupational dose limits and radiation dose limits for individual members of the public. Chapter 11 addresses these regulations.

For this chapter, a dose rate limit at the restricted area boundary of 0.02 mSv/h (2 mrem/h) is of particular interest.

5.1 Discussion and Results Section 5.1, Rev. 1 Page 5.1-8

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The 10 CFR 72.106 radiation dose limits at the controlled area boundary for design basis accidents are:

• Any individual located on or beyond the nearest boundary of the controlled area may not receive from any design basis accident the more limiting of a total effective dose equivalent of 0.05 Sv (5 rem), or the sum of the deep-dose equivalent and the committed dose equiva-*

lent to any individual organ or tissue (other than the lens of the eye) of 0.5 Sv (50 rem). The lens dose equivalent may not exceed 0.15 Sv (15 rem) and the shallow dose equivalent to skin or any extremity may not exceed 0.5 Sv (50 rem). The minimum distance from the spent fuel or high-level radioactive waste handling and storage facilities to the nearest boundary of the controlled area must be at least 100 meters.

- In accordance with ALARA practices, sufficiently low dose rates have to be established in the vicinity of the storage cask, at its surface and in the area close to the cask. Table 5.1-1 presents an overview of the total dose rates (sum of the neutron radiation, direct gamma radiation, induced gamma radia-tion and gamma radiation from fuel hardware) around the storage cask for normal, off-normal and accident conditions. The dose rates calculated within different fuel failure scenarios discussed above are presented as well. All values presented include 2cr statistical uncertainty as described in Section 5.4.

The following conclusions can be made:

• Even without impact limiters the storage cask complies with the transport requirements ac-cording to 10 CFR 71.

• The dose rates from the uniform loading pattern (TR 1) are higher than those from the region-alised ones at the shell side of the storage cask, which is decisive for the distant dose rate due to its magnitude.

• Fuel failure under normal or off-normal conditions does not lead to an increase of the dose rate. For that reason, the storage site evaluations are performed with an undamaged fuel model.

• Fuel failure under accident condition leads to a local increase of the dose rate at the storage cask surface. On the shell side, this effect vanishes with increasing distance from the cask.

5.1 Discussion and Results Section 5.1, Rev. 1 Page 5.1-9

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS At a distance of 1 m from the cask, the dose rates caused by the undamaged fuel are higher.

The same is true also for the fuel failure case, when the storage cask remains upright ori-ented. The fuel failure followed by the concentration of the rubble at the lid side of the cask can only occur in case of over-tipping, for which the shell side of the storage cask plays a major role for the external dose rate. Summarising the accident related findings, the accident without fuel damage is bounding and generates the highest dose rate at the site boundary.

The maximum dose rate at the cas!< surface under normal (and off-normal) conditions is dominated by tlll- The total dose rate of e.g. a cold cask loaded according to TR1 sums up as follows: 1.47 mSv/h with the contributions from primary gamma radiation, neutrons including secondary gamma radiation, and hardware radiation, respectively. The details about dose rate distributions at the cask surface (and at a distance of 1 m from the surface as well) are presented in Section 5.4.

At 1 m distance from the cask surface and the total dose rate at maximum is 0.162 mSv/h with the contributions from primary gamma radiation, neutrons including secondary gamma radiation, and hardware radi-ation, respectively.

An annual dose (8766 hour0.101 days <br />2.435 hours <br />0.0145 weeks <br />0.00334 months <br /> annual occupancy) from a standalone storage cask without any addi-tional protection is presented in Figure 5.1-7 as a function of distance from the cask. The angles of 45° and 90° are analysed as they reproduce different shielding geometries (see Figure 5.1-1 ). It can be seen that the distant annual doses are not very different with a marginal excess _ , . The neutron radiation share as well as the combined contribution of primary, secondary and 6 °Co gamma

- radiation from the fuel hardware (PG+SG) are displayed. The annual dose limits for a controlled area (10 CFR 72.104) and a restricted area (10 CFR 20.1301, recalculated to an annual dose for easier presentation) are shown in red colour. Easy to see, the 0.25 mSv/a-limit is met at a distance of

- At least - around the storage cask has to be restricted as the dose rate exceeds 0.02 mSv/h within this circle.

Figure 5.1-8 plots the annual dose from the bounding cask array (see Figure 5.1-6) at various dis-tances from the centre of the long side of the array. As in Figure 5.1-7, the total *annual dose and the contributions from gamma and neutron radiation are displayed. This conservative arrangement_

of controlled area on its long side. Almost

- space has to be reserved for the restricted area according to 10 CFR 20.

5.1 Discussion and Results Section 5.1, Rev. 1 Page 5.1-10

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 An additional contribution to the annual dose at the controlled area boundary (10 CFR 72.104) comes from the transfer cask (CLU),

the assessment of the dose rate is performed assuming that the transfer cask is not surrounded with any concrete or metal structure. The shielding model according to Appendix 5-3 (see Section 5.5) is used to determine the dose rate at different distances from the cask. In order for the annual doses to be estimated, a total duration for a single CLU handling has been assumed. Together with an estimated amount of

- a total duration has been considered. The resulting annual dose as a function of the distance is presented in Figure 5.1-9. The off-normal conditions involving the transfer cask include e.g. a loss of external power, equipment failure or human failure leading to the extension of the handling operation.

Therefore, it is expected that the corresponding annual dose from the transfer cask under off-normal conditions would be At a provisional controlled area boundary of approx. 400 m the annual dose from the transfer cask amounts to 0.015 mSv/a (0.020 mSv/a at 385 m from the cask) under assumption that the CLU is operated outside of the building. When taking the structures of the NPP building into account (e.g.

50 cm concrete), at least a factor-lower dose is expected (cf. Figure 5.1-8 and Figure 5.1-10).

This means that the contribution of the transfer cask to the annual dose (10 CFR 72.104) is negligibly small.

According to Section 2.2 ACS and natural phenomena are no credible events for the transfer cask exclusively handled inside the NPP building and are thus not be taken into account in the shielding analysis.

A potential modification of the free-field storage is to place the storage cask in a storage building. At least -are needed, so that the 10 CFR 72.104 dose limit is complied with at a distance of 100 m (see Figure 5.1-10).

After an accident, when all moderator material is assumed to be lost, no presence of any additional material (e.g. building remnants) is assumed to provide additional shielding. The 30 days-dose in this case is dominated by neutron radiation but remains safely under the dose limit of 10 CFR 72.106 at 100 m distance. The dose limit is strictly met** distance (see Figure 5.1-11 ).

5.1 Discussion and Results Section 5.1, Rev. 1 Page 5.1-11

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 5.1-7 Annual dose as a function of distance from the storage cask Figure 5.1-8 Annual dose from the storage cask array as a function of distance from the array 5.1 Discussion and Results Section 5.1, Rev. 1 Page 5.1-12

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 5.1-9 Annual dose from transfer cask as a function of distance from the cask 5.1 Discussion and Results Section 5.1, Rev. 1 Page 5.1-13

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 5.1-1 Total dose rates in the vicinity of a single storage cask under different condi-tions Maximum Dose Rate at the Shell Surface Maximum Dose Rate at the Lid Surface for Normal and Off-Normal Operation for Normal and Off-Normal Operation (for Accident), mSv/h (for Accident), mSv/h Loading Pattern Fuel Failure Fuel Failure 3%/10% 3 % / 10 %

Cold Cask Warm Cask Cold Cask Warm Cask (100% (100%

bottom I top) bottom I top) 1.47 1.55 0.042 0.041 1R1 (6.29) ( ) ( 0.324) ( )

1R2 1.00 I 1.05 - 0.050 I 0.050 -

( 7.02) - ( 0.370) -

1R3 1.34 I 1.42 - 0.047 I 0.048 -

( 6.61) - ( 0.352) -

Maximum Dose Rate at 1 m from Cask Shell Maximum Dose Rate at 1 m from Cask Lid for Normal and Off-Normal Operation for Normal and Off-Normal Operation (for Accident), mSv/h (for Accident), mSv/h Loading Pattern Fuel Failure Fuel Failure 3 %/10 % 3%/10 %

Cold Cask Warm Cask Cold Cask Warm Cask (100% (100%

bottom / top) bottom I top) 0.162 0.158 0.016 0.016 TR1

( 2.32) ( ) ( 0.110) ( )

1R2 0.133 I 0.127 - 0.017 I 0.017 -

( 2.57) - ( 0.129) -

1R3 0.134 I 0.129 - 0.017 I 0.017 -

(2.43) - ( 0.123) -

Maximum Dose Rate at 2 m from Cask Shell Maximum Dose Rate at 2 m from Cask Lid for Normal and Off-Normal Operation for Normal and Off-Normal Operation (for Accident), mSv/h (for Accident), mSv/h Loading Pattern Fuel Failure Fuel Failure 3%/10% 3%/10%

Cold Cask Warm Cask Cold Cask Warm Cask (100% (100%

bottom I top) bottom I top) 0.093 0.090 0.009 0.009 TR1

( 1.22) ( ) ( 0.072) ( )

1R2 0.076 I 0.071 - 0.013 I 0.011 -

( 1.35) - ( 0.083) -

1R3 0.078 I 0.073 - 0.013 I 0.011 -

( 1.28) - ( 0.079) -

5.1 Discussion and Results Section 5.1, Rev. 1 Page 5.1-14

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 5.1-10 Annual dose from the storage building as a function of distance Figure 5.1-11 Annual dose 'from the storage cask array under accident storage conditions 5.1 Discussion and Results Section 5.1, Rev. 1 Page 5.1-15

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 (@)GNS Summarising the evaluation, to comply with the dose limit of 0.25 mSv/a to any real individual at or beyond the controlled area boundary (10 CFR 72.104) one has to:

• Make sure that the boundary is far off from the storage cask array, or

• Guarantee that the storage casks are placed into a storage building In this case, the dose limit is met at the minimum dis"'"

tance of 100 m (10 CFR 72.106).

A detailed site specific evaluation of dose at the controlled area boundary must be performed for each ISFSI in accordance with 10 CFR 72.212.

The complete loss of the moderator material as a result of an accident seriously affects the dose generated by the considered bounding storage cask array. Assuming an accident duration of 30 days, the accumulated dose at the controlled area boundary would be** safely below the dose limit of 50 mSv.

For the accident conditions the compliance with the 10 CFR 72.106 limit is demonstrated without any additional requirements.

List of References

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

[2] C.J. Werner (ed.), MCNP User's Manual - Code Version 6.2, LA-UR-17-29981, 2017

[3] Title 10 CFR Part 20 Standards for Protection Against Radiation U.S. Nuclear Regulatory Commission 5.1 Discussion and Results Section 5.1, Rev. 1 Page 5.1-16

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 5.2 Source Specification Name, Function Date Signature Prepared Reviewed I

5.2 Source Specification Section 5.2, Rev. 1 Page 5.2-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 The procedure to identify bounding source terms for each loading pattern TR1 to TR3 is presented in this section. For each fuel assembly described in Section 2.1 the following quantities are known:

• lattice parameters and geometry,

• minimum initial enrichment (see also Table 5.2-1 ),

• maximum end burn-up (with burn-up uncertainty, see Table 5.2-1 ),

• maximum allowed decay heat.

With these parameters burn-up and depletion calculations are performed (see Section 2.1 for an extended description). The nodal burn-up profile data as well as the nodal value for the moderator density are provided in Table 2.1-2. To accurately account for an axial distribution, 24 individual calculations are executed for a single geometry resulting in nodal values for the physical SNF prop-erties. Cooling times of up to more than 45 years are analysed for each SNF separately, and decay heat, gamma and neutron source terms are compared to find the maximum values. This helps to unambiguously identify the minimum cooling times, at which the decay heat required by the certain loading pattern <111) is just reached. Otherwise, the particular SNF are not au-thorised for loading.

The decay heat as well as the radiation _sources are calculated for every SNF geometry using ORIGAMI with the necessary libraries generated beforehand using TRITON (see Appendix 5-1) as discussed in Section 2.1. The validity of the ORIGAMI output is checked against multiple TRITON calculations (see Appendix 5-2).

What concerns axial distributions of the sources, some model assumptions have to be mad For two important model ingredients the assumptions are made based on the available information from representative power plants:

1. To obtain the bounding source terms have been analysed and the node wise minimum density is chosen (Section 1.2.2). The source term strengths calculated with the minimum enveloping density profile are higher than the other ones.

2. A number of existing discharged fuel assemblies has been analysed targeting the axial dis-tributions of the decay heat as well of the radiating source. Concluding this analysis, con-servative axial distributions of the source terms are identified (Section 1.2.2).

5.2 Source Specification Section 5.2, Rev. 1 Page 5.2-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The burn-up profiles used in the analysis compare nicely with the representative profiles utilised in Ref. [1]. The moderator density axial profile used in the analyses represents a bounding node-wise minimum of profiles corresponding to the complete irradiation histories of the considered fuel as-semblies.

It also compares nicely to the cycle-average axial void fraction from Ref. [1], which in turn is representative for the assemblies with a discharge burn-up of above 30 GWd/MgHM [2].

The radiation sources relevant for the dose rate outside of the cask consist of gamma and neutron radiation from the active zone of the SNF as well as from gamma radiation due to 6°Co from the activated hardware including fuel rod plena and tie plates.

Each basket position is characterised by the maximum allowed decay heat power (see Section 4.1 ).

The minimum cooling times required to reach a certain decay heat for a particular loading pattern are presented in Table 5.2-2.

Table 5.2-1 Boundary conditions for the burn-up calculations Table 5.2-2 Minimum cooling times (in years) needed to reach certain decay heat 5.2 Source Specification Section 5.2, Rev. 1 Page 5.2-3

Non-Proprietary V ersion 1014-SR-00002 Proprietary Inform ation withheld per 10CFR 2.390 Rev. 1 @GNS 5.2.1 Gamma Source The gamma-radiation fuel is analysed in seven energy groups (see Table 5.2-3), representing ener-gies relevant for the dose rate outside of the storage cask. The gamma particles with lower energy are so well shielded that they do not signi ficantly contribute to the outer dose rate. The high energy gamma particles possess tiny source term strengths and, therefore, do not contribute to the dose rate either.

Table 5.2-3 Gamma energy structure Energy group, i = 1 2 3 4 5 6 7 Average group energy, MeV 0.5 75 0.85 1.25 1.75 2.25 2.75 3.5 Lower group energy, MeV 0. 45 0.7 1 1.5 2 2.5 3 Upper group energy, MeV 0 .7 1 1.5 2 2.5 3 4 The gamma source is determined for every SNF type described in Section 2.1. The total source strength in each gamma en~rgy group is calculated by summation of the 24 axial nodal values stem-ming from the conservative axial burn-up profile discussed above.

This confirms close to linear behaviour of the total gamma source as a function of the assembly burn-up.

5.2 Source Specification Section 5.2, Rev. 1 Page 5.2-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS This justifies the use of the high burn-up profile for the generation of the axial source strength distri-butions. The axial gamma source strength profile used in present shielding calculations are displayed in Figure 5.2-1. Different shapes of the profiles in various groups can be related to the most contrib-uting nuclides.

Figure 5.2-1 Axial gamma source strength distributions As discussed above, the nodal gamma sources are determined. It is checked that no increase of the source terms over time occurs. For all seven groups the sources monotonously decrease. The re-sulting gamma sources based on Table 5.2-2 for every SNF type are summarised in Table 5.2-4 to

-

• Table 5.2-10. The sources are normalised to one megagram (Mg) of heavy metal to account for the differences in the mass of the SNF including the mass in the shielding model. The bold value are the maximum sources for each fixed decay heat.

Besides gamma particles stemming from the fuel pellet stack, there is also a gamma radiation from activated hardware: the end fittings, the plenum springs, and the grid spacers. The primary source of activity in the non-fuel regions of a SNF arise from the activation of 59 Co to 6 °Co. The activities of 6

°Co are determined during burn-up and depletion calculations together with the calculations for the primary gamma radiation.

This value is scaled with the maximum burn-up of the particular SNF and corresponding masses of hardware. The flux scaling factors of 0.1 for the top 5.2 Source Specification Section 5.2, Rev. 1 Page 5.2-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS end fittings and of 0.2 for the bottom end fittings and plenum springs are applied according to PNL-6906 [3]. The top handles and the lower parts of the SNF bottom end pieces are conservatively taken 59 into account by assuming that the whole Co from the top or bottom end piece is concentrated in the upper respectively lower tie plate. The activities are finally decayed to the desired cooling time from Table 5.2-2. The gamma sources for each SNF type are given in Table 5.2-11 to Table 5.2-14.

Some older SNF may have higher 59 Co impurity level in their structural elements (including spacers),

the minimum cooling times specified in Section 2.1, however, imply that the activity of 6°Co is largely cooled down.

Yet another gamma source arises from (n,y) reactions in the materials of the storage cask. This source is properly accounted for in MCNP, when neutron calculations are performed in a coupled neutron-gamma mode, which is the case for the present shielding analysis.

The bounding gamma source for each loading pattern TR1 to TR3 is constructed from the maximum individual sources corresponding to the selected decay heat. The bounding gamma source terms used in the shielding analysis are reported in Table 5.2-15 forTR1; in Table 5.2-16 forTR2, and in Table 5.2-17 for TR3 decoded The bounding gamma energy release rate for the gamma radiation from the spent fuel is summarised in Table 5.2-18.

Table 5.2-4 SNF type dependent gamma sources for energy group i=1 5.2 Source Specification Section 5.2, Rev. 1 Page 5.2-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 5.2-5 SNF type dependent gamma sources for energy group i=2 Table 5.2-6 SNF type dependent gamma sources for energy group i=3 Table 5.2-7 SNF type dependent gamma sources for energy group i=4 5.2 Source Specification Section 5.2, Rev. 1 Page 5.2-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 5.2-8 SNF type dependent gamma $O,urces for energy group i=5 Table 5.2-9 SNF type dependent gamma sources for energy group i=6 Table 5.2-10 SNF type dependent gamma sources for energy group i=7 5.2 Source Specification Section 5.2, Rev. 1 Page 5.2-8

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @)GNS Table 5.2-11 SNF type dependent gamma sources for the bottom end fittings Table 5.2-12 SNF type dependent gamma sources for the top end fittings Table 5.2-13 SNF type dependent gamma sources for the plenum springs 5.2 Source Specification Section 5.2, Rev. 1 Page 5.2-9

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 5.2-14 SNF type dependent gamma sources for the grid spacers Table 5.2-15 Bounding gamma source term for TR1 Table 5.2-16 Bounding gamma source term for TR2 5.2 Source Specification Section 5.2, Rev. 1 Page 5.2-10

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Table 5.2-17 Bounding gamma source term for TR3 Table 5.2-18 Bounding energy release rate per Mg heavy metal 5.2.2 Neutron Source The neutron sources are determined in an analogous way as the gamma ones. The two relevant neutron energy spectra - from spontaneous fission and from (a,n) reactions - exhibit different axial distributions due to the weaker burn-up towards the ends of the SNF (see Figure 5.2-2).

It is checked, whether energy distributions of these two spectra could be realised by internal means of the MCNP code system (continuous energy distributions, see [4]). While the energy spectrum 244 from the spontaneous fission could be nicely described by the Watt spectrum from Cm (parame-ters a= 0.902523 MeV, b = 3.72033 Mev-1 , see [4]), for the (a,n) energy spectrum no internal func-tion has been found as it is rather a modified Maxwell distribution currently not implemented in the code. Finally, for (a,n)-neutrons a histogram function generated by the performed burn-up and de-pletion calculations is utilised in the analysis (see Table 5.2-25).

The subcritical neutron multiplication is properly taken into account during particle transport with MCNP.

5.2 Source Specification Section 5.2, Rev. 1 Page 5.2-11

• Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 The nodal neutron source is determined in the l;,ame fashion (simultaneously) as the gamma one and is shown in Table 5.2-19 for spontaneous fission and in Table 5.2-20 for the (a,n)-reactions. The bounding neutron source for each loading pattern TR1 to TR3 is constructed from the maximum individual sources corresponding to the selected decay heat. The bounding source terms used for the shielding analysis are reported in Table 5.2-21 for TR1, in Table 5.2-22 for TR2, and in Table 5.2-23 for TR3.

The total neutron sources for spontaneous fission and for (a,n)-reactions as a function of energy are displayed in Table 5.2-24 and Table 5.2-25, respectively.

Figure 5.2-2 Axial neutron source strength distributions 5.2 Source Specification Section 5.2, Rev. 1 Page 5.2-12

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 5.2-19 SNF type dependent neutron sources from spontaneous fission Table 5.2-20 SNF type dependent neutron sources from (a,n)-reactions Table 5.2-21 Bounding neutron source for TR1 Table 5.2-22 Bounding neutron source for TR2 5.2 Source Specification Section 5.2, Rev. 1 Page 5.2-13

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 5.2-23 Bounding neutron source for TR3 Table 5.2-24 Bounding spontaneous fission neutron source as a function of energy 5.2 Source Specification Section 5.2, Rev. 1 Page 5.2-14

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 5.2-25 Bounding (a,n)-reaction neutron source as a function of energy List of References

[1] NUREG/CR-7224, Axial Moderator Density Distributions, Control Blade Distributions, and Axial Burnup Distributions for Extended BWR Burnup Credit, ORNL/TM-2015/544

[2] CAL-DSU-NU-000005, Rev. 00A, CSNF Disposal Container, BWR Axial Profile, OCRWM, 2004

[3] PNL-6906 Vol. 1 to Vol. 3, UC-85 A. Luksic, Spent Fuel Assembly Hardware: Characterization and 10 CFR 61 Classification for Waste Disposal, 1989

[4] C.J. Werner (ed.}, MCNP User's Manual - Code Version 6.2, LA-UR-17-29981, 2017 5.2 Source Specification Section 5.2, Rev. 1 Page 5.2-15

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 5.3 Model Specification Name, Function Date Signature Prepared I

Reviewed I

5.3 Model Specification Section 5.3, Rev. 1 Page 5.3-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 The shielding analysis of the DSS is performed with MCNP6 2.0 [1 ]. For the storage cask, a separate MCNP calculation is performed for each position group A to F (see Figure 5.0-1) for every twelve sources (seven gamma groups from the fuel pellet stack, two neutron spectra, 6°Co radiation from top end fittings, bottom end fittings, and plenum springs).

In this section, the shielding models used in the calculation are discussed. The information about individual constituents of the cask and the materials used in models are described.

In total, a set of shielding models for the storage cask and cask array are prepared to cover all the aspects of the safe storage. Except possible fuel failure scenarios due to the storage period beyond 20 years, none of the off-normal conditions have any impact on the shielding analysis. Therefore, normal and off-normal conditions are generally identical.

The shielding models for a single storage cask are as follows:

The shielding models for a bounding array of storage casks are as follows:

5.3 Model Specification Section 5.3, Rev. 1 Page 5.3-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS In all the array calculations it is additionally checked, at which distance a boundary of the restricted area has to be set. This boundary is regulated by the restricted boundary dose rate limit of 0.02 mSv/h (2 mrem/h) according to 10 CFR 20.1301 [3].

In order to speed the calculations up, a surface source file is utilised for the calculations of the storage cask arrays. For a generation of the surface source a uniform loading pattern (TR1) is used, because it generates the highest dose rates at the cask surface as well as at distances of 1 m and 2 m (see Section 5.4).

5.3.1 Description of the Radial and Axial Shielding Configurations The technical drawings of the storage cask (see Section 1.5) are used to create MCNP models used for the shielding calculations.

The elevation cut through the model is presented in Figure 5.1-2. The axial null of the scale corre-sponds to the cask bottom edge.

Being of low relevance for the shielding analysis, the screws, compression springs and gaskets are not modelled. When appropriate they are substituted by air, e.g. the heads of the lid bolting or the top of the basket, or by surrounding material, e.g. inside the lid.

For the model with fuel reconfiguration due to failure under normal (off-normal) conditions, an addi-tional fuel region with 3 % (10 %) of the source volume and strength according to [2] is considered at the canister bottom (see Figure 5.3-1). The rubbleised mixture of fuel and cladding is relocated within the basket cell, the mass packing fraction for the rubble is 0.58 [2]. Conservatively it is as-sumed that the rest of the fuel is not present anymore thus reducing the shielding effects. Solely the bottom nozzles are taken into account in the 10 %-rubble scenario (see Figure 5.3-1, right). Other-wise, the standard sources from the uniform loading pattern are attributed to the remaining fuel pellet 5.3 Model Specification Section 5.3, Rev. 1 Page 5.3-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 1 0CFR 2.390 Rev. 1 @GNS stack and contribute to the external dose rate with a strength of 97 % (90 % ). The end fittings and plenum springs from the standard NCT model contribute with a full strength.

After an accident, a 100 % fuel failure is assumed. As no damage of the basket structures occurs (see Chapter 3), the rubble consisting of fuel and cladding (the rest of the SNF structures is conser-vatively neglected) remains in the basket cell (see Figure 5.3-2).

Figure 5.3-1 Source location in fuel failure models - 3 % (normal conditions, left) and 10 %

(off-normal conditions, right)

Figure 5.3-2 Accident rubble of 100 % at the bottom (left) and at the top (right) of the cask 5.3 Model Specification Section 5.3, Rev. 1 Page 5.3-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The principal components of the shielding model are explained in the following subsections. The methods and main measurements are presented.

5.3.1.1 Spent Fuel All 69 SNF are placed into the corresponding basket cells. Conservatively, design is selected for the shielding model.

The fuel rods including cladding are modelled individually. The cross section through the fuel model in its lower part, where all the rods are present, is presented in Figure 5.3-3.

It is assumed that the SNF are complete and do not contain dummy rods. Possible loss of self-shielding due to possible absence of fuel rods is overcompensated by the actual loss of source strength at this location. The heavy metal mass of the modelled SNF amounts to 184.8 kg.

The end fittings enter the calculations in a simplified fashion as tie plates only. This approach is justified, since the source strength is scaled with the full mass of the end fittings. The plenum springs are modelled homogenised as steel pieces with reduced density (

Figure 5.3-3 Spent fuel model (dimensions in mm) 5.3 Model Specification Section 5.3, Rev. 1 Page 5.3-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 5.3.1.2 Basket The basket in the shielding model (see Figure 5.1-1) consists of and round segments made of aluminium (SB-209 5454). The mounting elements are left out and replaced by air. As discussed in Section 5.1 and shown in Figure 5.1-3, the thicknesses of the sheets are minimised and the outer diameter of the round segments is reduced according to the design tolerances.

5.3.1.3 Shielding Elements Shielding elements out of aluminium (SB-209 5454) are modelled as solid blocks (see Figure 5.3-4).

Figure 5.3-4 Shielding elements 5.3 Model Specification Section 5.3, Rev. 1 Page 5.3-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 5.3.1.4 Canister The elevation view of the canister is presented in Figure 5.3-5.

- Figure 5.3-5 Cask elevation cut (dimensions in mm) 5.3.1.5 Cask Body and Lid The cask body with lid is modelled with the unfavourable combination of the tolerances in radial direction.

Several so-called splitting levels (thin black lines in the model illustra-tions) are implemented into various components of the calculation model in order to increase its statistical performance using variance reduction techniques.

The LAP are modelled as solid pieces made of steel (upper trunnions) or DCI (lower tilting studs),

the corresponding flattenings of the cask body possess minimum dimensions.

5.3 Model Specification Section 5.3, Rev. 1 Page 5.3-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 The cask lid is modelled with the lid moderator plate mounted on its lower surface. The bolting of the lid is not modelled; the bolt heads are replaced with air.

5.3.1.6 Cooling Fins The cooling fins are modelled explicitly (see Figure 5.3-6). It was found that homogenisation of this areas does not necessarily lead to conservative evaluation.

Figure 5.3-6 Cooling fins (dimensions in mm) 5.3.1.7 Moderators The moderator rods made of polyethylene ( are modelled They are analysed in two different configurations, hot and cold (under accident conditions not modelled at all), the design geometry representing the situation shortly after loading and the state of thermal equilibrium (see Figure 5.1-4 and Figure 5.1-5).

Divergent from the design density of the resulting polyethylene at equilibrium is conser-vatively reduced according to the maximum temperature over the entire length of the rod (no axial temperature profile is assumed, see chapter 3). For the moderator plates (bottom and lid ones) only the density is reduced in equilibrium state, no expansion is implemented.

5.3.1.8 Environment and Detectors Depending on the modelling situation, the storage cask or array of casks is surrounded by sufficient amount of air to take scattering effects into account.

5.3 Model Specification Section 5.3, Rev. 1 Page 5.3-8

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS of the storage cask, 10 m surrounding air is selected. To analyse the storage site up to 500 m of air are considered.

The information about particular dose rates is gained from a mesh of detectors positioned all around the storage cask. A cylindrical mesh is utilised for the model with a single storage cask, while a rectangular raster is more convenient when exploring the dose rate field around the storage cask array. Besides this geometry-independent mesh of detectors, separate volumetric detectors are modelled in order to control the calculation process.

5.3.2 Shield Regional Densities Compositions and densities of the materials used in the shielding model are presented in Table 5.3-1. For the moderator rods two densities are given, the nominal one for the shielding configuration shortly after loading (cold), and the low one for the equilibrium configuration (hot). The steel specifi-cation for the retention ring ( is very similar to that of the for this

- reason no new material has been introduced.

The materials in Table 5.3-1 are arranged from inside to outside .

For the moderator material the temperature effects are studied as discussed in Section 5.1. -

. 1 5.3 Model Specification Section 5.3, Rev. 1 Page 5.3-9

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The design basis for the material data is given in Chapter 8.

5.3 Model Specification Section 5.3, Rev. 1 Page 5.3-10

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 5.3-1 Material properties in the shielding model 5.3 Model Specification Section 5.3, Rev. 1 Page 5.3-11

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 5.3-1 Material properties in the shielding model (cont.)

5.3 Model Specification Section 5.3, Rev. 1 Page 5.3-12

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @)GNS Table 5.3-1 Material properties in the shielding model (cont.)

List of References

[1] C.J. Werner (ed.), MCNP User's Manual - Code Version 6.2, LA-UR-17-29981, 2017

[2] NUREG-2224, Dry Storage and Transportation of High Burnup Spent Fuel Office of Nuclear Material Safety and Safeguards, November 2020

[3] Title 10 CFR Part 20 Standards for Protection Against Radiation U.S. Nuclear Regulatory Commission 5.3 Model Specification Section 5.3, Rev. 1 Page 5.3-13

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 5.4 Shielding Evaluation Prepared Reviewed 5.4 Shielding Evaluation Section 5.4, Rev. 1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS As discussed in Section 5.3 the MCNP6 [1] code is used for the shielding analysis. The cross section data are based on ENDF/B-VII data. The MCNP code system is benchmarked against experimental data for a broad spectrum of gamma [2] and neutron [3] problems. Described shielding problems cover a wide range of energies and material compositions and involve both scattering and deep penetration. A good agreement between measured and calculated values has been demonstrated for all the validation scenarios.

The dose rates are calculated using volumetric mesh tallies (f4), multiplied by an appropriate flux-to-dose-rate conversion factor, heavy metal mass of the SNF in the shielding model, and by the total source strength (per megagram (Mg) heavy metal) for every radiation source term. Since the mesh is stretched around the entire geometry of the storage cask, the locations of dose rate maxima are determined explicitly. Its use also allows for the evaluation of the wrapping dose rate distributions.

Volumetric detectors are used to determine the dose rates from the storage cask array at various distances.

The calculations are performed separately The elementary external dose rates for the evaluation of an arbitrary loading pattern provided In this evaluation, the loading patterns TR1 to TR3 are assessed.

The calculation with the bounding cask array are performed based on the generated surface source file from the individual storage cask loaded according to TR 1.

Each set of the MCNP calculations is foreseen with a message digest (md5) providing its unique identification. The summary of the cases is given in Table 5.4-1.

5.4 Shielding Evaluation Section 5.4, Rev. 1 Page 5.4-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @)GNS Table 5.4-1 Message digest overview As MCNP calculates fluxes, these values have to be converted into dose rate (dose) using corre-sponding response functions. The conversion of the spectral neutral and gamma flux density to the ambient equivalent dose is performed with the flux-to-dose-rate conversion factors according to ANSI/ANS-6.1.1-1977 [4]. The conversion factors are exhibited in Table 5.4-2 and Table 5.4-3 for gamma radiation and neutrons, respectively.

5.4 Shielding Evaluation Section 5.4, Rev. 1 Page 5.4-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 G Table 5.4-2 Conversion factors for gamma radiation Conversion Coefficient Conversion Coefficient Gamma Energy, Gamma Energy, (Including Quality Factor), (Including Quality Factor),

MeV 2 MeV 2 m5v/h/(y/cm *s) m5v/h/(y/cm *s) 0.01 3.96E-05 1.4 2.51E-05 0.03 5.82E-06 1.8 2.99E-05 0.05 2.90E-06 2.2 3.42E-05 0.07 2.58E-06 2.6 3.82E-05 0.1 2.83E-06 2.8 4.01E-05 0.15 3.79E-06 3.25 4.41E-05 0.2 5.01E-06 3.75 4.83E-05 0.25 6.31E-06 4.25 5.23E-05 0.3 7.59E-06 4.75 5.60E-05 0.35 8.78E-06 5 5.80E-05 0.4 9.85E-06 5.25 6.01E-05 0.45 1.08E-05 5.75 6.37E-05 0.5 1.17E-05 6.25 6.74E-05 0.55 1.27E-05 6.75 7.11E-05 0.6 1.36E-05 7.5 7.66E-05 0.65 1.44E-05 9 8.77E-05 0.7 1.52E-05 11 1.03E-04 0.8 1.68E-05 13 1.18E-04 1 1.98E-05 15 1.33E-04 Table 5.4-3 Conversion factors for neutrons Conversion Coefficient Neutron Energy, (Including Quality Factor),

MeV 2 mSv/h/(n/cm *s) 2.50E-08 3.67E-05 1.00E-07 3.67E-05 1.00E-06 4.46E-05 1.00E-05 4.54E-05 1.00E-04 4.18E-05 1.00E-03 3.76E-05 0.01 3.56E-05 0.1 2.17E-04 0.5 9.26E-04 1 1.32E-03 2.5 1.25E-03 5 1.56E-03 7 1.47E-03 10 1.47E-03 14 2.08E-03 20 2.27E-03 5.4 Shielding Evaluation Section 5.4, Rev. 1 Page 5.4-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The external radiation levels calculated with MCNP for an individual storage cask are modified by the corresponding statistical uncertainties provided by MCNP in the same tally. Thereby, instead of using a Gaussian error propagation, two standard deviations of MCNP are conservatively added to every single tallied value (penalising error propagation).

A sample input file for MCNP (neutron case) is presented in Section 5.5.

5.4.1 Dose Rates in the Near Field of the Cask Under Normal and Off-Normal Condi-tions The maximal values of the dose rates in the vicinity of the storage cask for all three loading patterns assuming that the storage cask has recently been loaded (cold shielding model) are presented in Table 5.4-4. The contributions of the gamma and neutron radiation from fuel and fuel hardware are presented as well. The uncertainties from the penalising error propagation are included.

Among three loading patterns analysed, the uniform one (TR1) is the most demanding from the shielding point of view. It generates the maximum dose rate at the storage cask surface as well as at 1 m and 2 m from the cask surface.

The corresponding dose rate wrap ups on the surface of the storage cask loaded with the fuel ac-cording to loading pattern TR1 are displayed in Figure 5.4-1. The distributions for other two loading patterns are similar, solely the absolute values change.

While one can still visually see the contributions from the nozzles and plena in the dose rate distribution at 1 m from the cask (see Figure 5.4-2, left), they nearly disappear at 2 m from the cask (see Figure 5.4-2, right).

Figure 5.4-3 demonstrates fractional standard deviations - Gaussian error propagation - of the cal-culated dose rate for TR1 at the surface of the cask and at a distance of 2 m from the cask.

The dose rates on the lid side of the storage cask are much lower than on the shell side.

In 1 m from the cask lid the dose rate distribution is close to be uniform, (see Figure 5.4-4, right).

• 5.4 Shielding Evaluation Section 5.4, Rev. 1 Page 5.4-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 5.4-4 Maximum external dose rates for the cold storage cask 5.4 Shielding Evaluation Section 5.4, Rev. 1 Page 5.4-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 5.4-1 Dose rate distributions (TR1, cold model) at the surface of the storage cask in mSv/h: fuel gamma (top left), fuel hardware gamma (top right), neutron (bottom left) and total (bottom right) 5.4 Shielding Evaluation Section 5.4, Rev. 1 Page 5.4-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Figure 5.4-2 Total dose rate distributions (TR1, cold model) at 1 m (left) and 2 m (right) from the storage cask in mSv/h Figure 5.4-3 Fractional standard deviation distributions at the surface of the storage cask (left) and in 2 m from the cask surface (right) for the cold model in mSv/h 5.4 Shielding Evaluation Section 5.4, Rev. 1 Page 5.4-8

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 G Figure 5.4-4 Total dose rate distributions (TR1, cold model) at the lid surface of the storage cask (left) and in 1 m (right) from the storage cask lid in mSv/h Table 5.4-5 provides maximum external dose rates for the storage cask under regular storage con-ditions, when the thermal equilibrium is attained and the moderator material expands (hot model).

Fuel failure under normal and off-normal conditions are evaluated for TR1 only (see Table 5.4-6),

since this loading pattern delivers the highest dose rates. Two cases according to NUREG-2224 [5]

are considered: 3 % failed fuel forming a rubble under normal conditions and 10 % fuel failure under*

off-normal conditions of storage. Fuel reconfiguration is considered for loading after 20 years of dry storage [5], the source terms for this particular scenario have been adjusted accordingly by selecting maximum source terms among all possible combinations of fuel assemblies after 20 years (see Ta-ble 5.2-2).

The redistribution of the fuel does not affect the shell dose rate in a negative way. The 5.4 Shielding Evaluation Section 5.4, Rev. 1 Page 5.4-9

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS neutron dominated maximum dose rates (at the surface and 2 m from the surface of the storage cask) are lower than those from the failure unaffected shielding models.

Table 5.4-5 Maximum external dose rates for the hot storage cask In total, sufficiently low dose rates are established in the vicinity of the storage cask in accordance with ALARA practices. Important to mention that the bounding dose rates are presented. In reality, lower values are expected.

5.4 Shielding Evaluation Section 5.4, Rev. 1 Page 5.4-10

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Table 5.4-6 Maximum external dose rates assuming fuel failure during storage 5.4.2 Dose Rates in the Near Field of the Cask under Accident Conditions The calculated dose rates for the storage cask after design basis accident are presented in Table 5.4-7 for all three loading patterns. Additionally, a 100 % fuel reconfiguration shifted towards the e, bottom and the top of the storage cask is summarised in Table 5.4-8.

For this reason, the regular accident cask model with upright oriented casks is selected to evaluate the dose rates at the site boundary.

5.4 Shielding Evaluation Section 5.4, Rev. 1 Page 5.4-11

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 5.4-7 Maximum external dose rates for design basis accident 5.4 Shielding Evaluation

• Section 5.4, Rev. 1 Page 5.4-12

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Table 5.4-8 Maximum external dose rate for design basis accident with complete fuel dam-age 5.4.3 Long Term Radiation Load on the Cask Materials Besides the requirements on the external dose rates or dose equivalents, there are further demands on the radiation load on the storage cask and its materials (see chapter 8). Among other things, it is to be demonstrated that the materials of the storage cask are able to tolerate the radiation load on the long term. For this purpose, the energy doses and neutron flux densities are determined for different components of the storage cask for the three loading patterns (TR1 to TR3). The resulting values are the maxima of the three individual values determined for each loading pattern. The results of the calculations are presented in Table 5.4-9.

5.4 Shielding Evaluation Section 5.4, Rev. 1 Page 5.4-13

Non.:Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Table 5.4-9 Energy doses and neutron fluxes for different storage cask components 5.4.4 Dose Rates and Annual Doses at the Storage Site Regarding the horizontal orientation of the storage cask relative to the nearest site boundary, it is not important, which side of the cask is directed towards this boundary. The azimuthal distribution of the dose rate at 2 m from the storage cask (see Figure 5.4-2) is relatively smooth without stringent maxima. This statement is supported by the range dependence of the annual dose from the standalone cask (see Figure 5.1-7),

- For the calculation of the annual dose an exposure time of 8766 hours0.101 days <br />2.435 hours <br />0.0145 weeks <br />0.00334 months <br /> is assumed.

The neutron dose equivalent is mainly due to the spon-taneous fission, the major gamma radiation contribution comes from the fuel gammas at energies Table 5.4-10 Relative contributions to the annual dose at 100 m from the storage cask 5.4 Shielding Evaluation Section 5.4, Rev. 1 Page 5.4-14

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Clearly seen that at the minimum distance to the controlled area boundary of 100 m according to 10 CFR 72.106 the annual dose equivalent limit to public of 0.25 mSv according to 10 CFR 72.104 is In case the larger distance cannot be confronted with, a deployment of a storage building made of e.g. concrete is one of the possible solutions.

Assuming a bounding array - storage casks (see Figure 5.1-6), one can determine a minimum distance to public needed to meet the requirements of 10 CFR 72.104. This point is located in the centre of the long side of the array from the cask array (see Figure 5.1-8). An area of roughly has to be classified as restricted to comply with the requirements of 10 CFR 20 [6]. For easier understanding the line representing a dose rate limit I

- of 0.02 mSv/h according to 10 CFR 20.1301 is scaled to the annual dose equivalent.

In case it is necessary to meet the requirements of 10 CFR 72.104 at the minimum distance to public of100m*-isrequired around the entire array of the storage casks (see Figure 5.1-10). For this scenario, the annual dose equivalent to any real individual beyond the controlled area would not exceed 0.25 mSv (10 CFR 72.104).

For design basis accident, it is assumed that-storage casks have lost their neutron moderators completely. The upright arrangement as discussed in Section 5.4.2 is retained. For the duration of accident conditions 30 days are implied. The resulting annual dose equivalents are presented in Figure 5.1-11. It can be clearly seen thatthe limiting total effective dose equivalent of 50 mSv ac-cording to 10 CFR 72.106 is* safely complied with.

5.4 Shielding Evaluation Section 5.4, Rev. 1 Page 5.4-15

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 List of References

[1] C.J. Werner (ed.), MCNP User's Manual - Code Version 6.2, LA-UR-17-29981, 2017

• [2] Daniel J. Whalen, David E. Hollowell, and John S. Hendricks, Photon Benchmark Prob-lems, LA-12196, Los Alamos National Laboratory, 1991

[3] Daniel J. Whalen, David A. Cardon, Jennifer L. Uhle, and John S. Hendricks, Neutron Benchmark Problems, LA-12212, Los Alamos National Laboratory, 1991

[4] American Nuclear Society. Working Group ANS-6.1.1; American National Standards I nsU-tute. American national standard neutron and gamma-ray flux-to-dose-rate factors, La Grange Park, Ill.: The Society, 1977

[5] NUREG-2224, Dry Storage and Transportation of High Burnup Spent Fuel Office of Nuclear Material Safety and Safeguards, November 2020

[6] Title 10 CFR Part 20 Standards for Protection Against Radiation U.S. Nuclear Regulatory Commission 5.4 Shielding Evaluation Section 5.4, Rev. 1 Page 5.4-16

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev.1 5.5 Appendix Name, Function Date Signature

~ -~ -~---- -

Prepared Reviewed 1

5.5 Appendix Section 5.5, Rev. 1 Page 5.5-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Appendix 5-1: Computer Program Descriptions Appendix 5-2: Validation of the ORIGAMI calculations Appendix 5-3: Shielding Model for Cask Loading Unit Appendix 5-4: Sample Input File for MCNP 5.5 Appendix Section 5.5, Rev. 1 Page 5.5-2

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 Appendix 5-1 Computer Program Descriptions MCN P [1] is a fully three-dimensional program that describes the coupled transport of neutrons and gammas. MCNP works according to the Monte Carlo technique used to simulate the histories of individual particles from their origin to the point of their absorption or when they leave the volume of interest. A sufficient number of simulated particle histories allows for the description of the resulting location and energy-dependent particle densities and other physical quantities even for complex geometry arrangements. The results of Monte Carlo simulations are expected values of physical quantities according to their stochastic nature. The information for a specific transport problem is gained using detectors allowing for the tallying of the particle characteristics at the respective loca-tion. In this report, geometry-independent meshes of volume detectors are used, where the particle flow was determined using track length estimation.

The SCALE code system (actualwork using SCALE 6.2 [2]) is developed by the Oak Ridge National Laboratory (ORNL) on behalf of the Nuclear Regulatory Commission (NRC). SCALE is a modular system used for nuclear analyses and consists of the functional and control modules as well as of the nuclear cross section libraries. The cross section libraries are application dependent and can either be based on the energy groups or on the continuous energy distribution. The former ones are microscopic and have to undergo a problem and geometry specific preparation carefully handling the resonance regions. These preparations are fully automated and are controlled by the functional modules BONAMI [3] (region of not resolved resonances) and CENTRM [4] / PMC [5] (region of resolved resonances).

In the sequence with the TRITON [6] control module it is made possible to perform the analysis of the fuel assembly transverse cross section using the two dimensional (2D) transport calculations by the NEWT [7] functional module to obtain local neutron spectra. The subsequent burn-up and decay calculations are.performed by the ORIGEN-S [8] functional module. The cross sections from the 252 group libraries based on ENDF/B-Vll.1 are employed.

ORIGAMI [9] computes detailed isotopic compositions, decay heat, and gamma and neutron radia-tion source spectra for light water reactor assemblies containing U02 fuel by using the ORI GEN code with pre-generated libraries, for a specified assembly power distribution. ORIGAMI performs ORIG EN burn-up calculations for each of the specified power regions to obtain the spatial distribution of isotopes in the burned fuel.

Page 1 of2

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390

[1] C.J. Werner (ed.), MCNP User's Manual- Code Version 6.2, LA-UR-17-29981, 2017

[2] SCALE Code System, Version 6.2.2. B. T. Rearden, M.A. Jessee, Eds. ORNL/TM-200~39, 2017 -

[3] BONAM I: Resonance Self-Shielding by the Bondarenko Method in: SCALE Code System, Version 6.2.2, ORNL/TM-2005/39, 2017

[4] CENTRM: A Neutron Transport Code for Computing Continuous-Energy Spectra in General One-Dimensional Geometries and Two-Dimensional Lattice Cells in: SCALE Code System, Version 6.2.2, ORNL/TM-2005/39, 2017

[5] PMC: A Program to Produce Multigroup Cross Sections using Pointwise Energy Spectra from CENTRM in: SCALE Code System, Version 6.2.2, ORNL/TM-2005/39, 2017

[6] TRITON: A Multipurpose Transport, Depletion, and Sensitivity and Uncertainty Analysis Module in: SCALE Code System ORNL/TM-2005/39, Version 6.2.2, 2017

[7] NEWT: A New Transport Algorithm for Two-Dimensional Discrete Ordinates Analysis in Non-Orthogonal Geometries in: SCALE Code System, Version 6.2.2, ORNL/TM-2005/39, 2017

[8] ORIGEN: Neutron Activation, Actinide Transmutation, Fission Product Generation, and Ra-diation Source Term Calculation in: SCALE Code System, Version 6.2.2, ORNL/TM-2005/39, 2017

[9] ORIGAMI: A Code for Computing Assembly lsotopics with ORIGEN in: SCALE Code Sys-tem ORNL/TM-2005/39, Version 6.2.2, 2017 Page 2 of2

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 Appendix 5m2 Corroboration of the ORIGAM.1 calculations Page 1 of 4

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 Appendix 5-3 Shielding Model for Cask Loading Unit This supplement is focussed on describing the shielding model of the CASTOR geo69 cask loading unit (CLU) comprising the transfer cask, the transfer lock (or port) and further not explicitly modelled equipment. A shielding evaluation of the CLU is only of interest because of the occupational expo-sures of the personal in view of ALARA practices, In general, the shielding analysis of the CASTOR geo69 CLU represented by the transfer cask only is very similar to that of the storage cask. All the analysis methods are identical with those described in chapter 5. The vast part of the shielding model including contents, basket and canister is identical.

The only component which differs is the cask body itself. The transfer cask is designed to be a

- lightweight loading unit with a water neutron shield needed to facilitate To protect the personnel from gamma radiation a lead gamma shield is designed.

A detailed shielding model of the transfer cask is presented in Figure 5.5-2. The cask itself is con-servatively modelled without transfer cask lid. The basket is drained, both water chambers are filled with water.

Conservatively, the contents of the canister are assumed to be loaded according to the uniform loading pattern (TR 1). Typical dose rate distributions on the shell surface of the transfer cask and in 1 m from the cask are displayed in Figure 5.5-3.

- Regarding handling of the canister during initial loading into the storage cask, the configuration when the storage cask and the transfer cask are connected via transfer port (see Figure 5.5-4) is very important (see chapter 11).

The maximum external dose rates would be ob-served, when the canister is positioned in the middle of its travel. The time it takes to perform the whole reloading action between transfer and storage cask including cask closure is limited, never-theless the exposure of the personnel during this time has to be minimised. The individual handling operations complemented with corresponding dose levels are summarised in chapter 11.

Page 2 of 4

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 Figure 5.5-2 Shielding model of the transfer cask Figure 5.5-3 Dose rate distributions (in mSv/h) on the shell surface (left) and in 1 m distance from the transfer cask (right)

Page 3 of 4

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 Figure 5.5-4 Shielding model of the canister transfer Page 4 of 4

---~-------------------------,

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 Appendix 5-4 Sample Input File for MCNP Page 1 of 23

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev.a @GNS 6 Criticality Evaluation 6.0 Overview Prepared Reviewed 6.0 Overview Section 6.0, Rev. 0 Page 6.0-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 This chapter describes the proof of subcriticality for the content described in subsection 1.2.3 in accordance with requirements from 10 CFR 72.

The maximum values for the effective neutron multiplication factor k and the calculational bias with its uncertainty ~ku with a 95 % probability at a 95 % confidence level fulfil the acceptance criteria k + ~ku < 0.95 for all considered cask loadings and demonstrate the compliance with requirements for normal, off-normal and accident conditions during handling, packaging, transfer and storage, as required by§ 72.124 and§ 72.236.

6.0 Overview Section 6.0, Rev. 0 Page 6.0-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 6.1 Discussion and Results Name, Function Date Signature Prepared Reviewed 6.1 Discussion and Results Section 6.1, Rev. 0 Page6.1-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS The cask CASTOR geo69 (referred to as "storage cask") consists of the cask body (cast iron, polyethylene) with bolted lid system (stainless steel, polyethylene), as described in section 1.2, and is used to accommodate the cylindrical canister (stainless steel) with bolted lid system (stainless steel) during storage. The fuel basket is designed to accommodate up to 69 BWR FA and is placed into the canister.

During the loading and unloading of the FA into/out of the canister and for the transfer of the canis-ter into/out of the storage cask a transfer cask, as described in section 1.2, is used. The body of the transfer cask consists of stainless steel and lead shields combined with two independent water chambers for optimized protection against gamma and neutron radiation.

For ensuring the criticality safety of the storage and transfer casks, a combination of the following design measures is used:

- limitation of the fissile content of the fuel,

- geometrical positioning of the FA within a fuel basket and

- fixed neutron absorbing structures in the fuel basket.

The following conservative assumptions are made in the criticality safety analysis:

The applicable codes and standards are summarized as follows:

10 CFR 72, ANSI/ANS-8.1-2014 [1].

6.1 Discussion and Results Section 6.1, Rev. 0 Page 6.1-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 The storage cask is designed to exclude the water leakage into the canister cavity under normal, off-normal and accident conditions during storage as shown in the respective analyses in Section 3.5 and 3.6. Due to a very low reactivity of dry fuel, the behavior of the spent fuel as a result of accident conditions during dry storage (ACS) and the storage periods beyond 20 years do not need to be explicitly evaluated and are bounded by the reactivity of the fully flooded cask with pure unborated water, as assumed in the bounding model for normal conditions of storage (NCS), de-scribed in subsections 6.3.1.2 and 6.3.2.

As shown in chapters 3 and 12, the off-normal and accident conditions during the loading and un-loading have no impact on the structure of the canister, the fuel basket and the content. The nor-mal, off-normal and accident conditions during handling, packaging and transfer are thus also bounded by the reactivity of the fully flooded cask with pure unborated water, as assumed in the bounding model for NCS, described in subsections 6.3.1.2 and 6.3.2.

As the outer shell structures of the storage and transfer casks provide different neutron reflection conditions, the full neutron reflection is separately determined for fully flooded storage and transfer casks.

The final results of the criticality evaluation as well as the corresponding internal moderation and external moderation and reflection conditions are provided in the summary Table 6.1-1 for all in-vestigated FA types (FA no.). The results are obtained under bounding conditions for NCS, as dis-cussed in subsections 6.4.2.1 through 6.4.3, and include the maximum effective neutron multiplica-tion factors k and the calculational bias with its uncertainty ~ku. All contributions to the final results are determined with a 95 % probability at a 95 % confidence level, as discussed in subsec-tion 6.5.2.1 and Appendix 6-1.

- For reference purposes a unique alphanumeric identification number (calc. ID) is assigned to each calculation.

The following cases are evaluated in the criticality safety analysis:

- infinite array of flooded storage casks (reference case for NCS),

- infinite array of dry storage casks,

- single, fully reflected storage cask,

- single, fully reflected transfer cask.

The evaluated cases are bounding for normal, off-normal and accident conditions during handling, packaging, transfer and storage.

6.1 Discussion and Results Section 6.1, Rev. 0 Page 6.1-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 Table 6.1-1 Summary table of criticality evaluation Infinite array of flooded Infinite array of dry Case storage casks.

storage casks (bounding model for NCS)

Internal 100 % (Water) 0 % (Void)

FA no. moderation External moderation 0% (Void) 0 % (Void) and reflection Fuel type k+Aku calc. ID k +Aku calc. ID 1 GE 8x8-1 0.84386 20tZ09ny02 0.35051 20EF09bR02 2 GE 8x8-2 0.85061 20da09vS02 0.34374 20hN09sh02 3 SPC 8x8-2 0.84625 20rq09Tj02 0.33861 20DB09dw02 4 GE9B 8x8 0.88137 20ll09Vo02 0.36399 20JJ09zw02 5 GE12 LUA 0.90514 20Xb09ei02 0.39113 20La09rj02 6 ATRIUM-10A 0.93729 20nS09hZ02 0.42457 20rk09gn02 6.1 Discussion and Results Section 6.1, Rev. 0 Page 6.1-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @)GNS Table 6.1-1 Summary table of criticality evaluation (continued)

Single, fully reflected Single, fully reflected Case storage cask transfer cask Internal 100. % (Water) 100 % (Water)

FA no. moderation External moderation 100 % (Water) 100 % (Steel) and reflection Fuel type k+Aku calc. ID k+Aku calc. ID 1 GE 8x8~1 0.84387 20hr09qR02 0.84410 21Vu02YZ08 2 GE 8x8-2 0.85034 20bM09JV02 0.85044 21Iv02wr08 3 SPC 8x8-2 0.84629 20AV09iY02 0.84641 21HQ02yr08 4 GE9B 8x8 0.88145 20ed09nP02 0.88149 21oa02SS08 5 GE12 LUA 0.90527 20Np09QB02 0.90541 21NW02dx08 6 ATRIUM-10A 0.93725 20Lf09eS02 0.93740 21jq02Sv08 As provided in Table 6.1-1, the results demonstrate the compliance with requirements for normal, off-normal and accident conditions during handling, packaging, transfer and storage, as required by§ 72.124 and§ 72.236.

List of References

[1] ANSI/ANS-8.1-2014: Nuclear Criticality Safety in Operations with Fissile Material Outside Reactors.

6.1 Discussion and Results Section 6.1, Rev. 0 Page 6.1-5

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev.a @GNS 6.2 Spent Fuel Loading Name, Function Date Signature Prepared Reviewed 6.2 Spent Fuel Loading Section 6.2, Rev. 0 Page 6.2-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 The criticality safety analysis is performed for the content described in subsection 1.2.3. The calcu-lation model takes upper limits for the use of unirradiated fissile material (max. fuel density, max.

fuel enrichment) in the FA into account. All investigated fuel types are in a solid metal dioxide form (U02) .

6.2 Spent Fuel Loading Section 6.2, Rev. 0 Page 6.2-2

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev.a @GNS 6.3 Model Specification I

Name, Function I Date Signature Prepared I

Reviewed I

6.3 Model Specification Section 6.3, Rev. 0 Page 6.3-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 6.3.1 Description of Calculational Model 6.3.1.1 General Considerations This subsection addresses the assessment methodology used to evaluate criticality of the cask.

The CASTOR geo69 cask is designed for the accommodation of different types of BWR spent FA.

To take a large number of relevant model parameters into account, for example the fabrication tolerances and the uncertainties in material compositions of FA or basket structures, the calcula-tions of the neutron multiplication factors are performed using bounding models.

The analyzed FA represent square-pitched lattices of fuel rods. The qualitatively same impact of

- some model parameters on the system reactivity for all FA no. can be assumed. For example, the decrease of the moderator density or the increase of the absorber concentration of the fuel basket structures will reduce the reactivity of all FA no.

However, the distinctions of different FA no. lead to quantitative differences in the reactivity impact of the same model parameter, i.e. the criticality calculations must be performed for each FA type and each cask loading pattern separately. For such calculations the same set of conservative model parameters can be applied.

Based on these assumptions, bounding calculation models (bounding models) for NCS and ACS are derived as described below and shown in Figure 6.3-1.

Basic t----~

Bounding Bounding model model for NCS model for ACS Figure 6.3-1 Development of bounding models The bounding model for NCS is determined using sensitivity analyses based on a basic.calculation model (basic model), as described in subsection 6.4.2.1 . The basic model is based on the storage cask and content descriptions from section 1.2 and contains either nominal, representative or ex-pected to be bounding values for geometry, material compositions and densities of the cask, fuel basket and content.

Based on this basic model sensitivity analyses are perf9rmed for a full homogeneous cask loading with a reference FA (ATRIUM-10A) only, as described in subsection 6.4.2.1. The sensitivity ana-lyses bound all fabrication tolerances, uncertainties in material compositions, axial and radial FA displacements within the basket receptacles as well as optimum moderation conditions. The effect 6.3 Model Specification Section 6.3, Rev. 0 Page 6.3-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS of a variation of a certain parameter is described as a deviation {f1k in pcm, i.e. 10-5 ) of the calcu-lated k-value to the k-value of the basic model. If the variation of a certain parameter leads for the reference FA to an increase in reactivity, the same behavior can be expected for all the other FA no. The corresponding basic model of each FA no. is then adapted according to the possible toler-ance range of this parameter.

With the above mentioned approach a bounding model for NCS for each FA no. is developed. The bounding model for NCS is described in subsections 6.3.1.2 and 6.3.2.

On the basis of this bounding model for NCS additional proof of its conservativity as well as analy-ses concerning reflection conditions (separately for storage and transfer casks) and the cask be-havior under ACS are usually performed leading for each FA no. to a bounding model for ACS.

The proof of the conservativity of the bounding model for NCS and the analysis of the reflection conditions are provided in subsection 6.4.2.3.

As the damage of the cask wall and bottom as well as the loss of integrity of the cask and canister lid systems under ACS are excluded, as shown in chapters 3 and 12, no explicit bounding model for ACS is developed. ACS are bounded by the reactivity of the fully flooded cask with pure un-borated water, as assumed in the bounding model for NCS, described in subsections 6.3.1.2 and 6.3.2.

Based on the bounding model for NCS and taking into account the full neutron reflection for stor-age and transfer casks, the compliance with the requirements of § 72.124 and § 72.236 is demon-strated in subsection 6.4.3.

Every possible mixed loading of the cask with FA no. for which safe subcriticality can be demon-strated is bounded by the maximum effective neutron multiplication factor k of a full homogeneous cask loading with one of these FA no.

6.3.1.2 Model Configuration In the bounding calculation model for NCS the cylindrical canister and cask body with the corre-sponding bottom and top (lid system) structures are considered. The model represents an infinite array of densely packed and fully flooded storage casks. The volume outside the casks is filled with void.

The calculation model does not include an explicit consideration of neither the sealing system nor other outer parts, e. g. moderator rods inside the cask body, cooling fins, trunnions or impact limit-ers, as well as the axial gaps between the canister and the cask.

6.3 Model Specification Section 6.3, Rev. 0 Page 6.3-3

Non-Proprietary Version 1014-SR-00002 Rev. 0 Proprietary Information withheld per 10CFR 2.390 s

The radial and axial cross sections of the bounding model for NCS are shown in Figure 6.3-2 and Figure 6.3-3.

Figure 6.3-2 Radial cross section of the bounding model for NCS (dimensions in mm) 6.3 Model Specification Section 6.3, Rev. 0 Page 6.3-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Figure 6.3-3 Axial cross section of the bounding model for NCS (dimensions in mm) 6.3 Model Specification Section 6.3, Rev. 0 Page 6.3-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Figure 6.3-4 Cross sections of the analyzed FA types

• 6.3 Model Specification Section 6.3, Rev. 0 Page 6.3-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 Figure 6.3-5 Radial cross section of the bounding model for NCS for a 6.3 Model Specification Section 6.3, Rev. 0 Page 6.3-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS

- Figure 6.3-6 Axial cross section of the bounding model for NCS for a Compared to the basic model used as a starting point for the development of the bounding model for NCS, as described in subsection 6.4.2.1, the following modifications are incorporated in the bounding model for NCS (cf. subsection 6.4.2.2):

- bounding material densities and compositions,

- maximum clad inner diameter,

- minimum clad outer diameter,

- minimum inner dimension of basket receptacles,

- radial displacement of all FA towards the center of the fuel basket.

6.3 Model Specification Section 6.3, Rev. 0 Page 6.3-8

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS 6.3.1.3 Transfer Cask In the calculation model of the transfer cask, which is placed instead of the storage cask, only the relevant structures of the transfer cask (bottom, lid and shell) are modelled. The calculation model of the transfer cask is used only for the evaluation of reflection conditions during handling opera-tions, as described in subsection 6.4.2.3.3.

The radial and axial cross sections of the calculation model for the transfer cask are shown in Fig-ure 6.3-7 and Figure 6.3-8.

Figure 6.3-7 Radial cross section of the calculation model for the transfer cask (dimen-sions in mm) 6.3 Model Specification Section 6.3, Rev. 0 Page 6.3-9

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Figure 6.3-8 Axial cross section of the calculation model for the transfer cask (dimensions inmm) 6.3.2 Cask Regional Densities The bounding material densities and compositions are determined using the sensitivity analysis, as described in subsection 6.4.2.1, and are listed in Figure 6.4-1 and Table 6.4-2.

The fuel basket contains fixed neutron absorbing structures (boronated basket sheets). The pres-ence and the proper distribution of boron in the basket sheets at time of fabrication are ensured by quality measures. Loss of absorber material as a result of physical, chemical and corrosive mech-anisms can be excluded according to chapter 3.

6.3 Model Specification Section 6.3, Rev. 0 Page 6.3-10

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS The loss of absorber material through direct neutron absorption (and, thus, transmutation to a non-absorbing isotope) is inconsequential because any measurable depletion would take millions of years as a result of the extremely low neutron flux levels in a subcritical system.

6.3 Model Specification Section 6.3, Rev. 0 Page 6.3-11

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev.a @GNS 6.4 Criticality Evaluation Prepared Reviewed 6.4 Crftlcallty Evaluatron Section 6.4, Rev. 0 Page 6.4-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 G S 6.4.1 Calculational or Experimental Method The criticality calculations for the CASTOR geo69 are performed using the 3-dimensional Monte-Carlo program KENO-VI from the SCALE 6.2 code package [1]. The neutron multiplication factors are calculated using the 252-group neutron cross sections based on the ENDF/B-VI 1.1 evaluation (V7.1-252n, T = 293 K).

In the criticality calculations the number of neutron generations with 20,000 neutrons per genera-tion as well as the number of first neutron generations to be skipped is chosen in such a way, that the standard deviation of the calculated neutron m"ultiplication factors is below 20 pcm .

.The criticality safety analysis as well as the validation of the used calculation system, i.e. the Mon-te-Carlo program together with the cross-section library, is performed under the same or compara-

, ble boundary conditions.

6.4.2 Fuel Loading or other Contents Loading Optimization Based on the basic model, the sensitivity analysis for the full cask loading with the most reactive FA type .(ATRIUM-10A) is performed and bounding model for NCS is derived, as described in sec-tion 6.3.

6.4.2.1 Development of the Bounding Model for NCS The sensitivity analyses for the development of the bounding model for NCS include the following evaluations:

- material densities and compositions,

- pellet diameter,

- length of the active zone, 6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS

- clad inner and outer diameter,

- thickness of the FA internals,

- inner and outer dimensions of the FA channel,

- axial and radial displacement (including clustering) of the FA,

- axial position of part-length fuel rods,

- FA orientation within basket receptacles,

- temperature,

- dimension of the basket receptacles,

- thickness of the basket sheets,

- thickness of the outer sheets,

- canister dimensions,

- radial displacement of the canister within the cask cavity,

- outer boundary conditions for the cask,

- partial vertical and horizontal flooding of the cask cavity,

- moderator rod material in the cask wall,

- shielding element structure.

The results of the analysis are presented in subsections 6.4.2.1.1 through 6.4.2.1.22.

6.4.2.1.1 Material Densities arid Compositions 6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 Table 6.4-1 Determination of material densities for the bounding model 6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Table 6.4-2 Determination of material compositions for the bounding model 6.4.2.1.2 Pellet Diameter The pellet diameter is parametrically varied within a range of -200 µm/+150 µm at a fixed fuel den-sity of -The results given in Table 6.4-3 and Figure 6.4-3 show no statistically significant influence of this pa-rameter on the reactivity.

As a result, no changes need to be considered in the bounding model.

6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS 6.4.2.1.3 Length of Active Zone The length of the active zone is parametrically varied between +/-20 mm, that corresponds to a height of about two single fuel pellets. The results provided in Table 6.4-3 and Figure 6.4-3 show no statistically significant influence of this parameter on the reactivity.

As a result, no changes need to be considered in the bounding model.

6.4.2.1.4 Clad Inner and Outer Diameter The clad inner diameter is varied between -150 µm/+200 µm; the clad outer diameter between

+/-200 µm. The results for both variations are given in Table 6.4-3 and Figure 6.4-3. It can be seen that the clad thickness has a significant influence on the reactivity.

As a result, the maximum inner and minimum outer clad diameters are considered in the bounding model.

6.4.2.1.5 Thickness of the Water Channel The thickness of the water channel of the reference FA (ATRIUM-10A) is parametrically varied by the variation of the outer channel dimension between +/-200 µm at a fixed inner channel dimension.

The results given in Table 6.4-3 and Figure 6.4-3 show a significant increase of reactivity at smaller thicknesses of the water channel.

As a result, the minimum thicknesses of the FA internals, as already implemented in the basic model, are considered in the bounding model.

6.4.2.1.6 Inner and Outer Dimensions of the FA Channel The inner and outer dimensions of the FA channel are separately varied between +/-200 µm, result-

- ing in a variation of the wall thickness of the FA channel. The results of criticality calculations are given in Table 6.4-3 and Figure 6.4-3. It can be seen that the wall thickness of the FA channel has no statistically significant influence of the reactivity.

As a result, no changes need to be considered in the bounding model.

6.4.2.1.7 Axial Displacement of the FA The axial displacement of the FA is varied between +/-40 mm. The results provided in Table 6.4-3 and Figure 6.4-3 show no significant influence on reactivity.

As a result, no changes need to be considered in the bounding model.

6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-6

Non-Proprietary Version 1014-SR-00002 Rev. 0 Proprietary information withheld per 10CFR 2.390

@G S 6.4.2.1.8 Radial Displacement of the FA The reactivity impact of the radial displacement of the FA within the fuel basket is parametrically investigated via a simultaneous shift of all FA towards the basket center or towards the basket periphery The results of these calculations are shown in Table 6.4-3 and Figure 6.4-3. It can be seen that the displacement towards the basket center leads to a significant increase of reactivity.

As a result, the maximum possible displacements of the FA towards the basket center are consid-ered in the bounding model.

6.4.2.1.9 Clustering

- Additionally to the radial displacement of all FA towards the basket center or the basket periphery, as discussed in subsection 6.4.2.1.8, the building of FA conglomerates (clustering) within the fuel basket is investigated. The selected configuration is shown in Figure 6.4-1 and results in a signifi-cant decrease of reactivity (see Table 6.4-3) compared to the reference (basic) model with FA cen-tered within the basket receptacles.

As a result, the maximum possible displacements of the FA towards the basket center are consid-ered in the bounding model, as discussed in subsection 6.4.2.1.8.

Figure 6.4-1 FA clustering 6.4.2.1.10 Axial Position of Part-Length Fuel Rods The axial position of the part-length fuel rods is varied up to 40 mm beginning at the lower edge of the full-length fuel rods. The results presented in Table 6.4-3 and Figure 6.4-3 indicate the reactivi-ty decrease by the axial shift of the part-length fuel rods towards the center of the active zone.

6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 As a result, no axial displacement of the part-length fuel rods, as already implemented in the basic model, is considered in the bounding model.

6.4.2.1.11 FA Orientation within Basket Receptacles During the basket loading with FA different orientations of the FA internals (water rods/channel) relative to the neighboring FA are possible. This effect is investigated via a fuel lattice rotation of 90°, 180° and 270° for exemplary selected FA in the calculation model. The investigated configura-tion is presented in Figure 6.4-2 and shows no statistically significant influence (see Table 6.4-3) of the FA orientation within the basket receptacles on the reactivity.

As a result, the same orientation of all FA, as already implemented in the basic model, is consid-ered in the bounding model.

Figure 6.4-2 FA orientation 6.4.2.1.12 Temperature The temperature of the fuel and the structure materials is parametrically varied between 0 °C and 170 °C. The results presented in Table 6.4-3 and Figure 6.4-3 confirm the reactivity decrease by increased temperature due to the Doppler broadening.

As a result, the room temperature (20 °C), as already implemented in the basic model, is consid-ered in the bounding model.

6.4.2.1.13 Dimension of the Basket Receptacles The inner dimension of the basket receptacles is parametrically varied within the possible tolerance range between at a fixed thickness of the basket sheets. The results provided in 6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-8

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Table 6.4-3 and Figure 6.4-3 show a significant reactivity increase by decreased basket recepta-cles.

As a result, the minimum possible inner dimension is applied to all basket receptacles in the bounding model.

6.4.2.1.14 Thickness of the Basket Sheets The thickness of the borated aluminium sheets of the basket is parametrically increased up to

- As shown in Table 6.4-3 and Figure 6.4-3, the increased thickness of the basket sheets leads to a significant decrease of reactivity.

As a result, the minimum possible thickness of the basket sheets, as already implemented in the basic model, is considered in the bounding model.

6.4.2.1.15 Thickness of the Outer Sheets The thickness of the outer basket sheets is parametrically increased up to - T h e results given in Table 6.4-3 and Figure 6.4-3 show no statistically significant influence of this parameter on the reactivity.

As a result, no changes need to be considered in the bounding model.

6.4.2.1.16 Canister Dimensions The reactivity impact of the canister dimensions, such as inner and outer diameter (1.D. and O.D.)

as well as the wall thickness (T), is investigated via calculations for all possible combinations of minimum and maximum parameter values. The results are provided in Table 6.4-3 and confirm that the parameter set implemented in the basic model (maximum O.D. and T) is conservative.

As a result, no changes need to be considered in the bounding model.

6.4.2.1.17 Radial Displacement of the Canister The radial displacement of the canister within the cask cavity is investigated via a canister shift towards the cask wall. The result is provided in Table 6.4-3 and shows no influence of the radial canister position on the reactivity.

As a result, the canister centered within the cask cavity, as already implemented in the basic mod-el, is considered in the bounding model.

6.4.2.1.18 Outer Boundary Conditions The outer boundary conditions, such as external moderation and reflection, are evaluated for a single cask as well as for an infinite array of similar densely packed casks. For a single cask, the 6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-9

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @)GNS full reflection is achieved by surrounding the cask with 20 cm water reflector. The results of the investigations with and without external moderation (water or void) are shown in Table 6.4-3. It can be seen that the external moderation has no impact on the reactivity due to the thick cask wall.

As a result, an infinite array of similar, densely packed casks without external moderator, as al-ready implemented in the basic model, is considered in the bounding model.

6.4.2.1.19 Partial Flooding of the Cask The partial flooding of the cask is investigated for the vertical and horizontal orientations of the cask. The results for both orientations are given in Table 6.4-3 and Figure 6.4-3 and indicate that the fully flooded condition is conservative.

As a result, the fully flooded cask cavity with water (p = 1.0 g/cm 3 ), as already implemented in the basic model, is considered in the bounding model.

6.4.2.1.20 Moderator Rod Material The impact of the moderator rod material as well as of the corresponding boreholes on the reactivi-ty is investigated. The results are given in Table 6.4-3 and indicate that neglecting the moderator material and replacing it by the material of the cask body ( cast iron) is conservative.

As a result, no changes need to be considered in the bounding model.

6.4.2.1.21 Shielding Element Structure The reactivity impact of the drain support in one shielding element is investigated. The results are given in Table 6.4-3 and indicate that neglecting the drain support and replacing it by the material of the shielding element (aluminium) is conservative.

As a result, no changes need to be considered in the bounding model.

6.4.2.1.22 Summary Results of Sensitivity Analysis The results of the sensitivity calculations described in the previous subsections are summarized in Table 6.4-3 and presented in Figure 6.4-3.

6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-10

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Table 6.4-3 Summary results of the sensitivity analysis FA no. 6 FA no. 6 k-value for basic model 20dr09cc03 k-value for basic model 20dr09cc03 variation iik, pcm calc. ID variation iik, pcm calc. ID Pellet diameter, L\ in µm Thickness of water channel, L\ in µm

-200 -75 20ln09oF03 -200 66 20BC09Zb03

-150 -30 20un09uS03 -150 64 20ZW09cK03

-100 -41 20WZ09nh03 -100 13 20vx09FJ03

-50 -15 20ai09fw03 -50 13 20TC09LA03 0 0 20Sx09iF03 0 0 20rJ09Jb03 50 12 20Rm09fs03 50 -46 20Yj09vu03 100 -9 20Xv09ra03 100 -51 20RM09rZ03 150 36 20xK09Vg03 150 -71 20ho09Py03 Length of active zone, L\ in mm 200 -86 20yJ09Fb03

-20 8 20Dt09zS03 Inner dimension of FA channel, L\ in µm

-15 10 20WF09aQ03 -200 -28 20HW09KZ03

-10 -2 20cH09Cy03 -150 1 20ee09xg03

-5 -34 20mw09cw03 -100 15 20RN09vw03 0 0 20hr09qH03 -50 -36 20kO09QD03 5 -6 20aQ09do03 0 0 20Sz09vW03 10 3 20MU09hO03 50 -5 20AZ09tW03 15 -21 20yX09zr03 100 -14 20FI09qe03 20 -31 20So09UA03 150 -39 20MI09mK03 Clad inner diameter, L\ in µm 200 -29 20JT09PF03

-150 -452 20YL09eg03 Outer dimension of FA channel, L\ in µm

-100 -339 20Pk09nU03 -200 -22 20DO09FD03

-50 -170 20MR09KC03 -150 -25 20Vz09ap03 0 0 20Gz09tK03 -100 -6 20yW09yN03 50 168 20yto9Ow03 -50 21 20Cf09CN03 100 272 20gT09vt03 0 0 20vJ09AI03 150 474 20hd09Py03 50 37 20Ut09wv03 200 625 20xU09cC03 100 31 20Rz09FI03 Clad outer diameter, L\ in µm 150 -1 20yB09bg03

-200 759 20GE09hl03 200 -5 20EB09cR03

-150 583 20vO09xi03 Axial displacement of FA, L\ in mm

-100 333 20nR09ZP03 -40 -6 20xD09kv03

-50 222 20OQ09kK03 -30 -17 20wM09EO03 0 0 20ES09hD03 -20 12 20jx09yn03 50 -207 20Sn09oe03 -10 -30 20yD09HL03 100 -370 20Ww09qY03 0 0 20oW09RI03 150 -613 20aG09kN03 10 -20 201109TN03 200 -760 20XV09iY03 20 -47 20pX09qY03 30 15 20fW09Bo03 40 9 20aZ09XJ03 6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-11

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Table 6.4-3 Summary results of the sensitivity analysis (continued)

FA no. 6 FA no. 6 k-value for basic model 0.92065 20dr09cc03 k-value for basic model 0.92065 20dr09cc03 variation Ak, pcm calc. ID variation Ak, pcm calc. ID lacement of FA, /J. in mm Thickness of outer sheets, /J. in mm 510 20sO09B103 0 20ut09la03 459 20XO09Xb03 -4 20Rt09xV03 413 20jJ09Yw03 7 20sY09rg03 354 20Jz09Oo03 -2 20bF09LL03 273 20mu09jy03 -27 20mJ09yW03 152 20vz09EF03 5 20lu09oH03 0 20hd09rU03

-135 20el09ST03 -248 20zE09by03

-336 20Ee09Pg03 -47 20jv09Ba03

-519 20Nn09fU03 26 20ih09lf03

-739 20xe09uW03 0 20Jv09pj03

-967 20HE09KN03

-1359 2otq09jA03 to the cask wall -24 2oaL09cqo3 Axial h fuel rods, /J. in mm Outer boundary conditions 0 0 20Ri09vC03 Array, ext. mod. 0 % O 20wm09SY03 5 -7 20Py09oN03 Array, ext. mod. 100 % 4 2owio9TH03 10 -33 20uF09Jx03 Single, ext. mod. 100 % -8 20Mr09LJ03 15 -70 20lo09BY03 Sin le, ext. mod. 0 % -39 20pE09xn03 20 -78 20rZ09Au03 Partial floodin vertical , /J. in cm 25 -85 20ly09pr03 -200 -619 20UM09uz03 30 -145 20GX09LP03 -150 -405 20Pb09Kz03 35 -118 20is09cK03 -100 -193 20zN09RC03 40 -120 20JH09eK03 -90 -164 20Yo09nW03 FA orientation within basket receptacles -80 -80 20Hd09Uc03 Clustering -235 20ek09Fc03 -70 -23 20qo09Jv03 Rotation 19 20SY09XK03 -60 29 20IA09eP03 Material temperature, °C -50 5 20VA09ro03 0 -12 20nN09Et03 -40 -32 20jd09sj03 10 -36 20Jx09CU03 -30 17 20SR09OW03 20 0 20IJ09Xw03 -20 31 20ab09at03 30 -124 20xv09pr03 -10 O 20cd09aw03 40 -253 20oR09DF03 0 0 20rt09gh03 70 -483 20MH09BO03 Partial flooding (horizontal), number of dry FA rows 120 -1 034 20Gr09xg03 0 0 20kO09BF03 170 -1490 20Fq09vG03 1 -230 20jx09MZ03 Dimension of basket rece tacles, /J. in mm 2 -474 20OH09eX03 179 20SD09Wf03 3 -918 20Fy09Vy03 0 20oK09za03 4 -1523 20yt09LW03

-214 20rn09KB03 5 -2463 20sR09TV03

-368 20WN09UL03 Moderator rod material

-563 20ZY09dl03 Neglected 0 20JQ09jh03 Thickness asket sheets, /J. in mm PE -189 20Fc09va03 0 20EZ09ED03 Void 12 20li09CF03

-136 20Rm09NM03 Water -236 20hl09FS03

-224 20ar09Dj03 Shielding element structure

-315 20Xv09jv03 Drain su art -61 20JI09UQ03

-488 20Gi09Ao03

-551 20FM09DC03 6.4 Criticality Evaluation Section 6.4, Rev. O Page 6.4-12

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Figure 6.4-3 Graphical representation of the summary results 6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-13

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 1500 100 1000 50 E E g_ ~

(J 500 C2.

0

"'....S 0

"'....S -50

+I

~ -500

<I -

<I

-100

-1000 -150

-1500 -200

-10 -5 0 5 10 0 10 20 30 40 50 Axial position of part-length fuel rods, Radial displacement of FA, 6. in mm

6. in mm E

(J

a. -500

"'....S 500 o .....

~

~ .. ~

E(J C2.

"'....S 400 200 0

-200

.!!,-1000

.:.: -400

<1-1500 -- <I

-000

-2000 -800 0 50 100 150 200 -2 -1 0 2 3 4 Dimension of basket receptacles, Material temperature, °C

6. inmm 200 200 I

E o E 100 l

(J (J

a. -200 C2.

1~

"'S "'....S 0

_ -400 .!!,

.:.: i~ .:.:

<I ~~ <I -100

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-800 -200 0 0.2 0.4 0.6 0.8 1.2 0 2 3 4 5 6 Thickness of basket sheets, A in mm Thickness of outer sheets, A in mm 200

.iii..

0 E E " o------+---+----+--+---<

(J (J

.iii.

a. -200 C2.

.;:!. -400

....;-1000 + - - - + - - - - , 1 - - . . , . . - - - - + - - + - - - - - ;

<I <I -2000 + - - - + - - - - , 1 - - - + - - - - + - - + - - - - - ;

-000

-800 -3JOO - ' - - - ~ - - L _ _ _ . . . J . . __ ___.__ __.___ __.

-250 -200 -150 -100 -50 0 0 1 2 3 4 5 6 Partial flooding (vertical), A in cm Partial flooding (horizontal), number of dry FA rows Figure 6.4-3 Graphical representation of the summary results (continued) 6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-14

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS 6.4.2.2 Results for the Bounding Model for NCS As discussed in subsection 6.4.2.1, the following conservative assumptions for the basic model are confirmed:

- fully flooded gap between fuel pellet and clad,

- nominal pellet diameter,

- nominal length of the active zone,

- minimum thickness of the FA internal structures,

- nominal dimensions of the FA channel,

- nominal axial position of the FA,

- - axial position of part-length fuel rods beginning at the lower edge of the active zone,

- same orientation of all FA within the basket receptacles,

- nominal temperature,

- nominal thickness of the basket sheets,

- nominal thickness of the outer sheets,

- maximum canister outer diameter with maximum wall thickness,

- no radial displacement of the canister within the cask,

- void outside the cask (external moderation 0 %),

- infinite array of densely packed casks,

- fully flooded cask (internal moderation 100 %),

- neglected moderator rods and sheets,

- neglected drain support in the shielding element.

Based on the sensitivity analysis in subsection 6.4.2.1, the following changes are applied to the basic model to develop the bounding model for NCS:

bounding material densities and compositions,

- maximum clad inner diameter,

- minimum clad outer diameter,

- minimum inner dimension of the basket receptacles,

- radial displacement of all FA towards the center of the fuel basket.

6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-15

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 The criticality calculations using the bounding model for NCS are performed for all analyzed FA.

The results are summarized in Table 6.4-4 and include the maximum effective neutron multipli-cation factors k and the calculational bias with its uncertainty ~ku. All contributions to the final re-sults are determined with a 95 % probability at a 95 % confidence level, as discussed in subsec-tion 6.5.2.1 and Appendix 6-1.

A sample input file for the bounding model for NCS with the FA of type ATRIUM-1 OA is provided in Appendix 6-2.

Table 6.4-4 Results for the bounding model for NCS Infinite array of flooded Case storage casks (bounding model for NCS)

Internal 100 % (Water)

FA no. moderation External moderation 0 % (Void) and reflection Fuel type k+Aku calc. ID 1 GE 8x8-1 0.84386 20IZ09ny02 2 GE 8x8-2 0.85061 20da09vS02 3 SPC 8x8-2 0.84625 20rq09Tj02 4 GE9B 8x8 0.88137 201109Vo02 5 GE12 LUA 0.90514 20Xb09ei02 6 ATRIUM-10A 0.93729 20nS09hZ02 6.4.2.3 Proof of Conservativity of the Bounding Model for NCS Additional evaluations as a proof of the conservativity of the bounding model for NCS is based on this model and includes:

- partial cask loading,

- planar fuel enrichment,

- reflection conditions for storage and transfer casks.

The results of the analysis are presented in subsections 6.4.2.3.1 through 6.4.2.3.3.

6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-16

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 G S 6.4.2.3.1 Partial Cask Loading The impact of different empty basket positions on the reactivity is evaluated. The investigated sin-gle empty basket positions (No. 6, 9, 32, 49, 51, 63, 64 and 69) are radially distributed within the fuel basket, as shown in Figure 6.4-4. The results of these different calculations are given in Table 6.4-5 and Figure 6.4-5 and confirm that the full cask loading is conservative.

Figure 6.4-4 Evaluated empty basket positions Table 6.4-5 Evaluation results for the partial cask loading FA no. 6 k-value for NCT model 20Xg09mz04 variation Ak, pcm calc. ID Number of empty basket position 6 -93 201n09oF03 49 -409 20un09uS03 63 -1002 20WZ09nh03 69 -1463 20ai09fw03 64 -1197 20Sx09iF03 51 -741 20Rm09fs03 32 -415 20Xv09ra03 9 -168 20xK09Vg03 6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-17

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS 500 E

u

c. -500 o

-s.....

-!!- -1000 ....

~

.II:

<I -1500 '

-2000 0 20 40 60 80 Number of empty basket position Figure 6.4-5 Graphical representation of results for the partial cask loading 6.4.2.3.2 Planar Fuel Enrichment 6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-18

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 Figure 6.4-6 Cross section of the fuel lattice and the group numbers (Group no.) with dif-ferent enrichments in the generic model Table 6.4-6 Investigated enrichment distributions The results of criticality calculations given in Table 6.4-7 and Figure 6.4-7 and confirm that the use of the planar-averaged uniform enrichment is conservative.

6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-19

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Table 6.4-7 Evaluation results for the planar fuel enrichment Figure 6.4-7 Graphical representation of results for the planar fuel enrichment 6.4.2.3.3 Reflection Conditions The impact of different reflection conditions is separately evaluated for the storage and for the transfer casks. For the storage cask, the analysis is performed using the bounding model for NCS, as described in subsections 6.3.1.2 and 6.3.2. For the transfer cask, the structures of the storage cask in the bounding model for NCS are replaced by the structures of the transfer cask, as de-scribed in subsection 6.3.1.3. Single cask configurations with the following reflector materials are investigated:

- void (vacuum),

- 20 cm water shell (1 g/cm 3 ),

- 50 cm concrete shell (2.3 g/cm 3 , referred to as "reg-concrete" in the Standard Composition Library chapter of the SCALE manual),

- 50 cm steel shell (8.03 g/cm 3 , referred to as "SS316" in the Standard Composition Library chapter of the SCALE manual).

6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-20

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS The results of these calculations are given in Table 6.4-8 and show no significant reactivity in-crease compared to the bounding model for NCS. In a conservative way, the following reflection conditions are selected to be bounding during handling, packaging, transfer and storage:

• 20 cm water shell for the storage cask,

- 50 cm steel shell for the transfer cask.

Table 6.4-8 Evaluation results for neutron reflection conditions FA no. 6 FA no. 6 k-value for NCS model 20Xg09mz04 k-value for NCS model 20Xg09mz04 variation Ak, pcm calc. ID variation Ak, pcm calc. ID Reflection conditions for storage cask Reflection conditions for transfer cask I-Void -12 21oM02FU08 Void -65 21Qh02dP08 20 cm water 8 21VH02Jc08 20 cm water -75 21LU02hX08 50 cm concrete -24 21CD02tz08 50 cm concrete -47 21WD02Cu08 50 cm steel -8 21uF02Le08 50 cm steel 22 21ST02JB08 6.4.3 Criticality Results 6.4.3.1 Evaluation of Cask Arrays under Normal Conditions of Storage 6.4.3.1.1 Configuration The evaluation model incorporates the following changes compared to the bounding model for NCS, described in section 6.3:

• infinite array of dry storage casks: array of dry densely packed storage casks with void be-tween the casks.

According to the sensitivity analysis in subsection 6.4.2.1, the external moderation has no impact on the reactivity of the infinite array of fully flooded casks and, taking into account the low reactivity level of internally dry casks, does not need to be analyzed.

6.4.3.1.2 Results The criticality calculations are performed for all analyzed FA. The results are summarized in Table 6.4-9 and include the maximum effective neutron multiplication factors k and the calculation-al bias with its uncertainty b.ku. All contributions to the final results are determined with a 95 %

probability at a 95 % confidence level, as discussed in subsection 6.5.2.1 and Appendix 6-1.

6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-21

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS The results for the infinite array of dry storage casks show that the maximum effective neutron mul-tiplication factors including the calculational bias with its uncertainty fulfil the acceptance criteria k + Aku < 0.95 for all investigated cask loadings and demonstrate the compliance with the require-ments of § 72.124 and § 72.236. Moreover, the comparison with the results of criticality calcula-tions using the bounding model for NCS provided in subsection 6.4.2.2 demonstrates that the reac-tivity of an infinite array of dry storage casks is bounded by the analysis of fully flooded casks.

Table 6.4-9 Results for infinite arrays of dry storage casks Infinite array of dry Case storage casks Internal 0 % (Void)

FA no. moderation External moderation 0 % (Void) and reflection Fuel type k +Aku calc. ID 1 GE 8x8-1 0.35051 20EF09bR02 2 GE 8x8-2 0.34374 20hN09sh02 3 SPC 8x8-2 0.33861 20DB09dw02 4 GE9B 8x8 0.36399 20JJ09zw02 5 GE12 LUA 0.39113 20La09rj02 6 ATRIUM-10A 0.42457 20rk09gn02 6.4.3.2 Evaluation of Single Storage Cask 6.4.3.2.1 Configuration The evaluation model incorporates the following changes compared to the bounding model for NCS, described in section 6.3:

- single, fully reflected storage cask: single, fully flooded storage cask surrounded with 20 cm water reflector.

As the water leakage into the canister cavity as well as any deformations of basket structures and content important to criticality safety under normal, off-normal and accident conditions of storage 6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-22

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS are excluded, the configurations for a single storage cask under normal, off-normal and accident conditions of storage are bounded by the corresponding single, fully flooded and fully reflected storage cask, as described above.

6.4.3.2.2 Results The criticality calculations are performed for all analyzed FA. The results are summarized in Table 6.4-10 and include the maximum effective neutron multiplication factors k and the calcula-tional bias with its uncertainty Llku. All contributions to the final results are determined with a 95 %

probability at a 95 % confidence level, as discussed in subsection 6.5.2.1 and Appendix 6-1.

The results show that the maximum effective neutron multiplication factors including the calcula-tional bias with its uncertainty fulfil the acceptance criteria k + ~ku < 0.95 for all investigated cask 9 loadings and demonstrate the compliance with the requirements of§ 72._124 and§ 72.236.

Table 6.4-10 Results for a single, fully reflected storage cask Single, fully reflected Case storage cask Internal 100 % (Water)

FA no. moderation External moderation 100 % (Water) and reflection Fuel type k+Aku calc. ID 1 GE 8x8-1 0.84387 20hr09qR02 2 GE 8x8-2 0.85034 20bM09JV02 3 SPC 8x8-2 0.84629 20AV09iY02 4 GE9B 8x8 0.88145 20ed09nP02 5 GE12 LUA 0.90527 20Np09QB02 6 ATRIUM-10A 0.93725 20Lf09eS02 6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-23

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS 6.4.3.3 Evaluation of Single Transfer Cask 6.4.3.3.1 Configuration The evaluation model incorporates the following changes compared to the bounding model for NCS, described in section 6.3:

- single, fully reflected transfer cask: single, fully flooded transfer cask surrounded with 50 cm steel reflector.

As any deformations of basket structures and content important to criticality safety under normal, off-normal and accident conditions of storage are excluded, the configurations for a single transfer cask under normal, off-normal and accident conditions during handling, packaging and transfer operations are bounded by the corresponding single, fully flooded and fully reflected transfer cask,

.* as described above.

6.4.3.3.2 Results The criticality calculations are performed for all analyzed FA. The results are summarized in Table 6.4-11 and include the maximum effective neutron multiplication factors k and the calcula-tional bias with its uncertainty ~ku. All contributions to the final results are determined with a 95 %

probability at a 95 % confidence level, as discussed in subsection 6.5.2.1 and Appendix 6-1.

The results show that the maximum effective neutron multiplication factors including the calcula-tional bias with its uncertainty fulfil the acceptance criteria k + Aku < 0.95 for all investigated cask loadings and demonstrate the compliance with the requirements of § 72.124 and § 72.236.

6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-24

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Table 6.4-11 Results for a single, fully reflected transfer cask Single, fully reflected Case transfer cask Internal 100.% (Water)

FA no. moderation External moderation 100 % (Steel) and reflection Fuel type k+Aku calc. ID 1 GE 8x8-1 0.84410 21Vu02YZ08 2 GE 8x8-2 0.85044 21lv02wr08 3 SPC 8x8-2 0.84641 21HQ02yr08 4 GE9B 8x8 0.88149 21oa02SS08 5 GE12 LUA 0.90541 21NW02dx08 6 ATRIUM-10A 0.93740 21jq02Sv08 List of References

[1] B. T. Rearden and M.A. Jessee, Eds.

SCALE Code System, ORNL/TM-2005/39, Version 6.2.2 Oak Ridge National Laboratory, Oak Ridge, Tennessee (2017)

Available from Radiation Safety Information Computational Center as CCC-834 6.4 Criticality Evaluation Section 6.4, Rev. 0 Page 6.4-25

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev.O I(?;\ GNS

\f::Y 6.5 Critical Benchmark Experiments Name, Function Date Signature Prepared Reviewed 6.5 Critical Benchmark Experiments Section 6.5, Rev. O Page 6.5-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 6.5.1 Benchmark Experiments and Applicability 6.5 Critical Benchmark Experiments Section 6.5, Rev. 0 Page 6.5-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 Figure 6.5-1 Frequency distribution of correlation coefficients between the application and selected benchmarks 6.5.2 Results of the Benchmark Calculations 6.5.2.1 Bias Determination (5.1)

(5.2) 6.5 Critical Benchmark Experiments Section 6.5, Rev. 0 Page 6.5-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS (5.3)

(5.4)

(5.5)

(5.6) 6.5 Critical Benchmark Experiments Section 6.5, Rev. 0 Page 6.5-4

Non-Proprietary Version 10,14-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS (5.7)

(5.8)

(5.9)

(5.10)

(5.11)

(5.12) 6.5.2.2 Validation Results 6.5 Critical Benchmark Experiments Section 6.5, Rev. 0 Page 6.5-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2_.390 Rev. 0 @GNS 6.5 Critical Benchmark Experiments Section 6.5, Rev. 0 Page 6.5-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Table 6.5-1 Results of benchmark calculations 6.5 Critical Benchmark Experiments Section 6.5, Rev. 0 Page 6.5-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Table 6.5-1 Results of benchmark calculations (continued)

List of References

[1] International Handbook of Evaluated Criticality Safety Benchmark Experiments NEA Nuclear Science Committee September 2019 Edition, NEA/NSC/DOC(95)03 6.5 Critical Benchmark Experiments Section 6.5, Rev. 0 Page 6.5-8

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev. O @GNS 6.6 Appendix Prepared Reviewed 6.6 Appendix Section 6.6, Rev. 0 Page 6.6-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Appendix 6-1: Methodological uncertainty Appendix 6-2: Sample Input File for Bounding Model for NCS (Cale. ID: 20nS09hZ02) 6.6 Appendix Section 6.6, Rev. 0 Page 6.6-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 7 Containment 7.0 Overview Prepared Reviewed 7.0 Overview Section 7.0, Rev. 1 Page 7.0-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS In this chapter the compliance of the CASTOR geo69 storage cask as part of the Dry Storage System (DSS) regarding the containment requirements from 10 CFR 72 is shown and the containment boundary is described. For further information on the design, see also Chapter 1.2 and the design drawings and design parts lists in Section 1.5, appendices 1-1, 1-4 and 1-5. All item numbering in this Chapter correspond to those design parts.

The storage cask is loaded with the content described in Appendix 7-1 and based on Subsection 1.2.3.

7.0 Overview Section 7.0, Rev. 1 Page 7.0-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @)GNS 7.1 Containment Boundary Prepared Reviewed 7.1 Containment Boundary Section 7.1, Rev. 1 Page 7.1-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The CASTOR geo69 DSS is designed for FA with moderate burn-up as well as high burn-up fuel with an averaged burn-up above 45 GWd/MgHM (cf. Section 1.2.3). A double (inner and outer) containment as described in the following subsections in more detail is used.

7.1 Containment Boundary Section 7.1, Rev. 1 Page 7.1-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 G 7 .1.1 Containment Vessel The CASTOR geo69 DSS containment vessel is constituted by the following subassemblies (item numbers according to the design drawing and design parts lists listed in Section 1.5, appendices 1-1, 1-4 and 1-5):

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

- I tightening plug (Item 10) 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 (Items 62, 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 , as well as pressure switch in the cask lid, cap screws (Item 37) and metal gasket (Item 71 (Ag) with a torus diameter of The outer jackets of all metal gaskets of the containment system are made of silver (Ag). The el canister body (Item 2) is designed by welding Items 2-2 to 2-5 together.

The inner and outer containment boundaries are shown in Figure 7.1-1.

7.1 Containment Boundary Section 7.1, Rev. 1 Page?.1-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 (@)GNS Figure 7.1-1: lnn*er and outer containment (canister and cask) 7.1.2 Containment Penetrations As described in Section 7.1.1, there is one service orifice through the lid of the inner containment and two service orifices through the lid of the outer containment.

7.1 Containment Boundary Section 7.1, Rev.1 Page 7.1-4

Non-Proprietary Version 1014-SR-00002 Proprietary lnf(?rmation withheld per 10CFR 2.390 Rev. 1 @GNS The service orifice through the inner containment is shown in more detail regarding the leakage path in Figure 7.1-2. The service orifices through the outer containment are shown in more detail regarding the leakage paths in Figure 7.1-3 and Figure 7.1-4.

Figure 7.1-2: Containment boundary detail at the canister lid with tightening plug Figure 7.1-3: Containment boundary detail at the cask lid with protection cap 7.1 Containment Boundary Section 7.1, Rev. 1 Page7.1-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 7.1-4: Containment boundary detail at the cask lid with pressure switch 7 .1.3 Seals and Welds The monolithic cask body , the canister 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 systems.

The potential leakage paths of the inner and outer containment are shown in Figure 7.1-5.

Envi ronment I Metal gasket (Item 44, Ag )

in the protection cap (Item 11 3)

Metal gasket (Item 69, Ag )

in the cask lid (Item 55)

Metal gasket (Item 71 , Ag )

in the pressure switch Metal gasket (Item 16, Ag ) Metal gasket (Item 13, Ag) in the canister lid (Item 3) in the tightening plug (Item 10)

Welded cladding s Fuel matrix Figure 7.1-5: Leakage paths of the inner and outer containment 7.1 Containment Boundary Section 7.1, Rev . 1 Page 7.1-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 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 (see Section 8.5, Appendix 8-4).

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 each containment, a maximum standard helium leakage rate of (leak test criterion) is proven by measurement after both, assembly prior to commissioning (acceptance test, see Section 10.1) and loading at a nuclear facility (e.g. nuclear power plant (NPP), see Chapter 9).

This corresponds to a reference air leakage rate of 7.1.4 Closure The SNF is double contained by the canister and by the cask. As described above, each containment and its service orifices are equipped with a closure. consisting of a lid and a metal gasket. The canister lid is attached to the canister body with thread bolts (Item 6). The cask lid is attached to the cask body by hexagonal screws (Items 62, 63).

This double containment is designed to maintain containment integrity during normal conditions of storage (NCS) and off-normal conditions of storage as well as for accident conditions of storage (ACS) and natural phenomena events.

7.1 Containment Boundary Section 7.1, Rev. 1 Page 7.1-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 7.2 Containment Requirements for Normal Conditions of Storage Prepared Reviewed 7.2 Containment Requirements for Normal Conditions of Storage Section 7.2, Rev. 1 Page 7.2-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 7.2.1 Release of Radioactive Material As the structural integrity and the redundant containments are not impaired under NCS (cf.

Chapter 3), the design leakage rate of the considered containment is not greater than 10-7 ref*cm 3/s (leak-tight according to ANSI N14.5 [1]). Therefore, no dedicated activity release calculations and corresponding dose calculations as described in [2] are required.

Nevertheless, the activity mobilization inside the canister is given as described in Appendix 7-3.

Starting with the activity content from Table 7.6-1, using the release fractions fs, fo, fv, fr and fc for normal conditions from Table 7.6-3 and using the value for the free gas volume V inside the canister from Table 7.2-4, the activity mobilization inside the cask is shown in Table 7.2-1 for gases and volatiles, in Table 7.2-2 for fines and crud and in Table 7.2-3 summed up for the nuclide mixture for NCS. For nuclides that may be available as gases or volatiles, an additional contribution as fines is taken into account.

Table 7 .2-1: Mobilized activity and activity concentration for gases and volatiles for NCS 7.2 Containment Requirements for Normal Conditions of Storage Section 7.2, Rev. 1 Page 7.2-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 7.2-2: Mobilized activity and activity concentration for fines and crud for NCS 7.2 Containment Requirements for Normal Conditions of Storage Section 7.2, Rev. 1 Page 7.2-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Table 7 .2-3: Mobilized activity and activity concentration summed up for the nuclide mixture for NCS 7.2.2 Pressurization of Containment Vessel The canister inside the cask contains SNF during NCS. The interior space inside the canister is drained, dried, evacuated and backfilled with helium gas prior to final closure. The interior space inside the cask with a loaded canister is dried, if applicable, evacuated and backfilled with helium gas prior to final closure. 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 absolute pressure of PHe,o (cf. Appendix 7-2) inside the canister and PHe,cask,o (cf. Appendix 7-2) inside the cask.

With the procedure described in Appendix 7-2, the maximum internal pressures for NCS are calculated as follows. Structural integrity and containment of the canister and the storage cask are not impaired in NCS (cf. Chapter 3) and both containment barriers remain leak-tight. Therefore, the pressures are calculated for the canister and the cask separately. There are no combustible gases inside the containment.

The maximum absolute pressure Pu inside the canister assuming fuel rod failure is obtained by Pu= PHe,o

• T 9as / T He,o (cf. Appendix 7-2) with the boundary conditions given in Table 7.2-4.

The maximum absolute pressure Pu inside the cask is obtained by Pu= PHe,cask,O

• Tgas / T He,cask,O (cf. Appendix 7-2) with Tgas (covering NCS value for the filling gas of the cask in Section 4.4).

7.2 Containment Requirements for Normal Conditions of Storage Section 7.2, Rev. 1 Page 7.2-4

Non-Proprietary Version 1O14-SR-OOOO2 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The maximum normal operating pressure (MNOP) is the value of the upstream absolute pressure Pu for NCS, reduced by the atmospheric pressure at mean sea level, i. e. 101.3 kPa. With a maximum absolute pressure Pu Table 7 .2-4: Boundary conditions for canister pressure calculation (NCS) assuming fuel rod failure Parameter Symbol Value Reference Initial canister filling gas pressure PHe,0 Appendix 7-2 Initial canister filling gas temperature THe,O Appendix 7-2 Gas temperature (covering value from canister el gas mixture)

Fuel rod failure fraction Tgas fa 1%

Section 4.4 Table 7.6-3 Fission gas release fraction fo 15 % Table 7.6-3 Maximum produced amount of fission gas per GFG Appendix 7-1 loading Amount of mobilized fission gas in the canister nFG Cale.

(GFG . fs . fo)

Maximum amount of fuel rod filling gas helium GFR Appendix 7-1 per loading Amount of mobilized fuel rod filling gas helium nFR Cale.

(GFR . fa)

Sum of gas released from the content into the n Cale.

canister (nFG + nFR)

Minimum free gas volume inside the canister V Chapter 1 List of References

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

[2] NUREG-2215, April 2020 Standard Review Plan for Spent Fuel Dry Storage Systems and Facilities 7.2 Containment Requirements for Normal Conditions of Storage Section 7.2, Rev. 1 Page 7.2-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 (@)GNS 7 .3 Containment Requirements for Off-Normal Conditions of Storage Prepared Reviewed 7.3 Containment Requirements for Off-Normal Conditions of Storage Section 7.3, Rev. 1 Page 7.3-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 7.3.1 Release of Radioactive Material As the structural integrity and the redundant containments are not impaired under off-normal conditions (cf. Chapter 3), the design leakage rate of the considered containment is not greater than 10-7 ref*cm 3/s (leak-tight according to [1]). Therefore, no dedicated activity release calculations and corresponding dose calculations as described in [2] are required.

Nevertheless, the activity mobilization inside the canister is given as described in Appendix 7-3.

Starting with the activity content from Table 7.6-1, using the release fractions fs, fo, fv, fr and fc for off-normal conditions from Table 7.6-3 and using the value for the free gas volume V inside the canister from Table 7.3-4, the activity mobilization inside the cask is shown in Table 7.3-1 for gases and volatiles, in Table 7.3-2 for fines and crud and in Table 7.3-3 summed up for the nuclide mixture for off-normal conditions. For nuclides that may be available as gases or volatiles, an additional contribution as fines is taken into account.

Table 7 .3-1: Mobilized activity and activity concentration for gases and volatiles for off-normal conditions 7.3 Containment Requirements for Off-Normal Conditions of Storage Section 7.3, Rev. 1 Page 7.3-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 7.3-2: Mobilized activity and activity concentration for fines and crud for off-normal conditions 7.3 Containment Requirements for Off-Normal Conditions of Storage Section 7.3, Rev. 1 Page 7.3-3

Non-Proprietary Version 1014-SR-00002 Rev. 1 Proprietary Information withheld per 10CFR 2.390 s

Table 7.3-3: Mobilized activity and activity concentration summed up for the nuclide mixture for off-normal conditions

• The mobilized content and the mobilized activity concentration in the canister increases for those mobility types where the release fraction fs is used (gases, volatiles and fines) compared to NCS by the factor of 10.

7.3.2 Pressurization of Containment Vessel The canister inside the cask contains SNF during off-normal conditions of storage. The interior space inside the canister is drained, dried, evacuated and backfilled with helium gas prior to final closure. The interior space inside the cask with a loaded canister is dried, if applicable, evacuated and backfilled with helium gas prior to final closure. 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 absolute pressure of PHe,o (cf. Appendix 7-2) inside the canister and PHe,cask,o (cf.

Appendix 7-2) inside the cask.

With the procedure described in Appendix 7-2, the maximum internal pressures for off-normal condition are calculated as follows. Structural integrity and containment of the canister and the storage cask are not impaired for off-normal conditions (cf. Chapter 3) and both containment barriers remain leak-tight. Therefore, the pressures are calculated for the canister and the cask separately. There are no combustible gases inside the containment.

The maximum absolute pressure Pu inside the canister assuming fuel rod failure results with the boundary conditions given in Table 7.3-4.

7.3 Containment Requirements for Off-Normal Conditions of Storage Section 7.3, Rev. 1 Page 7.3-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 (@)GNS The maximum absolute pressure Pu inside the cask, obtained by Pu = PHe,cask0

• T gas I T He,cask,0 ( cf. Appendix 7-2) with T gas (covering value for the filling gas of the cask in Section 4.5).

Table 7.3-4: Boundary conditions for canister pressure calculation (off-normal conditions) assuming fuel rod failure Parameter Symbol Value Reference Initial canister filling gas pressure PHe,0 Appendix 7-2 Initial canister filling gas temperature Appendix 7-2 THe,O Gas temperature (covering value from canister Tgas Section 4.5 I gas mixture)

- Fuel rod failure fraction Fission gas release fraction Maximum produced amount of fission gas per loading Amount of mobilized fission gas in the canister fs fo GFG 10%

15%

Table 7.6-3 Table 7.6-3 Appendix 7-1 nFG Cale.

(GFG

• fs

• fo)

Maximum amount of fuel rod filling gas helium GFR Appendix 7-1 per loading Amount of mobilized fuel rod filling gas helium nFR Cale.

(GFR . fs)

Sum of gas released from the content into the n Cale.

canister (nFG + nFR)

Minimum free gas volume inside the canister V Chapter 1 List of References

[1] ANSI N14.5-2O14, American National Standard For Radioactive Materials - Leakage Tests on Packages for Shipment

[2] NUREG-2215, April 2020 Standard Review Plan for Spent Fuel Dry Storage Systems and Facilities 7.3 Containment Requirements for Off-Normal Conditions of Storage Section 7.3, Rev. 1 Page 7.3-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 7.4 Containment Requirements for Accident Conditions of Storage Prepared Reviewed 7.4 Containment Requirements for Accident Conditions of Storage Section 7.4, Rev. 1 Page 7.4-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 7 .4.1 Fission Gas Products For the cask content compliant with the stipulations in Section 1.2.3, the maximum produced amount of fission gas of GFG is obtained (cf. Appendix 7-1). This amount corresponds to a cask loading of 69 spent uranium oxide (UOX) FA with maximum values for the final discharge burn-up of **** and the heavy metal mass of * * *

• each. Therefore, all possible cask loadings are covered by the considered amount of fission gas.

For each fuel rod that is assumed to have failed, the fraction fo (cf. Table 7.6-3) of the produced fission gas is assumed to be released into the cavity of the canister. For accident conditions of storage all fuel rods are assumed to have failed (fs = 1.0, cf. Table 7.6-3).

According to [1], there are two separate cases to be analyzed for accid.ent conditions of storage:

accident fire conditions (ACS-fire) and accident impact conditions (ACS-impact).

According to [1], the fraction of fission gas release fo is used as 0.15 (15 %) for ACS-fire as well as 0.35 (35 %) for ACS-impact which includes an extra 20 % fraction of the pellet-retained fission gases that might be released during a drop impact. This results in a maximum mobilized amount of fission gas inside the canister of nFG = GFG

• fs

• f G, which

• for ACS-fire and -

-for ACS-impact.

The fission gas Xenon provides about-% of this amount (cf. Table 7.6-2). The distribution of all considered fission gases is listed as values in parentheses in Table 7.6-2.

7 .4.2 Release of Content As the structural integrity and the redundant containments are not impaired under ACS-fire and ACS-impact (cf. Chapter 3), the design leakage rate of the considered containment is not greater than 10-7 ref*cm 3/s (leak-tight according to [2]). Therefore, no d.edicated activity release calculations and corresponding dose calculations as described in [3] are req~ired.

Nevertheless, the activity mobilization inside the canister is given as described in Appendix 7-3.

Starting with the activity content from Table 7.6-1, using the release fractions fs, fo, fv, fr and fc for ACS-fire resp. ACS-impact from Table 7.6-3 and using the value for the free gas volume V inside the canister from Table 7.4-4 resp. Table 7.4-5, the activity mobilization inside the cask is shown in Table 7.4-1 for gases and volatiles, in Table 7.4-2 for fines and ,crud and in Table 7.4-3 summed up for the nuclide mixture for ACS-fire and ACS-impact. For nuclides that may be available as gases or volatiles, an additional contribution as fines is taken into account.

7.4 Containment Requirements for Accident Conditions of Storage Section 7.4, Rev. 1 Page 7.4-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @)GNS Table 7.4-1: Mobilized activity and activity concentration for gases and volatiles for ACS-fire and ACS-impact 7.4 Containment Requirements for Accident Conditions of Storage Section 7.4, Rev. 1 Page 7.4-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 7.4-2: Mobilized activity and activity concentration for fines and crud for ACS-fire and ACS-impact 7.4 Containment Requirements for Accident Conditions of Storage Section 7.4, Rev. 1 Page 7.4-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 s Table 7.4-3: Mobilized activity and activity concentration summed up for the nuclide mixture for ACS-fire and ACS-impact

- I The mobilized content and the mobilized activity concentration in the canister is about the same for ACS-fire and ACS-impact but differs in the contribution of gases and fines. This corresponds to the different release fractions listed in Table 7.6-3.

7.4.3 Pressurization of Containment Vessel The canister inside the cask contains SNF during ACS-fire and ACS-impact. The interior space inside the canister is drained, dried, evacuated and backfilled with helium gas prior to final closure.

The interior space inside the cask with a loaded canister is dried, if applicable, evacuated and backfilled with helium gas prior to final closure. 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 absolute pressure of PHe,o (cf. Appendix 7-2) inside the canister and PHe,cask,o (cf.

- Appendix 7-2) inside the cask.

With the procedure described in Appendix 7-2, the maximum internal pressures for ACS-fire and ACS-impact are calculated as follows. Structural integrity and containment of the canister and the storage cask are not impaired for ACS-fire and ACS-impact (cf. Chapter 3) and both containment barriers remain leak-tight. Therefore, the pressures are calculated for the canister and the cask separately. There are no combustible gases inside the containment.

Accident-Fire Conditions:

For ACS-fire the maximum absolute pressure Pu inside the canister assuming fuel rod failure is obtained by Pu= PHe,o

• Tgas / T He,o (cf. Appendix 7-2) with the boundary conditions given in Table 7.4-4.

7.4 Containment Requirements for Accident Conditions of Storage Section 7.4, Rev. 1 Page 7.4-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The maximum absolute pressure Pu inside the cask is obtained by Pu = PHe,cask,O

• T gas / T He,cask,O ( cf. Appendix 7-2) with T gas (covering ACS-fire value for the filling gas of the cask in Section 4.6).

Table 7.4-4: Boundary conditions for canister pressure calculation (ACS-fire) assuming fuel rod failure Parameter Symbol Value Reference Initial canister filling gas pressure PHe,O Appendix 7-2 Appendix 7-2 Initial canister filling gas temperature THe,O Gas temperature (covering value from canister Tgas Section 4.6 I gas mixture)

- Fuel rod failure fraction Fission gas release fraction Maximum produced amount of fission gas per loading Amount of mobilized fission gas in the canister fs fo GFG 100 %

15 %

Table 7.6-3 Table 7.6-3 Appendix 7-1 nFG Cale.

(GFG . fs . fo)

Maximum amount of fuel rod filling gas helium GFR Appendix 7-1 per loading Amount of mobilized fuel rod filling gas helium nFR Cale.

(GFR . fs)

Sum of gas released from the content into the n Cale.

canister (nFG + nFR)

Minimum free gas volume inside the canister V Chapter 1 Accident-lm~act Conditions:

Section 4.6 provides two scenarios for ACS-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:

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

• ACS-impact II: with release and reconfiguration of spent fuel For ACS-impact the maximum absolute pressures Pu (ACS-impact I) and Pu (ACS-impact II) inside the canister result with the boundary conditions given in Table 7.4-5.

The maximum absolute pressures Pu (ACS-impact I) and Pu (ACS-impact II) inside the cask are obtained by Pu = PHe,cask,o

• T gas / T He,cask,o (cf. Appendix 7-2) with T gas and Tgas respectively (covering ACS-impact values for the filling gas of the cask in Section 4.6).

7.4 Containment Requirements for Accident Conditions of Storage Section 7.4, Rev. 1 Page 7.4-6

Non-Proprietary Version 1O14-SR-OOOO2 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Table 7.4-5: Boundary conditions for canister pressure calculation (ACS-impact)

Value Value Parameter Symbol Reference (ACS-impact I) (ACS-impact II)

Initial canister filling gas pressure PHe,0 Appendix 7-2 Initial canister filling gas temperature THe,O Appendix 7-2 Gas temperature (covering value from Tgas Section 4.6 canister gas mixture)

Fuel rod failure fraction fa 100 % Table 7.6-3 Fission gas release fraction fo 35% Table 7.6-3 Maximum produced amount of fission GFG Appendix 7-1 gas per loading Amount of mobilized fission gas in the nFG Cale.

canister (GFG

• fa

• fo)

Maximum amount of fuel rod filling gas GFR Appendix 7-1 helium per loading Amount of mobilized fuel rod filling gas nFR Cale.

helium (GFR

• fa}

Sum of gas released from the content n Cale.

into the canister (nFG + nFR)

Minimum free gas volume inside the V Chapter 1 canister List of References

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

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

[3] NUREG-2215, April 2020 Standard Review Plan for Spent Fuel Dry Storage Systems and Facilities 7.4 Containment Requirements for Accident Conditions of Storage Section 7.4, Rev. 1 Page 7.4-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information wi_thheld per 10CFR 2.390 Rev. 1 @GNS 7.5 Containment Requirements for Short-Term Operations Prepared Reviewed 7.5 Containment Requirements for Short-Term Operations Section 7.5, Rev. 1 Page 7.5-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 1 OCFR 2.390 Rev. 1 @GNS 7 .5.1 Pressurization of Containment Vessel The canister inside the transfer cask ~ontains SNF during handling of the transfer cask inside the reactor building. The interior space inside the canister is drained, dried, evacuated and backfilled with helium gas prior to final closure of the canister. Therefore, no vapors or gases are present which could cause a reaction or explosion inside the canister.

The maximum absolute pressure Pu inside the canister during short-term operations is obtained by Pu= PHe,o

• T9as / T He,o (cf. Appendix 7-2) with T 9as (covering value for the filling gas of the *canister after helium backfilling in Section 4. 7).

During on-site transfer conditions, the canister is inside the CASTOR geo69 storage cask. 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 cask.

The canister and the cask filling gas temperatures for the different on-site transfer conditions (see Section 4.7) are lower than the temperatures used for storage conditions in the previous sections.

Therefore, the maximum absolute pressures inside the canister and the cask for on-site transfers are covered by the canister and cask pressures determined in the Sections 7.2 for NCS, 7.3 for off-normal conditions and 7.4 for ACS.

7.5 Containment Requirements for Short-Term Operations Section 7.5, Rev. 1 Page 7.5-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 7.6 Appendix Prepared Reviewed 7.6 Appendix Section 7.6, Rev. 1 Page 7.6-1

Non-Proprietary Version 1014-SR-00002 Rev. 1 Proprietary Information withheld per 10CFR 2.390 s

Appendix 7-1 Content Appendix 7-2 Determination of design pressure values Appendix 7-3 Activity mobilization Appendix 7-4 Assumptions 7.6 Appendix Section 7.6, Rev. 1 Page 7.6-2

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 Appendix 7-1 Content The activity content of the cask is described in Section 1.2.3. The spent uranium oxide (UOX) FA have a maximum heavy metal mass of * * *

• and a maximum FA-averaged final discharge burn-up of * * * *

• Based on the activity values per FA given in Section 1.2.3 the total activity content per cask loading with 69 FA is determined by multiplying the maximum value per nuclide over all FA types (FA No 1 to FA No 6 given in Section 1.2.3) with the number of FA per cask loading. These covering values are listed in Table 7.6-1.

Table 7 .6-1: Activity content of a cask loading with 69 FA For crud depositions, additional activity is taken into account. The surface specific crud activity is taken as 4.64

• 10 7 Bq/cm 2 (1254

• 10-5 Ci/cm 2 , see [1]). As 60 Co is the only significant contributor to crud activity after short cooling of the FA, the whole crud activity is assumed as 6°Co (also see

[11). Taking the fuel types in Section 1.2.3 into account, a conservative value for the crud activity of is estimated for a cask loading with 69 FA (kFA) by the following calculation:

With this conservative assessment, potential residual contamination on the inside surfaces of the package (e. g. from previous transports) is covered. According to [1], the decay of 6°Co is considered by the minimum time before loading of (cf. Section 1.2.3) and results in a crud activity of Page 1 of 3

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 The fission gas production is in very good approximation linearly dependent on the produced energy during reactor operation. The maximum fission gas masses are taken from Section 1.2.3 and summarized in Table 7.6-2 as specific gas production values for each fission gas (in unit mol/GWd) for UOX FA.

also given in Table 7.6-2.

Table 7.6-2: Specific production of fission gas The total produced amount of fission gas GFG in a loading is calculated via:

For the cask content compliant to the stipulations in Section 1.2.3, the maximum produced amount of fission gas is obtained for 69 UOX FA with a final discharge burn-up o f * * * *

• and a heavy metal mass of In this case, GFG is obtained.

The fuel rod filling gas helium is considered for the FA No 5 (GE 12) given in Section 1.2.3 as a covering amount for the other FA types. The gas volume in a fuel rod of land the number of fuel rods are given in Section 1.2.3. Higher values for the gas volume in a fuel rod of up to for other FA No in Section 1.2.3 result to a lower total gas volume for a FA because of lower numbers of FR per FA for those FA types.

a total amount of fuel rod filling gas for a cask loading Page 2 of 3

Non-Proprietary Version Propri~tary Information withheld per 10CFR 2.390 with 69 FA is calculated to by using the ideal gas law and set to GFR for the design pressure calculations in Sections 7.2, 7.3 and 7.4.

List of References

[1] NUREG-2215, April 2020 Standard Review Plan for Spent Fuel Dry Storage Systems and Facilities Page 3 of 3

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 Appendix 7-2 Determination of design pressure values Information on the maximum pressure Pu inside the canister and cask is required for various analyses.

The absolute pressure for the canister 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. In addition, the assumed maximum helium filling partial pressure of the canister is temperature-corrected by the gas temperature T gas under test conditions. Therefore, Pu for the canister is calculated with:

Pu= PHe,O

• Tgas / T He,O + n

• Runiv

• Tgas / V.

Procedural steps ensure a maximum absolute pressure of PHe,o inside the canister.

To determine the maximum absolute pressure Pu, the values of n and T gas are maximized while Vis 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 in Chapter 4, 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 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.

For maximum pressure considerations, the gas release from the content is calculated as follows:

- The fraction of failed fuel rods fs is assumed as 0.01 (1 %) for NCS, 0.1 (10 %) for off-normal conditions and as 1.0 (100 %) for ACS-impact and ACS-fire (see Table 7.6-3, according to [11).

- 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.

Page 1 of 2

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 The produced amount of fission gas is determined via burn-up calculations (cf.

Appendix 7-1).

- According to [1 ], the fraction of fission gas release fo is used as 0.15 (15 %) for NCS, off-normal and ACS-fire as well as 0.35 (35 %) for ACS-impact which includes an extra 20 %

fraction of the pellet-retained fission gases that may be released during a drop impact (see Table 7.6-3). 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.

After the canister 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 storage cask, either.

Procedural steps ensure a maximum absolute filling pressure of PHe,cask,o (cf. Chapter 9) inside the cask. With a free cask volume o f * * (cf. Chapter 1) a total amount of Helium in the cask results to about The assumed maximum helium filling partial pressure of the cask is temperature-corrected by the gas temperature in the cask T gas.cask under test conditions. Therefore, the absolute pressure Pu for the cask is calculated with:

Pu = PHe,cask,O

• T gas,cask / T He,cask,O.

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 irrapiation of the lid-end moderator disc, which is made of the ultrahigh molecular weight polyethylene - is given as about in Chapter 5. 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.

No residual water has to be assumed to be present as vapor after dispatch. Further gases are not formed during operation of the storage cask, either.

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 Page 2 of 2

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 Appendix 7-3 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 3 H, 85 K 129 substances from crud. Based on [1], the nuclides are classified as follows: and 1 as 106 134 137 6 gases, Ru, Cs and Cs as volatiles, all nuclides (except from °Co) from Table 7.6-,1 as particulate substances from fuel and 6°Co as particulate substance from crud (cf. Appendix 7-1 ).

In Appendix 7-2, the fraction of failed fuel rods fo and the fraction of fission gas release fo are introduced. According to [1],

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

• 10-5 for NCS, off-normal, ACS-impact and ACS-fire,

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

• 10-5 for NCS, off-normal, ACS-impact and an increased mass fraction of fF = 3

• 10-3 due to a conservatively assumed fuel oxidation for ACS-fire and

- the fraction of crud that spalls off rods is fc = 0.15 for NCS and off-normal conditions and fc = 1.0 for ACS-impact and ACS-fire.

The release fractions of radioactive materials that are considered in this analysis are summarized in Table 7.6-3.

Table 7 .6-3: Release fractions of radioactive materials

--1 Variable Fraction of Fuel Rods Symbol fs Normal Conditions (NCS) 0.01 Off-Normal Conditions (off-normal) 0.1 Accident-Fire Conditions (ACS-fire) 1.0 Accident-Impact Conditions (ACS-impact) 1.0 Assumed To Fail Fraction of Fission Gases Released fG 0.15 0.15 0.15 0.35 Due to a Cladding Breach Fraction of Volatiles Released fv 3E-05 3E-05 3E-05 3E-05 Due to a Cladding Breach Mass Fraction of Fuel Released fF 3E-05 3E-05 3E-03 3E-05 as Fines Due to a Cladding Breach Fraction of Crud fc 0.15 0.15 1.0 1.0 Spalling off Cladding .

Page 1 of 2

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 129 The activity concentration of the gases 3 H, 85 Kr and 1 is obtained by multiplying the total activity of the concerned nuclides from Table 7.6-1 with the fraction of failed fuel rods fs 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 fs 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.

List of References

[1] NUREG-2224, November 2020 Dry Storage and Transportation of High Burnup Spent Nuclear Fuel Page 2 of2

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 Appendix 7-4 Assumptions

- The crud activity of is estimated for a cask loading (cf. Appendix 7-1).

- An initial absolute fuel rod filling gas pressure of is assumed for all fuel rods of all FA in the cask loading (cf. Appendix 7-1 ).

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

Appendix 7-2).

Page 1 of 1

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev. 1 8 Materials Evaluation 8.0 Overview Prepared Reviewed 8.0 Overview Section 8.0, Rev. 1 Page 8.0-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev: 1 The materials evaluation presented in this Chapter ensures that materials will perform in a manner that supports the functions of the SSCs of the CASTOR geo69 DSS and the CLU. Following a brief summary of the system design in Section 8.1 this Chapter provides information on materials of con-struction, including mechanical properties, thermal properties and the technical basis for material properties. The material properties summarized in Section 8.2 form the basis for the structural, ther-mal, shielding, criticality and containment evaluation of the DSS and the CLU, if applicable.

Section 8.3 demonstrates that the materials will not undergo adverse environmental degradation, chemical reactions, or other reactions that could challenge the ability of SSCs to safety handle, package, transfer, and store the SNF over the intended storage period.

Information on SNF classifications, properties of the fuel rod cladding and the cover gas are given in Section 8.4 8.0 Overview Section 8.0, Rev. 1 Page 8.0-2

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev. 0 @GNS 8.1 System Design Prepared Reviewed

8. 1 System Design Section 8.1, Rev. 0 PageB.1-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 8.1.1 Drawings The principle design of the CASTOR geo69 DSS and the CLU is described in Section 1.2. Drawings and corresponding parts lists of the DSS and CLU components are included in Section 1.5. The parts lists and drawings include material specifications, alternatives, welding instructions and non-destructive examination (NOE) requirements.

8.1.2 Codes and Standards The codes and standards applicable for the CASTOR geo69 DSS design are listed in Table 8.1-1.

The containment system is designed in accordance with Division 3, Subsection WC, considering the underlying requirements with respect to mechanical, design, material and fabrication issues. Seals and gaskets that are part of the containment system are leak tested in accordance with ANSI N14.5

[1 ]. Special lifting devices are designed in accordance with ANSI N14.6 [2]. For materials not covered by the BPVC, material data according to the manufacturer's catalogue and test data are used.

Table 8.1-1: Applicable codes and standards for the desiijn of the CASTOR geo69 DSS Component Applicable Codes and Standards for Design Containment system with bottom/closure Division 3, Subsection WC plate Trunnions (storage cask)

ANSI N14.6 [2] as applicable LAP of the canister lid Industry standards LAP of the cask lid ANSI N14.5 [1] as applicable Seals and gaskets Material data according to manufacturer's catalogue Test data Moderator Material data according to manufacturer's catalogue GNS proprietary design methodology report Fuel basket without shielding elements Material data according to manufacturer's catalogue

• Test data Respective material standards including additional manu-Shielding elements facturer's catalogue, as applicable Protection cover Industry standards The CLU is designed in accordance with the applicable requirements for supports used during the handling of SNF. The structural design of the transfer cask and transfer lock follows the requirements according to the ASME BPVC, Section Ill, Division 1, Subsection NF [3]. The code is considered regarding the underlying requirements with respect to mechanical, design, material and fabrication issues. The applicable codes and standards for the design of the transfer cask and the transfer lock are listed in Table 8.1-2.

8.1 System Design Section 8.1, Rev. 0 Page 8.1-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 1 OCFR 2.390 Rev. 0 Table 8.1-2: Applicable codes and standards for the design of the CLU Component Applicable Codes and Standards for Design Structural components of transfer cask and ASME BPVC Section 111, Division 1, Subsection NF [3]

transfer lock Trunnions of the Transfer Cask ANSI N14.6 [2] as applicable Load attachment points of the transfer lock Industry standards Lead shield PB940R acc. DIN EN 12659 [ASTM B-29]

8.1.3 Welding 8.1.3.1 Fabrication and examination of welds in the DSS Containment welds of category A, Band C according to Division 3, Subsection WC-3251 exist in the canister of the DSS. The canister body consists of a welded stainless steel construction The welding fabrication specifications included in the drawing of the canister (see Section 1.5) and the ac-ceptance criteria for containment welds, as specified in Section 10.1, are consistent with the require-ments of Division 3.

Only gas tungsten and submerged arc welding consumables according to SFA-5.9M [4] are used as welding material for the fabrication of the containment welds in the canister. Only welding processes that are capable of producing welds in accordance with the welding procedure qualification require-ments of Section IX [5] and Division 3, Subsection WC-4000 are permitted. The manufacturer chooses the welding procedure under consideration of this requirement.

8.1.3.2 Fabrication and examination of welds in the CLU Transfer cask and transfer lock consist of a welded steel construction. The transfer cask exhibits primary and secondary member welded joints according to Subsection NF-1215 [3], whereas the structural skeleton of the transfer lock only exhibits secondary member welded joints. The drawings of the CLU components (see Section 1.5) include the fabrication specification and the required NOE procedure for each welded joint. The acceptance standards for NOE of welds in the transfer cask and the transfer lock, as specified in Section 10.1, are in accordance with the acceptance standards of Subsection NF-5300.

Only welding processes that are capable of producing welds in accordance with the welding proce-dure qualification requirements of Section IX [5] and Division 1, Subsection NF-4000 are permitted for welds in the transfer cask and the transfer lock. The manufacturer chooses the welding procedure under consideration of this requirement.

8.1 System Design Section 8.1, Rev. 0 Page 8.1-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 .

List of References

[1] ANSI N14.5-2014 American National Standard for Radioactive Materials -

Leakage Tests on Packages for Shipment

[2] ANSI N14.6-1993 Radioactive Materials - Special Lifting Devices for Shipping Containers Weighing 10 000 Pounds (4500 kg) or More

[3] ASME Boiler and Pressure Vessel Code (Edition 2017)

Section Ill Rules for Construction of Nuclear Facility Components Division 1 - Subsection NF

[4] ASME Boiler and Pressure Vessel Code Section II Part C, Edition 2017 SFA-5.9M "Specifi-cation for Bare Stainless Steel Welding Electrodes and Rods"

[5] ASME Boiler and Pressure Vessel Code (Edition 2017)

Section IX Welding, Brazing, and Fusing Qualifications 8.1 System Design Section 8.1, Rev. 0 Page 8.1-4

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev.1 @GNS 8.2 Material Properties Prepared Reviewed 8.2 Material Properties Section 8.2, Rev. 1 Page 8.2-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 This Section documents the material data to be used for the evaluations of the CASTOR geo69 DSS and the CLU. Specific materials of the parts according to the design parts lists (see Section 1.5, Appendixes 1-4 to 1-10) of cask, canister, basket, shielding elements and protection cover of the CASTOR geo69 DSS as well as of the transfer cask and transfer lock of the CLU are consid-ered.

If not stated otherwise, the applicable material properties according to the respective requirements of BPVC,Section II and Division 3 are taken into account. For applied materials other than specified as described above, the properties are considered as given below.

In the range of validity of the properties (e.g. temperature), the data can be interpolated linearly.

8.2.1 Properties of ASME Materials In this regard, reference is made to the BPVC,Section II and Division 3 and to Section 8.5, Appendix 8-1 of this SAR.

8.2.2 Properties of non-ASME Materials The following Subsections tabulate all properties of non BPVC,Section II materials that are relevant to the mechanical, thermal, shielding and chriticality evaluations in the respective Chapters of this SAR.

8.2.2.1 Aluminum EN AW-5051A Table 8.2-1: Physical Properties acc. to [1]

Specific Yield Tensile Thermal Heat Temperature Elongation 11 Heat Strength 11 Strength 11 Expansion Conductivity Capacity T Rt,0.2 Rm a A, Cp

[OCJ [MPa] [MPa] [%] [1(J 6l°CJ [W/(m K)] [J/(kg K)]

20 60 150 16 - 138 900 100 - - - 23.8 147 960 150 - - - - 151 -

200 - - - 24.8 155 990 210 45 90 20 - - -

1) Values differing from material standard, have to be determined in manufacturing 8.2 Material Properties Section 8.2, Rev. 1 Page 8.2-2

Non-Proprietary Version .

1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Temperature Limit: Tumit [ C] = 121 0

Poisson's Ratio v [-] = 0.33 Density p [kg/ dm3 ] = 2.68 Chemical composition according to [1] as given in Table 8.2-3.

8.2.2.2 Aluminum EN AW-5454, condition 0 Table 8.2-2: Physical Properties acc. to [2]

Yield Tensile Thermal Heat Con- Specific Heat Temperature Elongation Strength Strength Expansion ductivity Capacity T R,,0.2 Rm a ,t q,

[°CJ [MPa] [MPa] [%] [10-61°CJ [W/(m K)] [Jl(kg K)]

20 85 215 - 275 16 23.6 134.0 898 50 137.6 909 75 140.8 923 100 143.6 934 125 145.9 942 150 148.0 951 175 150.0 961 200 151.9 972 Temperature Limit: Tumit [°C] = 121 Poisson's Ratio v [-] = 0.33 Density p [kg/ dm3 ] = 2.69 Chemical composition according to [2] as given in 8.2-3.

Table 8.2-3: Chemical composition for Aluminium alloys in wt-%

Alloy Si Fe Cu Mn Mg Cr Zn Ti Add. Bal.

0.10 1.60 ENAW- 0.30 0.45 0.05 0.30 0.20 0.10 5051A max max max

~ ~ max max max s 0.15 Al 0.25 2.10 0.50 2.40 0.05 ENAW- 0.25 0.40 0.10 0.25 0.20 5454 max max max

~ ~ ~ max max

- Al 1.00 3.00 0.20 8.2 Material Properties Section 8.2, Rev. 1 Page 8.2-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 8.2.2.3 Ultra High Molecular Weight Low Pressure Polyethylene GUR 4120, tempered Table 8.2-4: Physical Properties acc. to Appendix 8-3 8.2 Material Properties Section 8.2, Rev. 1 Page 8.2-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 8.2.2.4 Stainless Spring Steel 1.4310 Table 8.2-5: Physical Properties acc. to [3]

Tensile Thermal Thermal Modulus of Specific Heat Temperature Density Strength Conductivity Expansion Elasticity Capacity T Rm A, a E Cp p

[°CJ [MPa] [Wl(m*K)] [10-SJ<'CJ [1ci3 MPa] [J/(kg K)] [kg/m 3]

1300 -

20 14.5 16.1 196 472 7920 1500 100 - 16.0 16.7 - 501 -

200 - 17.6 17.2 - 525 -

300 - 19.1 17.7 - 532 -

• 400 Table 8.2-6:

- 20.4 Chemical Composition acc. to [3]

18.1 - 555 -

Chemical composition

[wt-%]

min max C 0.05 0.15 Si 2.00 Mn 2.00 p 0.045 s 0.015 Cr 16.0 19.0 9.50 Ni 6.00 N 0.11 Mo 0.80 8.2 Material Properties Section 8.2, Rev. 1 Page 8.2-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 8.2.2.5 Aluminium-Boron Metal Matrix Composite Al-84C-MMC Table 8.2-7: Physical Properties acc. to Appendix 8-2 8.2 Material Properties Section 8.2, Rev. 1 Page 8.2-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 8.2-8: Chemical Composition acc. to Appendix 8-2 8.2 Material Properties Section 8.2, Rev. 1 Page 8.2-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @)GNS 8.2.2.6 Nickel alloy 2.4969 Table 8.2-9: Physical Properties acc. to Appendix 8-4 Tensile Density Strength (RT) p Rm

[kgldm3] [MPa]

8.20 Table 8.2-10: Chemical Composition acc. to Appendix 8-4 Chemical composition

[wt-%]

min max Cr 18.0 21.0 Fe - 2.00 Ti 2.00 3.00 Mn - 1.00 Si - 1.00 C - 0.13 Al 1.00 2.00 Co 15.0 21.0 s - 0.015 Cu - 0.2 B - 0.03 Pb - 0.0025 Zr - 0.15 Ni bal.

8.2 Material Properties Section 8.2, Rev. 1 Page 8.2-8

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @)GNS 8.2.2.7 Silver Table 8.2-11: Physical Properties acc. to Appendix 8-4 Hardness HV 50 +/-10 Table 8.2-12: Chemical Composition acc. to Appendix 8-4 Chemical composition Density

[wt-%] [kgldm3J Ag I 99.99 10.49 8.2.2.8 Lead ASTM B 29 (UNS L50049), EN 12659 Pb940R Table 8.2-13: Physical Properties acc. to [4]

Heat Specific Heat Temperature Conductivity Capacity T A, Cp

[°CJ [Wl(mK)J [Jl(kg K)J 0 33.5 -

25 - 129 127 - 132 150 33.0 -

200 -

227 - 137 250 31.8 -

300 - -

326 31.0 142 ASTMB 29 EN 12659 PB940R (UNS L50049)

Density p 11.33 11.34

[kg/ dm3 ]

8.2 Material Properties Section 8.2, Rev. 1 Page 8.2-9

Non-Proprietary Version 1014-S.R-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 8.2-14: Chemical composition acc. to [5]

ASTMB 29 EN 12659 PB940R (UNS L50049)

Element Max. content [wt-%]

Sb 0.0010 0.0010 As 0.0010 0.0010 Sn 0.0010 0.0010 Cu 0.0015 0.0050 Ag 0.0100 0.0080 Bi 0.0500 0.0600 Zn 0.0010 0.0005 Ni 0.0005 0.0020 Fe 0.0010 -

Se 0.0010 -

s 0.0020 -

Al 0.0005 -

Cd 0.0005 0.0020 Pb bal. bal.

8.2.2.9 High-Density-Polyethylen (PE-HD)

PE-HD is used as neutron shielding material in the transfer lock. All the properties relevant for shield-ing evaluations are identical to those of GUR 4120. PE-HD is not part of any further evaluations.

Thus it is reffered to Subsection 8.2.2.3.

8.2 Material Properties Section 8.2, Rev. 1 Page 8.2-10

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 8.2.3 Emmission Coefficients This subsection tabulates the emission coefficients of all materials used for the components of the DSS and CLU.

Table 8.2-15: Emission Coefficients Type/Grade I Alloy Emission Material Standard Reference (Denomination) Coefficient e [-]

Carbon Steel SA-182M F60 I

Stainless Steel SA-182M F304 I I Stainless Steel SA-182M F316 I I Stainless Steel SA-182M F316L I I Stainless Steel SA-182M FXM-19 I I Stainless Steel SA-182M F6NM I I Chromium Steel SA-193M B6 I I Chromium Steel SA-193M B7 I I Low Alloy Carbon Steel SA-193M Gr. B8 Cl. 1, Cl. 2 I I Chromium Steel SA-194M 6 I I Stainless Steel SA-240M (UNS S31803)

I I I

Stainless Steel Stainless Steel Stainless Steel SA-240M SA-240M SA-240M 304 304L 316 II I I

Stainless Steel SA-240M 316L I I Stainless Steel SA-240M XM-19 I I Stainless Steel SA-350M Gr. LF2 I I Stainless Steel SA-479M 304 I I Stainless Steel SA-479M 316L I I Stainless Steel SA-479M (UNS S41500)

I I I

Stainless Steel Low Alloy Carbon Steel Low Alloy Carbon Steel SA-479M SA-516M SA-540M XM-19 Gr. 70 [485]

B22 Cl. 3 I

I II Carbon Steel SA-675M 60 I I Low Alloy Carbon Steel SA-738M Gr. C I I Cast Iron Stainless Steel SA-874M SA-965M (JIS G 5504) 304 II II Stainless Steel SA-965M 316 I I Stainless Steel SA-965M 316L 8.2 Material Properties Section 8.2, Rev. 1 Page 8.2-11

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Type/Grade I Alloy Emission Material Standard Reference (Denomination) Coefficient e [-]

Stainless Steel SA-965M FXM-19 Stainless Steel EN 10270-3 1.4310 Aluminum SB-209 5454 Aluminum EN 573-3 AW-5051A Aluminum EN 573-3 AW-5454 UHMW LP-PE GUR 4120 Alum.-Boron Metal Matrix AI-B4C-MMC Composite

• Within the thermal evaluations s = 0,55 is conservatively used.

8.2.4 Bolt Applications All materials used for bolt applications or SSC important to safety follow the requirements for the mechanical properties, temperature limits and design stress intensity limits listed in BPVC,Section II, Part D, Table 4. They all have adequate resistance to corrosion (c.f. Section 8.3) and brittle fracture and a coefficient of thermal expansion similar to the materials being bolted together.

8.2.5 Seals 8.2.5.1 Metal gaskets Sealing of the cask and cannister containments is provided by a multi-layer metal gaskets (c.f. Ap-pendix 8-4) that are important to safety. They consist of an The spring is made o f * * * * * (Subsection 8.2.2.6), a nickel alloy which is a flexible material achieving and maintaining tight contact between the gasket sur-face and the corresponding surfaces of the counterparts.

8.2 Material Properties Section 8.2, Rev. 1 Page 8.2-12

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 1.0CFR 2.390 Rev. 1 @GNS Table 8.2-16 Design relevant properties for the respective metal gasket 8.2.5.2 Elastomeric seals All elastomeric seals used with the DSS and CLU are not important to safety.They are made ofVMQ and FKM, which are suitable to. the expected thermal and radiation conditions (c.f. Section 8.3) 8.2 Material Properties Section 8.2, Rev. 1 Page 8.2-13

Non-Proprietary Version 1014-SR-00002 Rev. 1 Proprietary Information withheld per 10CFR 2.390

@G S 8.2.6 Coatings 8.2.6.1 Surfaces of Storage Cask and Transfer Cask The outer and inner surf~ces of storage cask and transfer cask are coated for various reasons. On the one hand the paint protects the DCI from corrosion, on the other hand it provides a good and defined decontamination capability. Furthermore, the thermal evaluations take credit from the coat-ing regarding the emission coefficients as they are more favorable in terms of heat dissipation com-pared to pure DCI or stainless steel. The coatings have to fulfil the following requirements:

Outer surface of storage cask and transfer cask

• Stability over the entire intended storage time (60 years according to this SAR)

• Very god capability of decontamination according to DIN 25415 [12]

• Good mechanical resistance and adhesion

• Possibility of repair

• Corrosion category C3 (very long > 25 years) according to ISO 12944-6 [13]

• Emmission coefficient .:: 0.93 Inner surface of storage cask and transfer cask

• Stability over the entire intended storage time (60 years according to this SAR)

• Good mechanical resistance and adhesion

• Possibility of repair

• Corrosion category C3 (very long > 25 years) according to ISO 12944-6 [13]

• Emmission coefficient .:: 0.6 8.2.6.2 Lubricants Prior to assembly the threads of screws and bolts as well as the undersides of the screw heads shall be coated with lubricants. Only the tow lubricants MOLYKOTE D 21 and MOLYKOTE D 321 Rare permitted for application. As discussed in Section 8.3 they have no significant effect on the cask materials to which they are in contact with. The structural evaluation takes these two lubricants into account with regard to the coefficients of friction to be applied in the calculations.

8.2 Material Properties Section 8.2, Rev. 1 Page 8.2-14

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @)GNS List of References

[1] EN 573-3, 2019-08, Aluminium and aluminium alloys - Chemical composition and form of wrought products - Part 3: Chemical composition and form of products Material data base WIAM-Metallinfo 2002/2.2 EN 755-2, 2016-03, Aluminium and aluminium alloys - Extruded rod/bar, tube and profiles

- Part 2: Mechanical properties ASME Boiler and Pressure Vessel Code Section II Part D, Properties (metric), Edition 2017 Dubbel, Taschenbuch fur den Maschinenbau, Seite E 102, 20. Auflage, 2001 Aluminium-Taschenbuch 1995

[2] EN 573-3, 2019-08, Aluminium and aluminium alloys - Chemical composition and form of wrought products - Part 3: Chemical composition and form of products EN 485-2, 2018-10, Aluminium and aluminium alloys - Sheet, strip and plate - Part 2: Me-chanical properties ASME Boiler and Pressure Vessel Code Section II Part D, Properties (metric), Edition 2017 Aluminium-Taschenbuch, Band 1, Aluminium-Verlag Dusseldorf, 1995 Aluminium Werkstoffdatenblatter, Aluminium-Verlag Dusseldorf, 1998

[3] Stahl-Eisen-Werkstoffblatt SEW 310, 08.1992 EN 10270-3, 2011-10, Steel wire for mechanical springs - Part 3: Stainless spring steel wire

[4] ASTM Volume 02.04 Nonferrous Metals B 29 "Standard Specification for Refined Lead".

EN 12659, Edition 1999-09, Lead and lead alloys - Lead Values for heat conductivity und specific heat capacity: "VOi Heat Atlas Calculation Sheets for the Heat Transfer", Springer (2010)

[5] ASTM Volume 02.04 Nonferrous Metals B 29 "Standard Specification for Refined Lead".

EN 12659, Edition 1999-09, Lead and lead alloys - Lead Values for heat conductivity und specific heat capacity: "VOi Heat Atlas Calculation Sheets for the Heat Transfer", Springer (2010)

[6] Kern, D. Q.: "Process Heat Transfer", McGraw-Hill, Kogakusha/Japan (1965)

[7] VOi Warmeatlas, Ke4 ,,Strahlung technischer Oberflachen, VOi-Veriag Dusseldorf/Ger-many, 2002

[8] Gosse, J.: "Thermodynamik-Kompendium, VOi-Veriag Dusseldorf/Germany, 1986

[9] "Emissionsgradtabelle", VEB MeBgeratewerk "Erich Weinert" Kombinat Elektro-Apparate-Werke "Friedrich Ebert", Berlin/Germany

[1 O] White, F. M.: "Heat and Mass Transfer", Addison Wesley, 1991

[12] DIN 25415:2012-11 Radioactively contaminated surfaces - Method for testing and assessing the ease of decon-tamination 8.3 Environmental Degradation Section 8.3, Rev. 1 Page 14.0-15

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS

[13] ISO 12944-6:2018-06 Paints and varnishes - Corrosion protection of steel structures by protective paint systems -

Part 6: Laboratory performance test methods 14.0 Overview Section 14.0, Rev. 1 Page 8.2-16

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 8.3 Environmental Degradation Prepared

• Reviewed 8.3 Environmental Degradation Section 8.3, Rev. 1 Page 8.3-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 8.3.1 Chemical, Galvanic and other Reactions The materials of the CASTOR geo69 DSS and CLU have been reviewed and as a result, no safety related component is significantly influenced by chemical, galvanic or other reactions during loading and storage operations.

During the loading operations the basket, canister and transfer cask are in contact with the pool water. These materials are stainless steels, aluminium alloys, aluminium boron carbide metal-matrix-composite (MMC) and different polymers with high chemical resistance. These materials are all com-patible with the pool water. The Lead shield within the transfer cask is not exposed to the pool water.

The transfer lock is made of different stainless steels, coated carbon steel and polymer compounds.

No Influence of the ambient conditions on the materials is implied.

During storage, the storage casks are equipped with a protection cover. The components are either stainless steel or coated carbon steel and polymers. No influence of the ambient conditions on the materials made from stainless steel is implied. The coating is chosen to have high resistance against e.g. UV irradiation and influences from atmospheric conditions. Nevertheless, the coating is acces-sible from the outside, thus can be inspected visually and repaired if necessary.

The inner surface of the cask is appropriately coated. The interior of the CASTOR geo69 DSS is dried and filled with helium after the loading operations. This provides a dry and inert environment during storage. Corrosion reactions depend on the presence of water and/or oxygen. The dry inert helium gas atmosphere in the CASTOR geo69 DSS precludes corrosion during storage. Exterior surfaces and materials consist of stainless steels, aluminium alloys or coated materials. Therefore, chemical, galvanic or other reaction do not have to be assumed to a significant extent. Nevertheless the components are accessible and can be exchanged or repaired.

The materials of the CASTOR geo69 DSS and CLU are summarized in Table 8.3-1. In presence of water, dissimilar materials can form a galvanic couple. During loading, the aluminium alloys and the aluminium boron carbide (MMC) form a galvanic couple with stainless steel. Both types of aluminium develop a native passive layer that precludes significant corrosion effects. To minimize galvanic and other corrosive reactions, the aluminium components, which are in contact with stainless steels, are additionally anodized. In consequence, no significant galvanic reactions to the aluminium alloys or the stainless steels occur during the loading time.

The storage cask body is made of ductile cast iron and the exterior surfaces will be coated to pre-clude corrosion reactions at the surfaces. The cask c~vity is filled with helium and no corrosion has to be assumed.

8.3 Environmental Degradation Section 8.3, Rev. 1 Page 8.3-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The lids of the storage cask are made of stainless steel and are in contact with the zinc-coated alloyed steel bolts. The bolts of the lid are not directly exposed to the ambient weather, because they are covered by the protection cover during storage. Thus, no significant corrosion effects are to be implied for the bolts.

During loading operations of the canister, before drying and backfilling with helium, only minor amounts of hydrogen gas will be generated due to the minimized galvanic reaction of the aluminium and stainless steel and due to radiolysis of the water. This hydrogen will be evacuated from the canister during the drying process and no significant concentration of hydrogen can occur.

The lid system of the storage cask includes different auxiliary sealing rings and polymeric com-pounds which possess no safety related function. They simplify and shorten the dispatch of cask and canister (e.g. leak testing subsequent to loading). During the whole operational period of the cask theses components might deteriorate due to the impact of temperature and radiation. From this degradation no chemically reactive compound is generated an thus the adjacent materials are not negatively affected.

Lubricants are used to coat the screw threads. Only permitted lubricants are used for the coating of the screw threads. Before assembly or loading, all cask components will be inspected and freed from any form of contamination or marking. The lubricants have no significant effect on the cask materials.

There are no significant chemical, galvanic or other reactions that could reduce the integrity of the cask during the loading and storage operations.

Table 8.3-1:. Environment during loading and storage for DSS and CLU components Material I Component Environment during loading Environment during storage High alloyed stainless Steels: Stainless steels in contact with both The environment for these compo-borated and unborated water do not nents will be an inert helium at- .

SA-182M FXM-19 exhibit chemical or galvanic reactions mosphere. No further chemical, SA-193M B6 or interactions with spent fuel. galvanic or other reactions are as-SA-194M 6 sumed.

SA-240M 304 SA-240M 316 SA-240M 316L SA-240M XM-19 SA-479M 304 SA-479M 316L SA-479M XM-19 SA-965M FXM-19 Basket, canister 8.3 Environmental Degradation Section 8.3, Rev. 1 Page 8.3-3

Non-Proprietary Version 1014-SR-00002 Rev. 1 Proprietary Information withheld per 10CFR 2.390 s

Material I Component Environment during loading Environment during storage Aluminium boron carbide The aluminium boron carbide (MMC) The environment for these com po-(MMC) forms a galvanic couple with stainless nents will be an inert helium at-steels. The aluminium will beano- mosphere. No further chemical, dized to minimise any form of galvanic galvanic or other reactions are as-or other corrosion reactions. Due to sumed.

the short loading time, in which they are in contact with pool water, the neutron absorber material is not ex-posed to significant chemical, gal-Basket vanic or other reactions.

Aluminium alloys: The aluminium alloy forms a galvanic The environment for these compo-couple with stainless steels. These al- nents will be an inert helium at-SB-209 Alloy 5454 uminium components will be anodized mosphere. No further chemical, to minimise any form of galvanic or galvanic or other reactions are as-other corrosion reactions. Due to the sumed.

limited loading time in which they are in contact with pool water, the mate-rial is not exposed to significant chem-ical, galvanic or other reactions.

Basket, shieldinq elements Steels: These components are not in contact The environment for these compo-with the pool water. nents will be an inert helium at-SA-193M B7 mosphere. No further chemical, SA-540M B22 galvanic or other reactions are as-sumed.

Canister High alloyed stainless steels: Stainless steels in contact with both The transfer cask is not part of the borated and unborated water do not DSS.

SA-182M FXM 19 exhibit chemical or galvanic reactions SA-182M F316L or interactions with spent fuel. Components are exposed to ambi-SA-182M F304 ent conditions. Stainless steels ex-SA-182M F6NM hibit a native corrosion protection SA-193M B6 layer and no galvanic or other cor-SA-240M 316L rosion reactions have to be as-SA-240M XM 19 sumed. No chemical or other reac-SA-479M UNS S41500 tions are assumed.

SA-965M FXM19 SA-965M F316L SA-965M F304 Other stainless steel Transfer Cask Lead: The lead shield is not in contact with The transfer cask is not part of the pool water (enclosed by stainless DSS.

PB 940R steel components).

ASTM B-29 Components are not exposed to ambient conditions. No chemical or Transfer Cask other reactions are assumed.

8.3 Environmental Degradation Section 8.3, Rev. 1 Page 8.3-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Material I Component Environment during loading Environment during storage Aluminium Not in contact with pool water. The transfer lock is not part of the DSS.

Components are exposed to ambi-ent conditions. Aluminium and its alloys exhibit a native corrosion protection layer and no corrosion reactions have to be assumed. No Transfer Cask chemical or other reactions are as-sumed.

Polymers: Compounds with high chemical re- The transfer cask is not part of the sistance. No degradation when in DSS.

lglidur X contact with pool water is implied.

PTFE (Replaceable when transfer cask is The materials are exposed to am-FPM not in use) bient condition but no reactions are Klingersil C-4400 implied.

e'I VMQ Transfer Cask High alloyed stainless steels: Not in contact with pool water. The transfer lock is not part of the DSS.

SA-182M F304 SA-182M F316 Components are exposed to ambi-SA-182M F316L ent conditions. Stainless steels ex-SA-193M B6 hibit a native corrosion protection SA-194M 6 layer and no galvanic or other car-SA-240M F304 rosion reactions have to be as-SA-240M F316 sumed. No chemical or other reac-SA-240M F316L tions are assumed.

SA-479M F304 SA-479M F316L SA-516M 485 SA-540M B22 Cl. 3 Other stainless steel Transfer lock Carbon steel: Not in contact with pool water. The transfer lock is not part of the DSS.

SA-36M SA-516M 485 Material is appropriately coated SA-738M C and thus protected from ingress of moisture. Corrosion and either re-Transfer lock actions are excluded.

Polymers: Not in contact with pool water. The transfer lock is not part of the (Replaceable when transfer lock is not DSS.

lglidur X in use)

PE HD The materials are exposed to am-PA bient condition but no reactions are Rubber implied.

Transfer lock 8.3 Environmental Degradation Section 8.3, Rev. 1 Page 8.3-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Material I Component Environment during loading Environment during storage Ductile cast iron: This component is not in contact with The internal surfaces will be appro-the pool water. priately coated and the coating will SA-874M be exposed to an inert helium at-mosphere. No reactions are im-plied.

External surfaces will be appropri-ately coated and will be maintained with a fully coated surface.Coating is chosen to withstand the implied atmospheric influences and UV ir-radiation. No further chemical, gal-van ic or other reactions are as-sumed. External coating is acces-sible and can be repaired if neces-Storaqe Cask sarv.

Polyethylene: This component is not in contact with The neutron moderator material the pool water. has no contact to the outer envi-ronment. No chemical or galvanic reactions with DCI or steels are as-sumed. Other reactions are insig-Storage Cask nificant.

Polymer: Compounds with high chemical re- Auxilliary components which might sistance. No degradation when in decompose over time but have no FKM, VMQ contact with pool water is implied. safety related function.

Storaqe Cask Carbon steel: This component is not in contact with Material is appropriately coated the pool water. and thus protected from ingress of SA-738M C moisture. Coating is chosen to withstand the implied atmospheric influences and UV irradiation. No further chemical, galvanic or other reactions are assumed. Coating is accessible and can be repaired if necessary.Corrosion and other re-Protection cover actions are excluded.

High alloy stainless steel: This component is not in contact with Components are exposed to ambi-the pool water. ent conditions. Stainless steels ex-SA-194M 6 hibit a native corrosion protection layer and no significant galvanic or other corrosion reactions have to be assumed. No chemical or other Protection cover reactions are assumed.

8.3 Environmental Degradation Section 8.3, Rev. 1 Page 8.3-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Material I Component Environment during loading Environment during storage High alloyed stainless steels: This component is not in contact with Some parts are exposed to ambi-the pool water. ent weather. Stainless steels ex-SA-182M F60 hibit a native corrosion protection SA-182M F304 layer and no significant galvanic or SA-182M F316 other corrosion reactions have to SA-182M F6NM be assumed. No chemical or other SA-193M B6 reactions are assumed.

SA-240M 304 SA-240M 316 SA-240M UNS S31803 SA-479M 304 SA-479M 316 SA-479M 316L SA-479M UNS S41500 SA-965M F304 SA-965M F316 EN ISO 10270-3 1.4310 Storage Cask Constructional Steel This component is not in contact with This material has no contact to the the pool water. outer environment. No significant SA-675M 60 chemical or galvanic reactions with ductile cast iron or other steels are assumed. No other reactions are StoraQe Cask assumed.

Steels: This component is not in contact with Material is appropriately coated the pool water. and thus protected from ingress of SA-193M B7 moisture. Coating is chosen to SA-540M B22 withstand the implied atmospheric influences and UV irradiation. No Storage Cask further chemical, galvanic or other reactions are assumed. Coating is accessible and can be repaired if necessary.Corrosion and other re-actions are excluded.

Silver, stainless steel and All materials of the gasket exhibit a All materials of the gasket exhibit a nickel based alloy native corrosion protection layer and native corrosion protection layer no galvanic or other corrosion reac- and no galvanic or other corrosion tions have to be assumed. No chemi- reactions have to be assumed. No cal or other reactions are assumed. chemical or other reactions are as-Metal Gaskets sumed.

Organic Coating The coating is not in contact with the The coating on external surfaces is pool water. exposed to ambient weather but exhibits a good long-term durabil-ity.

Storage Cask Discoloration is not a concern.

Transfer lock No further chemical, galvanic or Protection cover other reactions are assumed.

8.3.2 Effects of Radiation on Materials Metals are not impaired by gamma radiation. Significant radiation damages due to neutron exposure is not expected for metals at neutron fluences below 1018 n/cm 2 . The expected neutron fluences of 8.3 Environmental Degradation Section 8.3, Rev. 1 Page 8.3-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 the components of the CASTOR geo69 DSS will be significantly lower (c.f. Section 5.4).

Elastomeric seals are exposed to gamma radiation and thus may undergo degradation. The elasto-meric seals do not have safety related functions and the degradation products have a similar com-position as the original molecule but different crosslinking or chain length. Thus, no harmful degra-dation products are expected.

The neutron shielding material polyethylene may be affected by irradiation analogously to elasto-meric materials. The irradiation of the neutron shielding amounts approximately The impairment through irradiation is insignificant and a loss of the I

neutron shielding ability does not have to be implied.

External coating is not exposed to significant amount of radiation from the cask inventory. The coat-ing will be chosen to be adequate for the implied UV-irradiation. Nevertheless, the coating is acces-sible from the outside, thus can be inspected visually and repaired if necessary.

There is no significant extend of degradation of any important to safety related component caused directly by the effect of the reactions. The same applies for the effects of reactions combined with the effects of exposure of the materials to neutron or gamma radiation.

8.3 Environmental Degradation Section 8.3, Rev. 1 Page 8.3-8

1014-SR-00002 Non-Proprietary Version Rev.1 Proprietary Information withheld per 10CFR 2.390 @GNS 8.4 Fuel Cladding Integrity Prepared Prepared 8.4.3 only Reviewed Reviewed 8.4.3 only 8.4 Fuel Cladding Integrity Section 8.4, Rev. 1 Page 8.4-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 8.4.1 Spent Fuel Classification The bounding cask array described in Section 1.4 comprises - CASTOR geo69 storage casks and provides interim storage capacity for up to - fuel assemblies. The generic ISFSI uses CASTOR geo69 casks for the storage of SNF.

Only undamaged FA complying with the limits specified in Subsection 1.2.3 are stored in the canister of CASTOR geo69 casks. FA which have been deformed or damaged during reactor operation or which are otherwise defective in their structural integrity and hence are not able to meet the pertinent fuel-specific or DSS-related regulations are not to be loaded into the cask. It is only allowed to load undamaged FA with completely filled grids into the cask. However, it is allowed to load undamaged FA with completely filled grids containing replacement fuel rods and/or replacement rods manufac-tured from solid material (dummy rods) used to displace an amount of water greater than or equal to that displaced by the original fuel rod(s).

Damaged FA, fuel debris and associated nonfuel hardware are not to be stored in the fuel canister of CASTOR geo69 storage casks.

During the operation of the NPP, fuel integrity has been, and continues to be, monitored. Through the detection of radiochemistry changes in the reactor coolant system, most fuel damage is as-sessed. When damaged rods are suspected, assemblies are inspected as they are removed from the core. All assemblies with positive indication of damage are again inspected in the spent fuel pool to determine amount and location of rods in the assembly that have failed cladding. If the FA is to be placed back in the reactor core, any failed rods are removed and replaced with nonfuel rods of equivalent dimensional properties. If the suspected damaged FA are at the end of their cycle, the assemblies may be stored in the spent fuel pool without repair. During this process, all known rod failures are noted and their assemblies are tracked. If the failure is visible from the exterior of the assembly, the damage may be videotaped. For assemblies that were removed from the reactor core and were not inspected at that time, inspections will be performed prior to loading these assemblies into the storage cask. This will ensure that there are no undetected failed rods in any assembly that is placed into a cask.

Under this failure detection process, inspections to date have found a limited amount of failures.

Where single failed rods have been identified and removed, they were stored in the spent fuel pool and would ultimately be stored in a canister that can contain fuel debris, but not in the canister of CASTOR geo69 storage cask.

This detection process, along with the history of plant operations and spent fuel storage in the fuel pool, provide a high level of confidence that the current SNF will meet the criteria for storage in the 8.4 Fuel Cladding Integrity Section 8.4, Rev. 1 Page 8.4-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS canister. In addition, based on the condition of the current spent fuel, the continued maintenance of the reactor coolant and spent fuel pool water chemistry requirements, and proper handling of the fuel, there is a high level of confidence that future SNF will meet the criteria for storage in the canister.

A cask-loading plan ensures that no damaged FA are loaded into the canister. If the structural integ-rity criterion is met, then approval for dry storage for a given assembly is issued. This qualification is documented and consequently referenced in the ISFSI operating procedures prior to loading spent FA into the canister.

The cask-loading plan provides a loading sequence based on the various characteristics of the FA being loaded. There are two main fuel-loading strategies used: uniform fuel loading and regionalized fuel loading.

Uniform fuel loading is used when the FA being loaded are all of similar burn-up rates, decay heat levels, and post-irradiation cooling times. In this case, the actual location of each assembly is less critical and assemblies can be placed at any location in the canister.

Regionalized fuel loading is used when high heat emitting FA are to be stored in the canister. This loading strategy allows these specific assemblies to be stored in locations towards the centre of the canister basket provided lower heat emitting FA are stored in the peripheral storage locations.

The following controls ensure that each FA is loaded into a known cell location within a qualified canister:

- A cask-loading plan is independently verified and approved.

- A fuel movement sequence is based upon the written loading plan. All fuel movements from any rack location are performed under controls that ensure strict, verbatim compliance with the fuel movement sequence.

Prior to placement of the canister lid, all FA and associated nonfuel hardware, if included, is either videotaped or visually documented by other means, and independently verified, by ID number, to match the fuel movement sequence.

Finally, a third independent verification is performed to ensure that the fuel in the canister is placed in accordance with the original cask-loading plan.

Based on the qualification process of the spent fuel and the administrative controls used to ensure that each FA is loaded into the correct location within a canister, incorrect loading of a canister is not considered a credible event.

8.4 Fuel Cladding Integrity Section 8.4, Rev. 1 Page 8.4-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 8.4.2 Cladding Mechanical Properties As described in section 8.4.1, the FA can be assumed undamaged at the time of loading into the canister of the CASTOR geo69 storage cask.

For at least one of the fuel types that shall be loaded into the canister of the CASTOR geo69 cask, the burn-up exceeds 45 GWd/MgHM (see Section 2.1 ). For all fuel types the intended dry storage time is beyond 20 years. Hence according to NUREG-2224 [1] age-related uncertainties connected with the extended dry storage of HBU undamaged SNF are to be considered in the safety analyses.

The chosen approach in this SAR is to supplement the design basis with safety analyses that demon-strate the DSS can still meet the pertinent regulatory requirements by assuming hypothetical recon-figuration of the undamaged HBU fuel contents into justified geometric forms. This approach demon-strates that after 20 years of dry storage, even if reconfigured, fuel can still meet the 10 CFR Part 72 requirements for thermal, confinement, criticality safety, and shielding during normal, off-normal, and accident conditions.

Following NUREG-2224 [1], the impact of cladding failures of Category 1 with breached rods (Sce-nario 1 (a) according to [1]) and with damaged rods (Scenario 1(b) according to [1])

on the fuel cladding and package component temperatures is evaluated in Chapter 4, is considered by the external dose and dose rate evaluation in Chapter 5 and on the canister pressure is evaluated in Chapter 7.

Rupture of 1 percent, 10 percent, and 100 percent of the fuel rods is assum~d for normal, off-normal, and accident conditions of storage, respectively.

The storage cask is designed to exclude the water leakage into the canister cavity under normal, off-normal and accident conditions of storage. Due to very low reactivity of dry fuel, the behaviour of the spent fuel as a result of accident conditions during dry storage beyond 20 years does not need to be explicitly evaluated and is bounded by the reactivity of the fully flooded cask with pure unborated water, as assumed in the bounding criticality safety model for normal conditions of storage (see Section 6.4).

No cladding mechanical properties and structural evaluation of the fuel rods under design-bases drop accident scenarios are required, once 100 percent rupture of the fuel rods and the resulting fuel reconfiguration for accident conditions of storage are evaluated as described in the safety analyses above.

8.4 Fuel Cladding Integrity Section 8.4, Rev. 1 Page 8.4-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 8.4.3 Cover Gas After exposure to oxidizing atmosphere, the fuel pellets may oxidize, expand and apply stress on the cladding. This potential oxidation is prevented by the measures/parameters applied during the drying procedure and the backfilling of the cavity with inert helium gas. This way it is ensured that the cladding will be protected against splitting from fuel pellet oxidation.

The canister is dried according to a specified drying procedure in accordance with the recommen-dations of R. W. Knoll and E. R. Gilbert [2]. Thereby, as much water as practicable is removed from the cavity and the pressure is evacuated to be less than or equal to 4.0 x 10-4 MPa. After evacuation, adequate moisture removal is verified by maintaining a constant pressure over an appropriate period without vacuum pump operation (or the vacuum pump is running but it is isolated from the cask with its suction vented to atmosphere). Moisture is removed to levels below 0.43 mole H2O.

Potential icing of the evacuation system line during evacuation is considered and excluded through adequate measures and use of suitable devices. This way possible ice blockage of the canister evacuation path are prevented.

The canister cavity is then backfilled with helium as inert gas for applicable pressure and leak testing.

The applied cover gas fulfils a defined quality specification that ensures a known maximum percent-age of impurities and is additionally verified by sampling. This way, the source of potentially oxidizing impurity gases and vapours are minimized and contaminants are adequately removed from the can-ister. The maximum quantity of oxidizing gasses (e.g., oxygen, carbon dioxide, and carbon monox-ide) are thereby limited to 1 mole per cask. This 1 mole limit reduces the amount of oxidants to below levels where cladding degradation is expected. Afterwards the inert gas atmosphere is maintained and no ingress from oxidizing atmosphere is implied.

In case the DSS confinement cavity is opened to an oxidizing atmosphere (as may occur in conjunc-tion with remedial welding, seal repairs), the process of evacuation and re-pressurisation is repeated.

List of References

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

[2] Evaluation of cover gas impurities and their effects on the dry storage of LWR spent fuel R. W. Knoll; E. R. Gilbert (November 1987)

Prepared for the U.S. Department of Energy under Contract DE-AC06-76RL0 1830 8.4 Fuel Cladding Integrity Section 8.4, Rev. 1 Page 8.4-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 8.5 Appendix Prepared Reviewed 8.5 Appendix Section 8.5, Rev. 1 Page 8.5-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Appendix 8-1: 1014-TR-00065 Rev. 2 Material Data Storage Cask CASTOR geo69 Appendix 8-2: 1014-TR-00011 Rev. 2 Material Qualification Appendix 8-3: 1014-TR-00012 Rev. 1 Material Qualification Appendix 8-4: 1014-TR-00017 Rev. 1 Material Qualification Metal Gaskets Appendix 8-5: 1014-TR-00048 Rev. 0 Material Qualification Ductile Cast Iron 8.5 Appendix Section 8.5, Rev. 1 Page 8.5-2

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev. 1 9 Operating Procedures 9.0 Overview Prepared Reviewed 9.0 Overview Section 9.0, Rev. 1 Page 9.0-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 This chapter includes a description of the designated operating procedures with the CASTOR geo69 DSS and CLU, comprising sufficient details regarding loading and unloading of the storage cask under usage of the CLU as well as preparation of the loaded storage cask and establishing the DSS configuration for long-term interim dry storage. The operations shall ensure the perfor-mance of the DSS and that it is operated in a manner consistent with the conditions assumed in the safety evaluation chapters of this SAR.

All operations shall be performed according to detailed written and approved procedures which shall comply with the content of this document, the applicable codes and standards and the CoC.

The preparation of these procedures, which are site-specific, is in the responsibility of the user.

The results from tests to be performed within the scope of the operations (e.g. leakage tests) shall be documented and become part of the quality documentation of the DSS.

The operational procedures have to be consistent with maintaining occupational radiation expo-sures as low as reasonably achievable (ALARA) as required by 10 CFR 20.1101 (b) [1 ]. Occupa-tional doses for operating procedures are estimated in Chapter 11.

In general, even if not explicitly mentioned, all components shall be subject to visual inspections prior to handling to ensure they are in proper condition.

List of References

[1] Title 10 CFR Part 20 Standards for Protection Against Radiation U.S. Nuclear Regulatory Commission 9.0 Overview Section 9.0, Rev. 1 Page 9.0-2

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev. 1 9.1 Procedure for Loading the Storage Cask Prepared Reviewed 9.1 Procedure for Loading the Storage Cask Section 9.1, Rev. 1 Page 9.1-1

• Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 The CASTOR geo69 is designated as transport and storage cask. The applicable loading-related preparations, tests and inspections of the storage cask described in the following comply with the storage regulations 10 CFR 72 and the CoC. The description include inspections made before loading to determine that the storage cask is in proper condition and radiation and surface contam-ination levels are within allowable regulatory limits. Casks deployed at a storage facility must be confirmed to meet all conditions of the 10 CFR 71 CoC prior to transport on public routes. Instanc-es in which the conditions of approval in the CoC were not observed in making a shipment shall be reported to the NRC.

The storage cask is loaded and closed in accordance with detailed written and approved proce-dures, including procedures for the preservation of screws with lubricant where necessary and tightening methods for the installation of all lids. Each screw is installed with either a nominal tight-ening torque or a nominal preload. Table 9.1-1 and Table 9.1-2 summarize the operations for preparation for loading and loading of contents, respectively.

9.1.1 Preparation for Loading It is assumed that the transfer cask is already in the reactor hall and that the transfer lock and fur-ther equipment necessary for the transhipment of the canister from the transfer cask into the CASTOR geo69 storage cask is available. Further, it is assured that all components of the CLU are in proper condition and ready for usage.

Table 9.1-1 Operations for preparation of loading

. Step Description Requirement 1 Accej;!tance of the emj;!t)l j;!ackaging or storage cask.

Delivery of the transport unit including the documents accompa-1.1 nvinJ:i transport.

1.2 Check of the validity of the Coe for the cask. Cask logbook Visual check of the packaging or storage cask for proper condi-1.3 tion, deformation, wear and corrosion.

Removal of impact limiters and if installed 1.4 on cask when accepted.

If the cask has previously been used with SNF:

1.5 49 CFR 173.443 [1]

Contamination control, decontamination in case of anv issues.

If the cask has previously been used with SNF:

1.6 10 CFR 71.47 [2]

External dose rate measurement.

1.7 Visual inspection of the empty storage cask 1.7.1 Visual inspection of the. outer surface coating of the cask.

Visual inspection of cask lid, blind flange and protection cap in installed 1.7.2 condition, as far as accessible.

9.1 Procedure for Loading the Storage Cask Section 9.1, Rev. 1 Page 9.1-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Step Description Requirement Visual inspection of the trunnions in installed condition, as far as ac-1.7.3 cessible.

1 Visual inspection of tilting studs and wear protection in installed condi-1.7.4 tion, as far as accessible.

Visual inspection of the preservation on wear protection and of the 1.7.5 hexagon screws.

1.7.6 3 Leak-tightness test of the wear protection.

Visual inspection of closure plate, hexagon head screws, sealing 1.7.7 2 screws with valve and seal plug in installed condition, as far as acces-sible.

Visual inspection of the preservation of the closure plate, hexagon 1.7.8 head screws and seal pluq.

3 Leak-tightness test of the closure plate and the sealing screws with 1.7.9 valve.

1.8 Transfer of the cask to handling position in the truck lock.

Load attachment on the trunnions, tilting of the cask into the vertical ANSI N14.6 1.8.1 orientation. NUREG-0612 Transfer to the handling position in the truck lock and closing of the 1.8.2 service platform around the cask.

~ Prei;!aration of the emi;!t~ cask for accei;!ting the canister.

2.1 Removal of the cask lid.

Visual inspection of the six threaded holes for the load attachment on 2.1.1 top of the cask lid.

Remove hexagonal screws and hexagon head screws for sealing:

- Removal of the three hexagon screws at marked positions (see numbering on the cask lid),

- Visual inspection of the respective threaded holes in the cask body, as far as accessible, 2.1.2 - Visual inspection of threads of the three guide bolts,

- Installation of the three guide bolts in the corresponding threaded holes in the cask body,

- Installation of the lifting pintle on the cask lid,

- Removal of the remaining hexagonal screws and hexagon head screws for sealing.

Removal of the cask lid.

2.1.3 The liftinq pintle remains installed on the cask lid.

2.1.4 Removal of the three guide bolts.

Visual inspection after removing the cask lid:

- Cask lid as a whole,

- Sealing groove for metal gasket in the cask lid after removal of the 2.1.5 metal gasket,

- Sealing surface for metal gasket on the cask body,

- Hexagonal screws and hexagon head screws for sealing,

- Threaded holes for screws in cask body.

Visual inspection of sealing surface protection of cask lid fit for impuri-2.1.6 ties and damaqe. Installation in lid fit.

2.2 Removal of blind flange from cask lid.

2.2.1 Removal of the cap screws.

2.2.2 Removal of blind flange via the LAP located in the centre.

9.1 Procedure for Loading the Storage Cask Section 9.1, Rev. 1 Page 9.1-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Step Description Requirement Visual inspection after removing the blind flange:

- Blind flange as a whole,

- Sealing groove for metal gasket in the blind flange after removal of 2.2.3 the metal gasket,

- Sealing surface for metal gasket on the cask lid,

- Cap screws,

- Threaded holes for ca screws in the cask lid.

2.3 Removal of protection cap from cask lid.

2.3.1 Removal of the cap screws.

2.3.2 Removal of protection cap via the two LAPs.

Visual inspection after removing the protection cap:

- Protection cap as a whole, .

Sealing groove for metal gasket in the protection cap after removal 2.3.3 of the metal gasket, Sealing surface for metal gasket on the cask lid, Cap screws, Threaded holes for ca screws in the cask lid.

2.4 Disassembly of the retention ring.

2.4.1 2.4.2 2.5 Installation of the lifting pintle on the canister lid.

2.6 Positioning of the transfer lock on top of the cask.

Canister transfer into transfer cask.

3.1 Positioning of transfer cask on top of the CASTOR geo69 cask.

Load attachment on the trunnions of the transfer cask and vertical ANSI N14.6 3.1.1 crane transfer to the handlin osition in the truck lock. NUREG-0612 Positioning of the transfer cask on the transfer lock on top of the CASTOR geo69. Two centre bolts provide adjustment.

3.1.2 The trunnions of the transfer cask remain attached to the traverse dur-in the entire transhi ment.

3.2 Transfer of the canister into the transfer cask.

3.2.1 Opening of the bottom lid of the transfer cask via the transfer lock.

ANSI N14.6 3.2.2 Attachment of the lifting pintle on the canister lid.

NUREG-0612 3.2.3 Lifting of the canister from the CASTOR geo69 into the transfer cask.

3.2.4 Closure of the bottom lid of the transfer cask via the transfer lock.

Settling of the canister on the bottom lid of the transfer cask and strike 3.2.5 off from the crane.

Transfer of the transfer cask with canister to the service position in the 3.2.6 reactor hall.

Closing of the platform around the transfer cask and disassembly of 3.2.7 the crane traverse.

9.1 Procedure for Loading the Storage Cask Section 9.1, Rev. 1 Page 9.1-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Step Description Requirement Preparation of canister and transfer cask for loading.

4.1 Preparation ~f the canister in the transfer cask.

ANSI N14.6 4.1.1 NUREG-0612 4.1.2 4.1.3 4.1.4 4.1.5 4.1.6 4.1.7 4.1.8 4.1.9 4.2 Preparation of the transfer cask.

4.2.1 4.2.2 4.2.3 4.2.4 ANSI N14.6 4.2.5 NUREG-0612 4.2.6 4.2.7 Removal of the sealing compound and removal of the hexagon screws not required 2 Removal of the sealing compound not required 3 A leak-tightness test is not required if the result of visual inspection of the preservation of the wear protection in step 1. 7 .5 and of closure plate in step 1. 7. 7 demonstrates condition according to specifi-cation.

4 In case the transfer cask lid is mounted.

9.1 Procedure for Loading the Storage Cask Section 9.1, Rev. 1 Page 9.1-5

Non-Proprietary Version 1014-SR-00002 ProprietanJ Information withheld per 10CFR 2.390 Rev. 1 9.1.2 Loading of Contents This chapter describes the loading process of the canister after its transfer inside the transfer cask to the underwater FA loading position on ground the spent fuel pool. In addition, the process of dispatching canister and storage cask including drying etc. is covered in detail. They are both loaded and closed in accordance with detailed written and approved procedures, including proce-dures for the preservation of screws with lubricant where necessary and tightening methods for the installation of all lids. Each screw is installed with either a nominal tightening torque or a nominal preload.

Both, canister as well as storage cask cavities are vacuum dried by adjusting and maintaining a vacuum pressure of After a sufficient drying period, the interior is disconnect-ed from the vacuum pump.

After successful completion of drying, the cavities are filled with inert gas.

In accordance with the thermal evaluation in Chapter 4 some constraints apply to the loading oper-ations to ensure compliance with temperature limits of both, cask components and fuel rod clad-ding and to prevent the water inside the canister from boiling:

Table 9.1-2 Operations for loading of contents Step Description Requirement

.1 Loading of the canister in the FA storage ~ool.

1.1 Loading of fuel assemblies.

Only the inventory defined in the valid CoC is permitted to be loaded.

An approved corresponding loading plan in compliance with the Coe is Coe,  :

1.1.1 required before loading to prove conformity between loading plan and Loading plan storage licence.

9.1 Procedure for Loading the Storage Cask Section 9.1, Rev. 1 Page 9.1-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Step Description Requirement Placing of fuel assemblies into the specified loading positions in com-pliance with the specified loading plan.

1.1.2 Before and after the loading of each individual fuel assembly, compari- Loading plan son of the fuel assembly number and the planned/implemented loading osition. To be recorded in the loadin Ian.

Final loading inspection of the canister by comparing the loading plan 1.1.3 with the actual loading configuration, following the "four-eye principle". Loading plan To be recorded in the loadin Ian.

1.2 Removal of transfer cask with canister from the FA storage pool.

Visual inspection of a new metal gasket and the sealing ring in the 1.2.1 sealing groove of the canister lid. Assembly of the metal gasket. Cask logbook Recordin of the metal asket identification.

Underwater removing of sealing surface protection from the lid fit in the 1.2.2 canister bod .

Underwater positioning of canister lid. Visual check for proper installa-1.2.3 tion.

• 1.2.4 ANSI N14.6 1.2.5 NUREG-0612 1.2.6 1.2.7 1.2.8 1.2.9 Lifting of the transfer cask (incl. canister) above the water surface of 1.2.10 the FA storage pool. Decontamination of the outer surface of the trans-fer cask with deionized water.

Placement of transfer cask in the service platform in the reactor hall 1.2.11 and closure of the platform. Removal of the traverse from the transfer cask trunnions.

Dewatering and drying of annulus between canister and transfer cask 1.2.12 cavit.

Preparation of canister and transfer cask before loading of canis-

~ ter into CASTOR eo69 cask.

Installation of the additional temporary neutron and gamma 2.1 shieldin .

2.2 Work and tests on the canister lid.

2.2.1 2.2.2 2.2.3 Check for proper installation of the canister lid 2.2.4 Disassembly of the lifting pintle from the canister lid.

Installation of an additional shielding plate (part of multi-equipment) on 2.2.5 the canister lid.

Vacuum drying (at vacuum pressure pv) of canister cavity and 2.2.6 verification of the 9.1 Procedure for Loading the Storage Cask Section 9.1, Rev. 1 Page 9.1-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 (@)GNS Step Description Requirement Setting of the blind plug and mounting of the quick connect incl. bond-2.2.7 ed seal.

2.2.8 Evacuation and helium filling of the canister via quick connect.

2.2.9 Visual inspection of a new Metal gasket.

Installation of new metal gasket, tightening plug and pressure nut and 2.2.10 tightening with nominal torque. Cask logbook Recording of the metal gasket identification.

2.2.11 Leakage test of the metal gaskets of the canister lid system. ANSI N14.5 [3]

2.3 Disassembly of the temporary additional shielding.

2.4 Installation of the lifting pintle on the canister lid.

3 Loading of the CASTOR geo69 cask.

3.1 Transfer of the canister into CASTOR geo69 cask.

Load attachment on the trunnions of the transfer cask and vertical ANSI N14.6 3.1.1 crane transfer from the service platform in the reactor hall to the NUREG-0612 handling position in the truck lock.

Positioning of the transfer cask on the transfer lock on top of the CASTOR geo69 storage cask.

3.1.2 Two center bolts provide adjustment.

The trunnions of the transfer cask remain attached to the traverse dur-ing the entire transhipment.

Load attachment to the lifting pintle on the canister lid and slight lifting ANSI N14.6 3.1.3 of the canister. NUREG-0612 3.1.4 Opening of the bottom lid of the transfer cask via the transfer lock.

3.1.5 Lowering of the canister into the CASTOR geo69 storage cask.

3.1.6 Closure of the bottom lid of the transfer cask via the transfer lock.

3.1.7 Transfer of the transfer cask to the service platform in the reactor hall.

Disassembly of the transfer lock. Installation of the temporary addition-3.1.8 al shielding and disassembly of the lifting pintle from the canister lid.

3.2 Assembly of the retention ring.

3.2.1 3.3 1 Closure of the storage cask.

3.3.1 Visual inspections of pressure switch and a new metal gasket.

Installation of the pressure switch without metal gasket and cap screws 3.3.2 in the cask lid. Tighteninq of the cap screws with nominal torque.

Check for proper installation of the pressure switch and removal of the 3.3.3 pressure switch.

Installation of the pressure switch with the new metal gasket and cap screws in the cask lid. Tightening of the cap screws with nominal 3.3.4 Cask logbook torque.

Recording of the metal gasket identification.

3.3.5 Check for proper installation of the pressure switch.

3.3.6 Functional test of the pressure switch.

Removal of the sealing surface protection from the cask lid fit and in-3.3.7 stallation of the guide bolts.

9.1 Procedure for Loading the Storage Cask Section 9.1, Rev. 1 Page 9.1-8

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Step Description Requirement Visual inspection and installation of a new metal gasket with clips in the 3.3.8 cask lid. Cask logbook Recordina of the metal aasket identification.

Lifting of the cask lid via the installed lifting pintle. Visual check of the 3.3.9 proper installation and condition of the moderator plate, as far as ac-cessible. Placement of the cask lid on the cask.

Installation of the hexagonal screws and removal of the guide bolts.

3.3.10 Installation of the three hexagon head screws for sealing. Tightening with nominal torque.

3.3.11 Check for proper installation of the cask lid.

Vacuum drying (at vacuum pressure pv) of cask cavity and 3.3.12 verification of the

~

Helium backfilling of the storage cask via quick connect (Item 60) and 3.3.13 adjustment of the internal pressure.

3.3.14 Visual inspection of a new metal gasket.

Installation of protection cap with metal gasket and cap screws in the 3.3.15 cask lid. Tightening of the cap screws with nominal torque. Cask logbook Recording of the metal gasket identification.

3.3.16 Check for proper installation of the protection cap.

3.3.17 Leakage test of the metal gaskets of the cask lid system. ANSI N14.5 3.3.18 Disassembly of the temporary additional shielding.

~ Pre~aration of the storage cask for on-site transfer Transfer of the storage cask out of the truck lock, tilting it into 4.1 horizontal position and positioning on the transfer vehicle.

4.2 Contamination control, decontamination in case of any issues. 49 CFR 173.443 [1]

4.3 External dose rate measurement. 10 CFR 71.47 [2]

~ On-ite transfer of the Storage cask to the storage facilitv.

The procedure for closing the storage cask includes the installation of a pressure switch at the former position of the blind flange for the direct storage on the NPP site without transport on public routes.

List of References

[1] Title 49 CFR Part 173 Shippers - General Requirements for Shipments and Packagings U.S. Nuclear Regulatory Commission

[2] Title 10 CFR Part 71 Packaging and Transportation of Radioactive Material U.S. Nuclear Regulatory Commission

[3] ANSI N14.5-2014 American National Standard for Radioactive Materials -

Leakage Tests on Packages for Shipment 9.1 Procedure for Loading the Storage Cask Section 9.1, Rev. 1 Page 9.1-9

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev. 1 9.2 Procedure for Unloading the Storage Cask Prepared Reviewed 9.2 Procedure for Unloading the Storage Cask Section 9.2, Rev. 1 Page 9.2-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The removal of contents from the canister in the FA storage pool may be performed via CLU Sys-tem or directly from the CASTOR geo69 DPC, depending on the crane capacity in the nuclear facility. Both cases are described in the following subsections. Mandatory for the complete duration of unloading of a cask is that the validity of the Coe is not exceeded.

Underwater FA unloading requires prior re-cooling of the canister and the loaded contents accord-ing to written and approved procedures to protect the FAs from damage due to thermal shock. The owner is responsible for this procedure, which is not described in detail in this subsection.

9.2.1 Direct Removal of Contents Table 9.2-1 summarizes the operations during direct removal of contents from the CASTOR I geo69 storage cask. The initial condition is that the impact limiters, if present, are already disman-tled, the storage cask has been accepted and the two sealing screws with valve installed in the closure plate have been replaced by two simple sealing screws sealed by an O-ring. The loaded cask is located in the service platform in the nuclear facility.

Table 9.2-1 Required steps for the unloading of contents directly from the DPC Step Description Requirement

.'.!. O~ening of the storage cask .

1.1 Removal of the cask lid.

Visual inspection of the threaded holes for the load attachment on top 1.1.1 of the cask lid.

Remove hexagonal screws and hexagon head screws for sealing:

- Removal of the three hexagon screws at marked positions (see numbering on the cask lid),

- Visual inspection of the threaded holes in the cask body, as far as accessible, 1.1.2

- Visual inspection of thread on the three guide bolts,

- Installation of the three guide bolts in the corresponding threaded holes in the cask body,

- Installation of the lifting pintle on the cask lid,

- Removal of the remaininQ hexaQon screws.

Removal of the cask lid. The lifting pintle remains installed on the cask 1.1.3 lid. Removal of the Quide bolts.

Visual inspection after removing the cask lid:

- Cask lid as a whole, 1.1.4

- Sealing groove for metal seal in the cask lid,

- Sealing surface for metal seal on the cask body,

- Hexagonal screws and hexagon head screws for sealing,

- Threaded holes for hexaQonal screws in cask body.

Visual inspection of sealing surface protection of cask lid fit for impuri-1.1.5 ties and damage. Installation in lid fit.

9.2 Procedure for Unloading the Storage Cask Section 9.2, Rev. 1 Page 9.2-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Step Description Requirement 1.2 Removal of blind flange from cask lid.

1.2.1 Removal of the cap screws.

1.2.2 Removal of blind flange via the LAP located in the centre.

Visual inspection after removing the blind flange:

- Blind flange as a whole,

- Sealing groove for metal gasket in the blind flange after removal of 1.2.3 the metal gasket,

- Sealing surface for metal gasket on the cask lid

- Cap screws, Threaded holes for ca screws in the cask lid.

1.3 Removal of protection cap from cask lid.

1.3.1 Removal of the cap screws.

I

- 1.3.2 1.3.3 Removal of protection cap via the two LAPs.

Visual inspection after removing the protection cap:

- Protection cap as a whole,

- Sealing groove for metal gasket in the protection cap,

- Sealing surface for metal gasket on the cask lid,

- Cap screws,

- Threaded holes for ca screws in the cask lid.

1.4 Disassembly of the retention ring.

1.4.1 1.4.2

~ Unloading of the canister.

2.1 Preparation of the canister for re-cooling.

I 2.1.1 Installation of the lifting pintle on the canister lid.

I el 2.1.2 2.1.3 Removal of tightening plug and pressure nut from the canister lid.

Evacuation of canister cavity via quick connect.

Flushing of the canister cavity with helium at ambient pressure via 2.1.4 uick connect for coolin the FAs.

2.1.5 2.1.6 2.2 Preparations before unloading.

Load attachment on the trunnions of the cask. Transfer of the cask to ANSI N14.6 2.2.1 the FA stora e col. NUREG-0612 2.2.2 Partial lowering of the cask into the FA storage pool.

2.2.3 Filling of the canister cavity with pool water.

2.2.4 Removal of the dewatering lance.

2.2.6 Lowering of cask and canister into the FA storage pool.

9.2 Procedure for Unloading the Storage Cask Section 9.2, Rev. 1 Page 9.2-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Step Description Requirement 2.2.5 Underwater removal of canister lid via the installed lifting pintle.

2.3 Underwater unloading of the fuel assemblies.

9.2.2 Removal of Conten(s via CLU Table 9.2-2 describes the steps to be. performed at a minimum for an unloading of a cask via the CLU. The initial condition is that the loaded storage cask has been accepted, the loaded cask has been opened, the canister has been transhipped from the storage cask into the transfer cask (ac-cording to steps 1 - 3 in Table 9.1-1) and the loaded transfer cask is located in the service platform in the nuclear facility. The inner and outer water chambers of the transfer cask must be flooded with deionized water prior to canister transhipment.

Table 9.2-2 Required steps for the unloading of contents via CLU Step Description Requirement 1 Preparation of canister and transfer cask for unloading.

1.1 Preparation of the transfer cask.

Inflate the annulus inflatable seal between bottom lid and bottom ring 1.1.1 of the transfer cask.

Connection of the annulus between canister and transfer cask cavity 1.1.2 wall to the deionized water service and flooding.

The connection remains durin the entire underwater loadin of FAs.

1.2 Preparation of the canister for re-cooling.

el 1.2.1 1.2.2 Removal of tightening plug and pressure nut from the canister lid.

Evacuation of canister cavity via quick connect.

Flushing of the canister cavity with helium at ambient pressure via 1.2.3 uick connect for coolin .the FAs.

1.2.4 1.2.5 i Unloading of the canister.

2.1 Preparations before unloading.

Load attachment on the trunnions of the transfer cask. Transfer of the ANSI N14.6 2.1.1 loaded transfer cask to the FA stora e ool. NUREG-0612 2.1.2 Partial lowering of the transfer cask into the FA storage pool.

Start continuous flushing of annulus between canister and transfer 2.1.3 cask cavit .

2.1.4 Filling of the canister cavity with pool water.

9.2 Procedure for Unloading the Storage Cask Section 9.2, Rev. 1 Page 9.2-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Step Description Requirement 2.1.5 Removal of the dewatering lance.

2.1.6 Lowering of transfer cask and canister into the FA storage pool.

2.1.7 Disassembly of the crane traverse.

2.1.8 Removal of canister lid under water via the installed lid hanger.

2.2 Unloading of the fuel assemblies under water.

9.2 Procedure for Unloading the Storage Cask Section 9.2, Rev. 1 Page 9.2-5

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev. 1 @G S 9.3 Preparation of the Storage Cask for Long-Term Dry Storage Prepared Reviewed 9.3 Preparation of the Storage Cask for Long-Term Dry Storage Section 9.3, Rev. 1 Page 9.3-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 This section describes the preparation of the storage cask for long-term dry storage and the set-up of the CASTOR geo69 DSS at the storage facility. The operations described in Table 9.3-1shall be performed subsequent to the transport on public routes. The operations described in Table 9.3-2 shall be performed on the storage cask after on-site transfer from the NPP.

9.3.1 Preparation for Long-Term Dry Storage after Transport on Public Routes The operations performed after receiving the package from carrier are in accordance with the re-quirements listed in 10 CFR 20.1906 [2], "Procedures for Receiving and Opening Packages".

The designated storage position of the DSS must be prepared prior to performance of the opera-tions specified in Table 9.3-1. Equipment and further components (e.g. pressure switch) necessary for establishing the CASTOR geo69 DSS must be available.

Table 9.3-1 Operations for preparation for long-term dry storage after transport on public routes Step Description Requirement 1 Surve~s and insj;!ections after recei!;!t of the j;!ackage.

Receipt of the transport unit including the documents accompa-1.1 nying the transport.

Visual check of the package for proper condition, deformation, wear and corrosion.

1.2 Check of type plates, labels and tamper-indicating devices on the packaae.

1.3 Removal of impact limiters 1.4 1 Contamination measurement on the cask. 49 CFR 173.443 I

el 1.5 1 1.6 Dose rate measurement on the cask.

Visual inspection of the cask.

10 CFR 71.47 I

1.6.1 Visual inspection of the outer surface coating of the cask.

Visual inspection of cask lid, blind flange and protection cap in installed 1.6.2 condition, as far as accessible.

Visual inspection of the trunnions in installed condition, as far as ac-1.6.3 cessible.

Visual inspection of tilting studs and wear protection in installed condi-1.6.4 tion, as far as accessible.

Visual inspection of the preservation on wear protection and of the 1.6.5 hexaQon screws.

1.6.6 2 Leak-tightness test of the wear protection.

Visual inspection of closure plate, hexagon head screws, sealing 1.6.7 screws with valve and seal plug in installed condition, as far as acces-sible.

Visual inspection of the preservation of the closure plate, hexagon 1.6.8 head screws and seal pluQ.

9.3 Preparation of the Storage Cask for Long-Term Dry Storage Section 9.3, Rev. 1 Page 9.3-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Step Description Requirement Leak-tightness test of the closure plate and the sealing screws with 1.6.9 valve.

Transfer of the storage cask to the service station of the storage

~ facilitv.

ANSI N14.6 2.1 Load attachment on the trunnions of the storage cask.

NUREG-0612 Tilting of the storage cask by application of the tilting studs and the ANSI N14.6 2.2 correspondinq tiltinq suooort. NUREG-0612 Vertical (crane) transfer (if necessary under usage of a supporting 2.3 vehicle) to the service station of the storage facility and positioning in the service platform.

Closing of the service platform around the storage cask and installation 2.4 of the temporary additional shielding.

~ Installation of the (!ressure switch in the cask lid.

3.1 Removal of cap screws, protection cap and metal gasket.

Controlled pressure normalisation in the cask cavity using correspond-3.2 inq equipment via quick connect.

3.3 Removal of cap screws, blind flange and metal gasket .

Visual inspections of

• Pressure switch, 3.4

• Cap screws and application of lubricant,

• New metal gasket,

• Sealing groove in pressure switch and surface in cask lid,

• Threaded holes Installation of the pressure switch without metal gasket and cap screws 3.5 in the cask lid. Tightening of the cap screws with nominal torque.

Check for proper installation of the pressure switch and removal of the 3.6 pressure switch.

Installation of the pressure switch with the new metal gasket and cap screws in the cask lid. Tightening of the cap screws with nominal 3.7 Cask logbook torque.

Recording of the metal gasket identification.

3.8 Check for proper installation of the pressure switch.

3.9 Functional test of the pressure switch.

3.10 3.11 Leakage test of the metal gasket in the pressure switch. ANSI N14.5 Visual inspections of

• Protection cap, 3.12

• Cap screws and application of lubricant,

• New metal gasket,

• Sealing groove in protection cap and surface in cask lid,

• Threaded holes Installation of the protection cap with the new metal gasket and cap screws in the cask lid. Tightening of the cap screws with nominal 3.13 Cask logbook torque.

Recording of the metal gasket identification.

3.14 Check for proper installation of the protection cap.

3.15 Leakage test of the metal gaskets in the protection cap. ANSI N14.5 9.3 Preparation of the Storage Cask for Long-Term Dry Storage Section 9.3, Rev. 1 Page 9.3-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @)GNS Step Description Requirement 3.16 Disassembly of the temporary additional shielding.

.1: Set-u~ of CASTOR geo69 DSS Control of the conservation status of the storage cask and implementa-4.1 tion of improvements, if necessary ANSI N14.6 4.2 Load attachment on the trunnions of the storage cask.

NUREG-0612 4.3 Removal of the service platform around the storage cask.

Vertical (crane) transfer of the storage cask to the designated storage 4.4 position in the storaQe facility. Disassembly of the crane traverse.

Visual inspection of the protection cover incl. attachments and load 4.5 attachment points.

Load attachment on the LAPs of the protection cover. Crane transfer of 4.6 the protection cover and installation onto the storaQe cask.

Installations of the cable conduit and feeding of the pressure switch 4.7 cables throuQh the cable conduit Connection of CASTOR geo69 DSS to the pressure monitoring sys-4.8 tern of the storage facility and check of functionality.

Long-term interim dr)l storage for the designated storage ~eriod

§ of the CASTOR aeo69 D55.

Contamination and dose rate measurement shall be performed as soon as practical after receipt of the package, but not later than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> after the package is received at the licensee's facility. If the package is received outside normal working hours, the monitoring shall be performed not later than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> from the beginning of the next working day.

2 A leak-tightness test is not required if the result of visual inspection of the preservation of the wear protection in step 1.6.5 and of closure plate in step 1.6.7 demonstrates condition according to specifi-cation.

9.3.2 Preparation for Long-Term Dry Storage after On-site Transfer from the NPP The procedure described in Table 9.3-2 are performed subsequent to the steps specified in Table 9.1-2.

Table 9.3-2: Operations for preparation for long-term dry storage after on-site transfer from the NPP Step Description Requirement 1 Surve)ls and ins~ections after recei~t of the storage cask.

1.1 Receipt of the storage cask including the accompanying documents.

1.2 Visual inspection of the cask.

1.2.1 Visual inspection of the outer surface coating of the cask.

Visual inspection of cask lid, blind flange and protection in installed condi-1.2.2 tion, as far as accessible.

Visual inspection of the trunnions in installed condition, as far as accessi-1.2.3 ble.

9.3 Preparation of the Storage Cask for Long-Term Dry Storage Section 9.3, Rev. 1 Page 9.3-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Step Description Requirement Visual inspection of tilting studs and wear protection in installed condition, 1.2.4 as far as accessible.

Visual inspection of the preservation on wear protection and of the hexa-1.2.5 qon screws.

Visual inspection of closure plate, hexagon head screws, sealing screws 1.2.6 with valve and seal plug in installed condition, as far as accessible.

Visual inspection of the preservation of the closure plate, hexagon head 1.2.7 screws and seal plug, as far as accessible.

i Set-LIQ of CASTOR geo69 DSS ANSI N14.6 2.1 Load attachment on the trunnions of the storage cask.

NUREG-0612 Vertical crane transfer of the storage cask to the designated storage posi-2.2 tion in the storage facility. Disassembly of the crane traverse.

Visual inspection of the protection cover incl. attachments and load at-2.3 tachment points.

Load attachment on the LAPs of the protection cover. Crane transfer of 2.4 the protection cover and installation onto the storage cask.

Installations of the cable conduit and feeding of the pressure switch cables 2.5 through the cable conduit Connection of CASTOR geo69 DSS to the pressure monitoring system 2.6 of the storage facility and check of functionality.

Long-term dn,: storage for the designated storage 12eriod of the

~ CASTOR geo69 D55.

List of References

[2] Title 10 CFR Part 20 Standards for Protection Against Radiation U.S. Nuclear Regulatory Commission 9.3 Preparation of the Storage Cask for Long-Term Dry Storage Section 9.3, Rev. 1 Page 9.3-5

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev. 0 GNS 9.4 Appendix Prepared Reviewed 9.4 Appendix Section 9.4, Rev. 0 Page 9.4-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 With intent no items.

9.4 Appendix Section 9.4, Rev. 0 Page 9.4-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 10 Acceptance Criteria and Maintenance Program 10.0 Overview Prepared Reviewed 10.0 Overview Section 10.0, Rev. 1 Page 10.0-1

Non-Proprietary Version.

1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS This chapter identifies programs to be conducted on the CASTOR geo69 DSS and the CLU to verify that the SSCs classified as important to safety have been fabricated, assembled, inspected, tested, accepted, and maintained in accordance with the requirements set forth in this SAR. The acceptance criteria and maintenance program *requirements specified in this chapter apply to each CASTOR geo69 DSS and CLU fabricated, assembled, inspected, tested, accepted and maintained for use under the scope of the CoC issued by the NRC in accordance with the requirements of 10 CFR 72.

The controls, inspections and tests set forth in this chapter in conjunction with the design require-ments described in previous chapters (incl. the parts lists and drawings presented in Section 1.5) ensure that the CASTOR geo69 DSS and the CLU, if applicable, will retain the stored radioactive materials; will maintain subcriticality control; will properly transfer the decay heat of the stored radi-oactive materials; and that radiation doses will meet regulatory requirements.

Identification and resolution of nonconformance shall be performed in accordance with the QAP as described in Chapter 14. Nonconformance reports shall become part of the. documentation of the DSS and the CLU.

10.0 Overview Section 10.0, Rev. 1 Page 10.0-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 {g)GNS 10.1 Acceptance Criteria Prepared Reviewed 10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Prior to the start of production of a CASTOR geo69 DSS or CLU, all relevant design criteria and production-related acceptance tests and criteria for the individual production and assembly steps must be available to the manufacturers commissioned with the production. The parts lists and draw-ings referenced in Section 1.5 define all necessary codes & standards as well as other applicable specifications which are implemented in the respective manufacturing documents. Components which are important to safety are fabricated in accordance with manufacturing parts lists and draw-ings (based on the design parts lists and drawings included in Section 1.5) and with approved fabri-cation and test plans which ensure a sufficient manufacturing sequence. Together with material and test specifications for the procurement of raw materials (plates, forgings, castings, etc.) and a fabri-cation specification for the assembly of parts to components and for the final assembly of the DSS and CLU these documents include all applicable manufacturing specifications. Material procurement of SSC which are important to safety shall require certification and documentation according to at least Section II of the BPVC as well as to Section II, Division 3 (DSS) and Division 1, Subsection NF

[1] (CLU), respectively, and 10CFR72, Subpart G. Material traceability must be ensured throughout production. Certificates and documentation become part of the final quality documentation of DSS and CLU, respectively. All manufacturing documents shall at a minimum comply with, but are not limited to, the specifications of this SAR.

The manufacturing process follows the requirements of GNS Quality Assurance Program (Chapter

14) to ensure that each component is produce in accordance with the applicable specifications. This includes the following examinations:

• In-process inspections during the fabrication,

• Receiving inspections for overall quality, functionality, dimensional compliance and certifica-tion requirements.

This section of the SAR provides a summary of the acceptance tests to be performed, demonstrating that each CASTOR geo69 DSS and CLU is fabricated, assembled, inspected, tested, and accepted for use in accordance with the applicable codes. and standards (see Section 2.0) and the design criteria of this SAR (especially the parts lists and the corresponding drawings included in Sec-tion 1.5). The tests shall ensure that the initial operation complies with the regulatory requirements.

All tests applicable to DSS and CLU shall be performed in accordance with written and approved procedures (10 CFR 21.162). The results shall be documented and become part of the final quality documentation of DSS and CLU, respectively.

10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 1 OCFR 2.390 Rev. 1 Test and measuring equipment and tools requiring calibration are subject to a monitoring process.

The respective valid calibration certificate shall be available at the time of usage and the identification number of the test and measuring equipment or tool used are documented.

Any occurring noncompliance with the design criteria (parts lists and drawings, etc.) leads either to further treatment of the part (to correct e.g. its dimensions) orto the manufacturing of a new exemplar of that particular part. Further treatment of DSS parts shall comply with Division 3 and further treat-ment of CLU parts shall comply with Section Ill, Division 1, Subsection NF of the BPVC [1]. The noncompliance and the further treatment or repair shall be documented and become part of the final quality documentation of DSS and CLU, respectively.

The storage cask must be conspicuously and durably marked according to 10 CFR 72.236(k) with a model number, a unique identification number and an empty weight. Therefore, two stainless steel nameplates providing the demanded information are permanently attached at eye level on the outer surface of the transport and storage cask.

10.1.1 Materials Testing All materials uses within the fabrication of SSC important to safety are procured according to manu-facturing parts lists and drawings, material and test specifications and the fabrication specification, which all shall at a minimum comply with, but are not limited to the requirements of this SAR. In particular, this specifies the codes and standards to be applied as well as the examinations and tests to be performed during fabrication and on the final product, including the respective acceptance criteria.

These are described below as an example for a material of particular importance namely DCI ac-cording to SA-874M [2] (and further specifications referenced therein) for the cask body. The follow-ing in-process examinations and test are foreseen:

10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @)GNS 10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 10.1-1: Rejection criteria for PT of sealing surfaces on castings In case of any nonconformities repairs of the casted cask body are not permitted.

Similar specifications apply to other materials used.

10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS All examinations and test shall be performed according to written and approved procedures and the results shall become part of the quality documentation of the particular part, component or the DSS.

10.1.2 Visual Inspection and Nondestructive Examination 10.1.2.1 General Prior to assembly, all items, parts and components of the CASTOR geo69 DSS and the CLU shall be subject to visual inspections regarding intactness, cleanness and non-corrosiveness.

The materials of construction shall be recipe inspected for visual and dimensional acceptability, tak-ing the manufacturing parts lists and the corresponding drawings as a basis. A check of the manu-facturing documentation (material certification and traceability) for completeness and factual correct-ness shall be included. The dimensions of DSS, CLU and all parts given in the manufacturing docu-mentation shall conform to the dimensions and tolerances specified in the manufacturing drawings.

A verification that each part consists only of the materials specified in the manufacturing part lists must also be included.

10.1.2.2 Weld Examinations Welding within the framework of the BPVC shall be performed using welders and weld procedures that are qualified in accordance with Section IX [12] of the BPVC and the respective applicable part of Section Ill, Division 3 or Subsections NF as applicable to the SSC. Procedures for other welds are in the choice of the manufacturer.

The examination of welds shall verify that the fabrication of DSS, CLU and components comply with the manufacturing drawings and referred to in the CoC. All weld examinations shall be performed in compliance with the appropriate Article of Section V of the BPVC [13] for the particular examination method. Weld seams exist on the following parts of the DSS and the CLU:

• 1014-DPL-36855 canister (welded canister body)

• 1014-DPL-13752 protection cover (weld seams between plate and ring)

• 1015-DPL-37509 transfer cask (especially inner liner, lead and water enclosures, which are welded to the head ring and bottom ring)

• 1015-DPL-38148 transfer lock (weld joint connections of the cylinder console to the frame)

The non-destructive examination procedures applicable for the DSS components depend on the category of the weld seams according to Division 3, WC-3251. Weld examination of containment welds of the canister body shall be performed in accordance with Division 3, WC-5200 with regard 10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information withhelc;i per 10CFR 2.390 Rev. 1 @GNS to the required examination method and acceptance criteria. Personnel performing non-destructive examinations shall be qualified in accordance with Division 3, WC-5520. For the weld seams in the protection cover, assessment group C according to ISO 5817 [14] applies, personnel shall be quali-fied according to ISO 9712 [15]. Section Ill, Division 1, NF-5210 [1] determines the required NDE methods of the welds in the CLU components, NF-5300 the acceptance criteria and NF-5500 the personal qualification.

Table 10.1-2 gives an overview on the required examinations of welds on the different parts of the DSS and the CLU. The required NDE procedures shall be performed in accordance with Section V

[13] of the BPVC, except for the protection cover.

Table 10.1-2: Required examination of welds in the DSS and the CLU Weld seam location Code I standard Welded joint category Examination method Division 3 Canister body A,B,C RT+ PT Subsection WC Protection cover ISO 5817 C VT+ PT Division 1 Primary member RT/PT Transfer cask Subsection NF Secondary member VT Division 1 Transfer lock Secondary member VT Subsection NF 10.1.2.2.1 Examination of Welds in the Canister Body Table 10.1-3 summarizes the required NOE of welds in the canister body. Location, types and size of each weld must be confirmed by the required examination method before commissioning. Defects in weld metal detected by the examinations shall be eliminated and repaired when necessary in accordance with Division 3, WC-4450 or the indication shall be reduced to an acceptable limit.

Table 10.1-3: Required Examination of Welds in the Canister Body Part Weld location t[mm] Weld type Required NDE Item 2 Item 2-3 Canister body Item 2 Item 2-4 Item 2 Item 2-6 Item 2 Item 2-6

- Single U butt weld RT+PT

• Item No. according to the respective parts list of storage cask and canister, referenced in Section 1.2 10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-8

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Radiographic Testing RT shall be performed in accordance with Section V, Article 2, except that fluorescent screens are not permitted for film radiography, the geometric unsharpness shall not exceed the limits of Sec-tion V, Article 2, T-274.2, and the image quality indicators of Division 3! Table WC-5111-1 shall be used in lieu of those shown in Section V, Article 2, T-276. Indications shown on the radiographs of welds and characterized as imperfections shall meet the acceptance criteria according to l;)ivision 3, WC-5320.

Penetration Testing PT shall be performed in accordance with Section V, Article 6. Imperfections producing indications that do not meet the acceptance criteria specified in Division 3, WC-5352 are unacceptable.

10.1.2.2.2 Examination of Welds in the CLU Table 10.1-4 summarizes the required NOE of each weld in the transfer cask and the transfer lock.

Defects in weld metal detected by the examinations shall be eliminated and repaired when necessary in accordance with Section Ill, Division 1, NF-4450 or the indication shall be reduced to an accepta-ble limit.

Radiographic Testing RT shall be in accordance with Section V, Article 2, except that the geometric unsharpness shall not exceed the limits of BPVC Section V, Article 2, T-274.2. Only the welded connections between head ring, liner and bottom ring in the transfer cask body require RT under consideration of the acceptance criteria specified in BPVC Section Ill, Division 1, Subsection NF-5320. Indications shown on the ra-diographs of the welds shall meet these acceptance criteria.

Visual Testing VT shall be performed in accordance with Section V, Article 9. When VT is performed on welds in the transfer cask or the transfer lock, the acceptance criteria according to BPVC Section Ill, Divi-sion 1, Subsection NF-5360 shall apply.

10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-9

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 10.1-4: Required Examination of Welds in the CLU Part Weld location

• t[mm] Weld type Required NDE Item 2 Item 2-3 Bevel groove weld RT Item 2 Item 2-4 Item 2 Item 2-6 V groove weld VT Item 2 Item 2-6 Fillet weld VT Item 2 Item 2-8 V groove weld VT Item 2 Item 2-8 Item 2 Item 2-10 Fillet weld VT Item 2 item 2-12 Bevel groove weld VT Item 2 Item 2-13 Fillet weld VT

- Transfer cask Item 2 Item 2-9 Item 2;-6 (2 half-shells)

Item 2 Item 2-7 Item 2 Item 2-18 Item 2 Item 2-18 V groove weld V'groove weld Fillet weld Fillet weld VT VT VT VT body Item 2-8 (2 half-shells) V groove weld VT Item 2 Item 2-11 Bevel groove weld VT Item 2 Item 2-17 Fillet weld VT Item 2 Item 2-17 Item 2-9 (2 half-shells) V groove weld VT Item 2 Item 2-10 V groove weld VT Item 2 Item 2-11 Bevel groove weld VT Item 2 Item 2-12 Bevel groove weld VT Item 2 Item 2-13 Fillet weld VT Item 2 Item 2-19 Fillet weld VT Item 2-1 O - Item 2-11 Fillet weld VT Item 2 Item 2-12 Bevel groove weld VT Item 30 Item 30-3 Fillet weld VT Transfer lock Item 30 Item 30-4 Fillet weld VT

• Item No. according to the respective parts list of storage cask and canister, referenced in Section 1:2 10.1.2.2.3 Examination of Welds in the Protection Cover For the non-destructive examination of the weld seams in the protection cover, the acceptance cri-teria of assessment group C according to ISO 5817 [14] apply. Each weld shall be tested via VT and PT. VT shall be performed according to ISO 17637 [16] or another applicable standard. PT shall be performed according to ISO 3452-1 [17] or another applicable standard. Defects detected by the 10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-10

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 examinations shall be eliminated and repaired when necessary or the indication shall be reduced to an acceptable limit.

10.1.3 Structural and Pressure Tests All test shall comply with the applicable cods and standards mentioned in the following respective subsections. If sufficient, instead of performing especially the pressure tests at each single cask of the CASTORgeo69 design type, a verification by calculation or, if necessary, by experimental test-ing for the design type shall be permitted.

10.1.3.1 Pressure Tests 10.1.3.1.1 DSS- Storage Cask and Canister Pressure tests during fabrication only apply to the components cask and canister of the DSS. For the pressure tests, the requirements of Division 3 apply. Hydrostatic tests according to WC-6200 shall be performed with a maximum internal overpressure of at least 1.25 times the design pressure.

The external design pressure of the canister is equal to the internal design pressure of the storage cask. Since this pressure exceeds th~ internal design pressure in the canister, the requirements of Division 3, WC-6610 apply. The canister shall be pressure tested with a test pressure equal to 1.25 times the external design pressure. The applied test pressures for the storage cask and canister are summarized in Section 3. 7 and 3.9. They cover the required minimum test pressures according to Division 3. The test pressure shall be maintained for at least 10 minutes. The canister test pres-sure shall not be exceeded by more than 6 % during the test. Examinations after the pressure test shall be in accordance with WC-6224. (Alternately, a pneumatic pressure test may be performed in accordance with WC~6300 using 1.2 times the design pressure.)

Deviating from the specifications of the BPVC, the pressure tests are carried out only once as a process qualification, i.e. on the initial samples of storage cask and canister. They are omitted during serial production. This is discussed and assessed in Section 2.0. If deemed necessary, a pressure test can of course be carried out in individual cases.

After the test period time of the hydrostatic tests is exceeded, the initiation of the examination for leakage starts. LT shall be performed as described in Subsection 10.1.4 with an internal pressure equal to 75 % of the hydrostatic test pressure. All joints, connections and regions of high stress are tested for leaks.

10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-11

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 10.1.3.1.2 CLU Since the CLU does not provide containment for the nuclear content, no pressure test in accordance with Division 3, WC-6200 or WC-6300 during fabrication is designated for the CLU.

10.1.3.2 Structural Tests Structural tests are performed on all LAP in terms of a static load test. The applied load depends on the maximum weight that will be attached to the LAP during handling of the DSS and CLU compo-nents, multiplied with a certain load factor. The load attachment points on the canister lid and the trunnions of storage cask and transfer cask are special lifting devices and thus tested with a load factor of three in order to fulfil the provisions of ANSI N14.6 [18]. All other lifting devices are tested with a load factor of 1.5. Furthermore, a hoist factor of 1.15 according to CMAA #70 applies to all lifting devices for slow crane operation. The loads for the static load tests are listed in Table 10.1-5.

All load tests shall be performed at room temperature and the static load shall be maintained for at least 30 minutes.

The static load tests of the trunnions and the tilting studs require the use of bearing shells as indi-cated in Figure 10.1-1. The requirements for the geometry of the bearing shells are specified in Subsection 10.1.3.2.1 and 10.1.3.2.2.

Table 10.1-5: Requirements for static load tests Lifting device (quantity)

Storage cask trunnions (2)

Load factor 3x1.15

-- Max. weight attached [Mg]

--Load[kN]

Transfer cask trunnions (2)

LAP on canister lid (12) 3x1.15 3x1.15 Tilting studs (2) 1.5 X 1.15 LAP on cask lid (6) 1.5 X 1.15 10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-12

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Section A-A Trunnion flange or Trunnion collar or wea, p\::f~n~*- - _oreaf,+l<ction colla, I' L .

I A VI I

I,,.,.._.,~---+-~--...,,.

Rt 1U I

• bearing !¥1i.eli . ' _j I

Ii ._ __s....;c"-'w"'--/-F--l ow 1 -

suspension dip or L -------i*

~I-frame Figure 10.1-1: Geometry of load application on the trunnions / tilting studs 10.1.3.2.1 Storage Cask Trunnions The static test load must be distributed equally on the two trunnions of the storage cask. The direction of load application must be perpendicular to the axis of the trunnions. Bearing shells made of bronze must be placed between lifting lug and the trunnion. The following requirements must be met:

After the load test, a visual inspection and MT or PT must be performed on the installed trunnions, as far as accessible, to verify no distortion or cracking has occurred. MT or PT shall be performed in accordance with Section V, Article 7 or Article 6, respectively, using the acceptance criteria specified in Division 3, WC-5342 orWC-5352. The trunnion deformation between the initial value prior to load-ing (null measurement) and after loading with the test load shall be determined via dial gauge (meas-uring accuracy+/- ***I). A permanent trunnion deformation exceeding the measuring inaccuracy of the dial gauge is unacceptable. Otherwise, the deformed trunnions shall be replaced by new ex-emplars and the load test shall be repeated.

The cap screws that are exposed to the highest loading during the static lead test shall be removed and visually inspected after the test. Thread testing of these screws and the corresponding threaded 10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-13

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS holes in the storage cask body shall be performed. Defective cap screws shall be replaced with new exemplars. The tightening torque of all cap screws shall be checked after the load test.

10.1.3.2.2 Storage Cask Tilting Studs The load test on the tilting studs of the storage cask has to be performed prior to the installation of the wear protection. The static test load must be distributed equally on the two tilting studs and must be applied perpendicular to the axis of the tilting studs. Bearing shells made of bronze must be used

• and the following requirements apply:

A visual inspection of the of the tilting studs as well as MT or PT of the whole shell surface of the tilting studs and in the area of the transition radius shall be performed after the test load is applied, to verify no distortion or cracking has occurred. The deformation of the tilting studs due to the load shall be determined via dial gauge (measuring accuracy +/- . A perm~nent deformation of the tilting studs exceeding the measuring inaccuracy of the dial gauge is unacceptable. In case of a non-permitted deformation, the cask body shall be replaced by a new exemplar.

10.1.3.2.3 Canister Lid The static load must be distributed equally on the threaded holes by using the lifting pintle for the canister lid. The load must be applied perpendicular to the cask surface. The test screws must be screwed into the lid by hand until the minimum required screw-in depth is reached. A visual inspection and a gauge test of the threaded holes and a check of the screw-in depth on engagement must be performed before and after the load test. No adverse effects on the trueness to gauge of the threaded holes are permissible. The canister lid must be replaced by a new exemplar if the threaded holes do not pass the static load test.

10.1.3.2.4 Cask Lid The static load must be distributed equally on the threaded holes by using the lifting pintle for the cask lid. The load must be applied perpendicular to the cask lid surface. The test screws must be screwed into the lid by hand until the minimum required screw-in depth is reached. A visual in-spection of the threaded holes and a check of the screw-in depth on engagement must be performed 10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-14

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 before and after the load test. No adverse effects on the trueness to gauge of the threaded holes are permissible. The cask lid must be replaced by a new exemplar if the threaded holes do not pass the static load test.

10.1.3.2.5 Transfer Cask Trunnions The static test load must be distributed equally on the two trunnions of the transfer cask. The direc-tion of load application must be perpendicular to the axis of the trunnions. Bearing shells made of bronze must be placed between lifting lug and the trunnion. The following requirements must be met:

After the load test, a visual inspection and MT must be performed on the installed trunnions, as far as accessible, to verify no distortion or cracking has occurred. MT shall be performed in accordance with Section V, Article 7, using the acceptance criteria specified in Division 3, NF-5342. The trunnion deformation between the initial value prior to loading (null measurement) and after loading with the test load shall be determined via dial gauge (measuring accuracy+/- ***I). A permanent trunnion deformation exceeding the measuring inaccuracy of the dial gauge is unacceptable Otherwise, the deformed trunnions shall be replaced by new exemplars and the load test shall be repeated.

The cap screws that are exposed to the highest loading during the static lead test shall be removed and visually inspected after the test. Thread testing of these screws and the corresponding threaded holes in the transfer cask shall be performed. Defective cap screws shall be replaced with new ex-emplars. The tightening torque of all cap screws shall be checked after the load test.

10.1.4 Leak Tests Subsequently to the pressure tests carried out on the initial samples of storage cask and canister, an integrated LT is performed on both containments according to Division 3, WC-6224. They are omitted during serial production as discussed and assessed in Section 2.0. If deemed necessary, a pressure test can of course be carried out in individual cases.

As justified in Section 2.0 the monolithic cask body, the canister body and the lids are considered leak-tight with respect to the required maximum permissible standard helium leakage rate of 10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-15

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 L = 1o-s Pa m3 /s due to demanding quality assurance measures regularly performed during fabrica-tion. Thus, LT of the containment is reduced to LT of the metal gasket sealing systems.

At any time the containment boundaries of storage cask or canister are completely set up (during assembly, dispatch after loading, etc.), helium LT (helium detection via mass spectrometer) shall be performed on all metal gaskets. LT is performed in accordance with Section V, Article 10 [13] and ANSI N14.5 [19].The leak detector shall have a sensitivity of at least 10-10 Pa m3 /s. A calibration of the leak detector is performed using a test leak before connecting the measuring setup to the test volume. Table 10.1-6 lists all containment boundaries with metal gaskets that require leak testing.

Table 10.1-6: Containment boundaries with metal gaskets requiring LT Containment Component 1

• Metal Gasket

• Component 2*

Cask body (Item 2) Item 69, Cask lid (Item 55)

Storage cask Cask lid (Item 55) Item 44 Protection cap (Item 113)

Cask lid (Item 55) Item 71 Blind flange (Item 89)

Canister body (Item 2) Item 16 Canister lid {Item 3)

Canister Canister lid (Item 3) Item 13 Tightening plug (Item 10)

• Item No. according to the respective parts list of storage cask and canister (see Section 1.5, Appen-dixes 1-4 and 1-5)

The leak tightness is considered proven when the calculated standard helium leakage rate is lower than the permissible one. In case of an unacceptable leakage rate the corresponding sealing barrier is to be depressurised and opened. The metal gasket or elastomeric seal is to be replaced, the sealing surface checked for cleanliness and damage and necessary repairs performed. Containment material may be repaired in accordance with Division 3, WC-2500 or disposed of. After the new

- gasket and the sealing surface are preserved, the containment boundary is to be closed again and a new leak test has to be performed.

10.1.5 Components 10.1.5.1 Valves, Rupture Discs, and Fluid Transport Devices There are no valves, rupture discs or fluid transport devices associated with the containment bound-aries of the CASTOR geo69 DSS. The only valve-like components in the DSS are quick connections installed in the vent and drain ports in the service orifice of canister and cask lid. After completion of drying and helium backfill operations, the service orifice in the canister lid is covered and sealed by the tightening plug and a metal gasket and the one in the cask lid by the protection cap and a metal gasket. LT to verify canister and cask containment boundaries is performed. The quick connections 10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-16

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 are thus not accessible during transport or storage unless the containment boundaries remain closed.

The pressure in the water chambers of the transfer cask is limited by means of pressure relief valves to be plugged to the quick connection in the top side of the outer water chamber. The pressure relief valves shall relieve at an overpressure of 10.1.5.2 Seals and Gaskets All elastomeric seals (sealing rings, O-rings, etc.) used with the DSS and CLU are not important to safety. However, prior to usage, each elastomeric seal (and the corresponding sealing surface) is visually inspected.

-I The sealing performance of metal gaskets and the properties of the corresponding materials are specified in Section 8.2. Each metal gasket (and the corresponding sealing surface) is visually in-spected prior to installation and helium leak tested after closure of the corresponding lid or cover lid.

The creep of the metal gaskets is limited to an extent that will not degrade its sealing performance during storage.

10.1.5.3 Miscellaneous 10.1.5.3.1 FA receptacle calibre test After assembly of the basket a FA receptacle calibre test is used to simulate the loading of the canister with SNF in the nuclear facility. A replica of the intended FA is successively lowered into each receptacle of the basket to check if the fuel elements fit into the basket as requested. The calibre test is successfully completed when the replica fits into each receptacle without without jam-ming.

10.1.5.3.2 Weighing of canister and transfer cask As required by Section 2.0 the total mass of the transfer cask carrying a canister loaded with SNF must not exceed the maximum crane capacity of

• However, the masses of the respective components of canister with basket and shielding elements and CLU tabulated in Section 1.2 may result into a total weight of the loaded transfer cask of more than this limit, if all tolerances and a traverse weighing lare included in the considerations. Therefore, following the respective as-sembly, weighing of both, the transfer cask as well as the canister with basket and shielding elements 10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-17

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 must be performed. The results of the weighing must be assessed with the limits listed in Ta-ble 10.1-7. If an actual manufactured transfer cask weights less than the tabulated limit canisters to be handled may weight respectively more unless the maximum crane capacity is not exceeded.

Table 10.1-7: Mass limits on transfer cask and canister (including internals)

Component' Transfer cask Canister (including internals) --

Mass limit [kg]

10.1.5.4 Shielding Test The performed shielding evaluation provided in Chapter 5 shows that the dose rates calculated un-der conservative assumptions are always considerably below the regulatory dose rate limits for NCS and ACS. The components and materials serving as radiation shield for neutron and/or gamma ra-diation are described in Section 1.2. Material and component tests are performed to eliminate the possibility of defects, uncontrollable voids or streaming paths in the shielding, which could lead to a deviation from the calculated radiation profile and a local violation of a dose rate limit. Significant degradation, gas release, or physical alteration is not to be assumed. Each shielding component is visually inspected to ensure homogeneity and the absence of cracks, voids, shrinkage holes, pin-holes and other defects in the material. The chemical composition of the materials and their proper-ties ascertain that they exhibit the desired shielding properties used in the calculations. Hence, the intended shielding performance can be assumed throughout all loading operations and the entire storage time including maintenance operations. No additional shielding tests are designated for ac-ceptance of shielding components.

10.1.6 Shielding Integrity As described in Section 1.2, the various components of the storage cask and CLU contribute in certain ways to the attenuation of neutron and gamma radiation. Heavy components with high den-sity mainly contribute to gamma shielding, lightweight materials with high hydrogen content to neu-tron attenuation. Regarding the storage cask the DCI cask body, the stainless steel lids, bottom closure plate and trunnions as well as the steel bars in the bottom end of the deep hole drillings for the moderator rods are particularly worth mentioning as gamma radiation shields. The canister body and lid (stainless steel) also make a major contribution to gamma shielding. Due to the carbon con-tent of the DCI cask body and the boron-carbide content of the structural basket sheets they both also appreciably contribute to neutron attenuation. The moderator rods and plates (UHMW-PE) 10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-18

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS mainly attenuate neutron radiation. The transfer cask and transfer lock components made of steal (head and bottom ring, liner, frame, housing, etc.) and lead provide gamma shielding, water and PE encapsulated by stainless steel provide neutron attenuation.

10.1.6.1 Moderator Material Information on the moderator material, tempered UHMW-PE ( , is provided in a material qualification report included as Appendix 8.3 in Section 8.5. An elongation stress F 150/10 of is required. Preparation of test specimens and determination of properties is per-formed in accordance with EN ISO 21304-2 [20]. After tempering, each pressed plate is visually inspected for complete sintering, surface texture and optical inhomogeneity. The material must be free from bubbles, voids, foreign particles, cracks and other defects in accordance with EN ISO 15527 [21 ]. The density of the moderator material is measured in accordance with EN ISO 1183-1, method A [22] for each pressed plate after tempering. A density~ is required. The co-efficient of thermal expansion a is measured for each pressed plate. For as is required at a test temperature of Dimensional checks are to be applied to the finished components.

The moderator rods and plates may be composed of multiple parts. Thereby, care must be taken to ensure that the individual PE sections are form-fitted to each other in order to prevent neutron win-dows.

All examinations and test shall be performed according to written and approved procedures and the results shall become part of the quality documentation of the particular part, component or the DSS.

10.1.6.2 Neutron Absorber Material Top, centre and bottom sheets of the fuel basket are made of the Aluminium-Boron Metal Matrix Composite (AI-B4C-MMC) ****. which consists of a * * *

• Al alloy containing boron carbide (B4C). Information on acceptance tests and acceptance criteria relevant for the AI-B4C-MMC is provided in a material qualification report included as Appendix 8-2 in Section 8.5. The acceptance tests ensure the required mechanical properties, thermal properties, and neutron attenuation perfor-mance of the material.

10 The minimum required B4C content in Furthermore, a B content in boron o f * * * * *

• is required. Chemical composition is determined for each batch of mate-rial. Boron distribution tests are performed on each extruded profile to ensure uniformity of boron in the fuel basket material. Tensile testing is performed on 10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-19

Non-Proprietary Version 1014-SR-00002 Rev. 1 Proprietary Information withheld per 10CFR 2.390 s

each extruded profile to demonstrate that the material will meet the required mechanical properties.

The acceptance criteria for tensile testing are summarized in Table 10.1-8. The final profiles are visually inspected and their dimensions are checked to meet the requirements according to the draw-ings.

All examinations and test shall be performed according to written and approved procedures and the results shall become part of the quality documentation of the particular part, component or the DSS.

Table 10.1-8: Mechanical Requirements for Al-84C-MMC neutron absorber material Property Tensile strength Rm [MPa]

Yield strength Rpo2 [MPa]

Elongation at fracture A[%]

Test condition: RT Test condition: 300°C 10.1.6.3 DCI, Stainless Steel and Lead For acceptance tests and acceptance criteria of the DCI cask body it is referred to Subsection 10.1.1.

Analogously, the lead shield and the stainless steel components of storage cask and CLU are pro-cured and manufactured according to manufacturing and test specifications and fabrication specifi-cations and the codes and standards assigned in the parts lists in the Appendix of Chapter 1. Ac-ceptance tests and criteria regarding all relevant material and component requirements (chemical composition, dimensions, etc.) are specified in these manufacturing documents.

The lead shield of the CLU may either be poured or composed of multiple cast parts or lead sheets.

Regardless of the semi-finished product, care must be taken to ensure that the individual lead sec-tions are form-fitted to each other in order to prevent windows for gamma radiation and that the minimum required thickness of the lead shield according to the design drawing is met. Visual exam-inations regarding cracks, pores, inclusions, scratches, grooves, or other types of defects that pos-sibly reduce the gamma shielding function shall verify the effectiveness. To test the effectiveness of a poured lead shield, gamma scans shall be performed.

10.1.7 Thermal Test The thermal evaluation (see Chapter 4) provides reliable results in terms of three-dimensional tem-perature distribution and temperature time dependency in the DSS during storage and in the transfer cask during short-term operations. The results indicate that the maximum temperatures during both, short-term operations as well as NCS, off-normal conditions and ACS (fire phase and subsequent 10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-20

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 cooling phase) are considerably below the limit values of each component. The materials used in the CASTOR geo69 DSS and CLU design are applied in several other transport and storage casks with a comparable design since many years. Thereby GNS exhibits many years of experience in modelling and simulation of their thermal behavior. The acceptance tests performed during material fabrication ensure that each material is manufactured in accordance with the applicable standard, which leads to reproducible thermal properties as specified in Section 8.2. The thermal properties of each material and the heat transfer between these materials determine the thermal behavior and heat dissipation in the DSS and CLU. Based on the established experience and the previous verifi-cation of the applied simulation models it is ensured by GNS that the simulations describe the actual thermal behavior of the DSS under all conditions of storage and of the CLU during short-term oper-ations with adequate accuracy. Therefore, no further thermal tests are designated for acceptance.

10.1 Acceptance Criteria Section 10.1, Rev. 1 Page 10.1-21

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS List of References

[1] ASME Boiler and Pressure Vessel Code Section Ill - Rules for Construction of Nuclear Facility Components Division 1, Subsection NF - Supports 2017 Edition

[2] ASME Boiler and Pressure Vessel Code Section II - Materials, Part A~ Ferrous Material Specifications 2017 Edition

[3] ASTM E 1806 Standard Practice for Sampling Steel and Iron for Determination of Chemical Composition, Edition 2018

[4] ASTM E 351 Standard Test Methods for Chemical Analysis of Cast Iron - All Types, Edition 2018

[5] ASTM E 8 Standard Test Methods for Tension Testing of Metallic Materials, Edition 2016

[6] ASTM E 399 Standard Test Method for Linear-Elastic Plane-Strain Fracture Toughness of Metallic Mate-rials, Edition 2020

[7] ASTM E 1820 Standard Test Method for Measurement of Fracture Toughness, Edition 2020

[8] Standard Test Method for Evaluating the Microstructure of Graphite in Iron Castings, Edition 2019

[9] ASTM E 2567 Standard Test Method for Determining Nodularity And Nodule Count In Ductile Iron Using Image Analysis, Edition 2016

[1 O] ASTM E 562 Standard Test Method for Determining Volume Fraction by Systematic Manual Point Count, Edition 2019

[11] ISO1502(1996)

ISO general purpose metric screw threads - Gauges and gauging 0.0 Section 0.0, Rev. 1 Page 14.0-22

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS

[12] ASME Boiler and Pressure Vessel Code Section IX-Welding, Brazing, and Fusing Qualifications 2017 Edition

[13] ASME Boiler and Pressure Vessel Code Section V - Nondestructive Examination 2017 Edition

[14] ISO 5817 (2014)

Welding - Fusion-welded joints in steel, nickel, titanium and their alloys (beam welding ex-cluded) - Quality levels for imperfections

[15] ISO 9712 (2021)

Non-destructive testing - Qualification and certification of NOT personnel

[16] ISO 17637 (2016)

Non-destructive testing of welds - Visual testing of fusion-welded joints

[17] ISO 3452-1 (2014)

Non-destructive testing - Penetrant testing Part 1: General principles

[18] ANSI N14.6 (1993)

Radioactive Materials - Special Lifting Devices for Shipping Containers Weighing 10 000 Pounds (4500 kg) or More

[19] ANSI N14.5 (2014)

Radioactive Materials - Leakage Tests on Packages for Shipment

[20] EN ISO 21304-2 (2021)

Plastics - Ultra-high-molecular-weight polyethylene (PE-UHMW) molding and extrusion ma-terials Part 2: Preparation of test specimens and determination of properties

[21] EN ISO 15527 (2018)

Plastics - Compression-molded sheets of polyethylene (PE-UHMW, PE-HD)

Requirements and test methods

[22] EN ISO 1183-1 (2019)

Plastics - Methods for determining the density of non-cellular plastics Part 1: Immersion method, liquid pycnometer method and titration method 14.0 Overview Section 14.0, Rev. 1 Page 10.1-23

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 10.2 Maintenance Program Name, Function Date Signature Prepared Reviewed 10.2 Maintenance Program Section 10.2, Rev. 0 Page 10.2-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 A periodically ongoing maintenance program shall be defined and incorporated into the CASTOR geo69 DSS and CLU operations manuals, which shall be prepared and issued prior to the delivery and first use of the DSS to each user. These documents shall delineate the detailed inspections, maintenance, tests, and parts replacement necessary to ensure continued structural, thermal, and confinement performance; radiological safety, and proper handling of the system in accordance with 10 CFR 72.136(9) regulations, the conditions in the CoC, and the design requirements and criteria contained in this SAR.

The CASTOR geo69 DSS is passive by design. There are no active components required to as-sure the performance of its safety functions. The pressure switch in the cask lid, which is connect-ed to the pressure monitoring systems, is a self-reporting component, which means that a defect of the pressure switch will be automatically detected. As a result, only minimal maintenance will be required over the CASTOR geo69 DSS lifetime, and this maintenance would primarily include replacement routines. Typical of such maintenance would be the reapplication of corrosion inhibit-ing materials on accessible external surfaces or the replacement of the pressure switch in case of a reported defect.

The number of lifting and handling operations performed with the transfer cask and storage cask shall be documented in the cask logbook. Trunnions and trunnion bolts shall be replaced when the minimum number of permissible load cycles specified in Section 3.5 is reached.

The CLU components shall be inspected prior to usage.

Any maintenance operations shall be performed according to written and approved procedures.

Results of tests, repaired or replaced parts and other maintenance operations shall be recorded and become part or the quality documentation of the DSS or CLU.

10.2.1 Structural and Pressure Tests Prior to each fuel loading, a visual examination in accordance with a written procedure shall be performed on the lifting trunnions of the transfer cask and on the trunnions and tilting studs of the storage cask. The examination shall inspect for indications of overstress such as cracking, defor-mation, or wear marks. Repairs or replacement in accordance with written and approved proce-dures shall be required if unacceptable conditions are identified. Testing to verify continuing com-pliance of the transfer cask and storage cask trunnions shall be performed in accordance with ANSI N14.6 [1].

10.2 Maintenance Program Section 10.2, Rev. 0 Page 10.2-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 As described in Chapter 7 and 12, there are no credible normal, off-normal, or accident c.onditions that can cause the structural failure of the DSS. Therefore, periodic structural or pressure tests on the DSS following the initial acceptance tests are not required as part of the storage maintenance program.

10.2.2 Leak Tests The internal pressure of the storage cask is continuously monitored during storage via the pressure switch in the cask lid, which is connected to the pressure monitoring system of the storage facility.

A leak in the containment boundary of the storage cask or the canister would lead to a pressure drop in the storage cask. Such a pressure drop is automatically reported by the pressure switch.

Therefore, frequent leak tests are not required as part of the storage maintenance program. Leak tests are only performed on an as-needed basis when the pressure switch reports a pressure drop in the storage cask to determine the location of the leak.

10.2.3 Subsystem Maintenance The CASTOR geo69 DSS is connected to a pressure monitoring system during storage. Mainte-nance activities shall be performed by the licensee to ensure the operational reliability of the pres-sure monitoring system over the complete duration of storage.

Maintenance and calibration testing will be required on the vacuum drying, helium backfill, and leakage testing systems. Cranes and lifting beams shall be inspected prior to each loading cam-paign to ensure that proper maintenance and continued performance is achieved. Additional tem-porary neutron and gamma shielding provided during loading and transfer operations with the CLU requires no maintenance.

10.2.4 Valves, Rupture Discs, and Fluid Transport Devices The pressure relief valves used on the water jackets of the transfer cask shall be calibrated on an annual basis (or prior to the next use if the period the transfer cask is out of use exceeds one year) to ensure the pressure relief setting is or replaced with factory-set relief valves.

List of References

[1] ANSI N14.6 (1993)

Radioactive Materials - Special Lifting Devices for Shipping Containers Weighing 10 000 Pounds (4500 kg) or More 10.2 Maintenance Program Section 10.2, Rev. 0 Page 10.2-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 10.3 Appendix Prepared Reviewed 10.3 Appendix Section 10.3, Rev. 0 Page 10.3-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 With intent no items.

10.3 Appendix Section 10.3, Rev. 0 Page 10.3-2

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev. 1 11 Radiation Protection 11.0 Overview Prepared Reviewed 11.0 Overview Section 11.0, Rev. 1 Page 11.0-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 This chapter discusses the design considerations and operational features that are incorporated in the CASTOR geo69 DSS and CLU design to protect personnel and the public from exposure to radioactive contamination and ionizing radiation during operating procedures and long-term interim dry storage. Occupational exposure estimates for operating procedures are provided in Section 11.3, including canister loading, dispatch and transhipment via CLU and closure of the storage cask in the NPP. The occupational exposure is estimated for the procedures for preparation of the DSS at the ISFSI, as well. An off-site dose assessment (beyond the site boundary) of the DSS for a generic open-air storage facility is provided in Section 11.4. Since the determination of off-site doses is nec-essarily site-specific, similar dose assessments are to be prepared by the licensee, as part of imple-menting the CASTOR geo69 DSS in accordance with 10 CFR 72.212.

11. 0 Overview Section 11.0, Rev. 1 Page 11.0-2

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev. 1 11.1 Ensuring that Occupational Radiation Exposures are as Low as Reasonably Achievable (ALARA)

Prepared Reviewed 11.1 Ensuring that Occupational Radiation Exposures are as Low as Reasonably Achievable (ALARA)

Section 11.1, Rev. 1 Page 11.1-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 11.1.1 Policy Considerations The CASTOR geo69 DSS has been designed in accordance with 10 CFR 72. DSS as well as CLU maintain radiation exposures ALARA consistent with 10 CFR 20. Licensees using the CASTOR geo69 DSS and CLU will utilize and apply their existing site ALARA policies, procedures and prac-tices for onsite activities to ensure that personnel exposure requirements of 10 CFR 20 are met.

Personnel performing operations on the DSS or the CLU shall be trained on the operations and be familiarized with the expected dose rates. Pre-job ALARA briefings should be held with workers and radiological protection personnel prior to work on or around the DSS and CLU. Worker dose rate monitoring, in conjunction with trained personnel and well-planned activities, will significantly reduce the overall dose received by the workers. When preparing or making changes to site-specific proce-dures, users shall ensure that ALARA practices are implemented and the 10 CFR 20 standards for radiation protection are met in accordance with the site's written commitments.

Users can further reduce dose rates around the DSS and the CLU within the admissible loading patterns by preferentially loading longer-cooled and lower burnup spent fuel assemblies in the pe-riphery receptacles of the fuel basket, while loading assemblies with shorter cooling times and higher burnups in the inner basket receptacles.

11.1.2 Design Considerations Consistent with the design criteria defined in Subsection 2.3.5, the radiological protection criteria that limit exposure to radioactive effluents and direct radiation from the DSS are as follows:

• 10 CFR 72.104 requires that for normal operation and anticipated occurrences, the annual

- dose equivalent to any real individual located beyond the owner-controlled area boundary must not exceed 0.25 mSv to the whole body, 0.75 mSv to the thyroid, and 0.25 mSv to any other critical organ. This dose would be a result of planned discharges, direct radiation from the DSS/CLU, and any other radiation in the area. The licensee is responsible for demon-strating site-specific compliance with these requirements.

• 10 CFR 72.106 requires that any individual located on or beyond the nearest owner-con-trolled area boundary may not receive from any design basis accident the more limiting of a total effective dose equivalent of 0.05 Sv. The sum of the deep dose equivalent and the com-mitted dose equivalent to any individual organ or tissue (other than the lens of the eye) shall not exceed 0.5 Sv. The lens dose equivalent shall not exceed 0.15 Sv and the shallow dose equivalent to skin or to any extremity shall not exceed 0.5 Sv. The licensee is responsible for demonstrating site-specific compliance with this requirement.

11.1 Ensuring that Occupational Radiation Exposures are as Low as Reasonably Achievable (ALARA)

Section 11.1, Rev. 1 Page 11.1-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS

• 10 CFR 20, Subpart C specifies occupational dose limits for adults. The annual total effective dose equivalent shall not exceed 0.05 Sv. The sum of the deep-dose equivalent and the committed dose equivalent to any individual organ or tissue other than the lens of the eye shall not exceed 0.5 Sv in a year. Additionally, an annual limit of 0.15 Sv for the lens dose equivalent and an annual limit of 0.5 Sv for the shallow-dose equivalent to the skin of the whole body or to the skin of any extremity shall not be exceeded. The licensee is responsible for demonstrating site-specific compliance with this requirement.

• 10 CFR 20, Subpart D, specifies dose limits for individual members of the public. The total effective dose equivalent from licensed operations shall not exceed 1 mSv in a year. The dose rate in any unrestricted area from external sources shall not exceed 0.02 mSv/h. The licensee is responsible for demonstrating site-specific compliance with this requirement.

11.1.3 Operational Considerations The following operational considerations that most directly influence occupational exposures have been incorporated into the design of the CASTOR geo69 DSS and CLU:

• Use of a DPC for transport and storage, which requires no unloading/transfer of the canister from transport into a separate storage cask and vice versa at the storage facility and thus reduces on-site radiation exposure;

• A totally-passive DSS design requiring minimum maintenance and monitoring (other than security monitoring) during storage;

• A self-reporting pressure switch in the cask lid that is connected to a pressure monitoring system at the storage facility during long-term interim dry storage to allow remote monitoring of the storage cask internal pressure;

• Mostly passive CLU designs requiring minimum human interactions in the vicinity;

• Remotely operated transfer lock, lifting gear, etc.;

• Use of remote handling equipment, where practical

• Use of e.g. quick connection couplings in the service orifices of the lids

• Use of additional temporary neutron and gamma shielding during canister dispatch for load-ing into the CASTOR geo69 storage cask;

• Inspections and function tests prior to actual loading

• A sequence of operations based on ALARA considerations.

• Dry run trainings for the workers prior to actual FA handling 11.1 Ensuring that Occupational Radiation Exposures are as Low as Reasonably Achievable (ALARA)

Section 11.1, Rev. 1 Page 11.1-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 (@)GNS Operating controls and limits that are necessary for compliance with regulatory requirements and ALARA objectives are specified in Chapter 13.

11.1.4 Temporary additional Shielding To minimize occupational dose during loading and unloading operations, a specially designed set of temporary shielding (not part of DSS or CLU but of multi-equipment) will be used during loading and unloading operations performed with the CASTOR geo69 CLU. The temporary shielding comprises additional temporary shielding (e.g. lead blankets, PE plates) for the lid area of transfer cask and storage cask and a shielding plate for the canister lid. Each temporary shield is described in Ta-ble 11.1-1 and the procedures for utilization are provided in Chapter 9. Table 11 .1-1 provides the minimum requirements for the use of temporary additional shielding. Users shall evaluate the need for additional temporary and temporary shielding and use of special tooling to reduce the overall exposure based on an ALARA review of cask loading operations and the loaded contents.

Table 11.1-1: Temporary additional shieldings to be used with the CASTOR geo69 DSS and the CLU Temporary Shield Description Utilization Temporary additional neutron Installation in the reactor hall after placement and gamma shielding (e.g. lead of transfer cask (including the loaded canister)

Transfer cask blankets, PE plates) for the lid in the service platform. Removal before trans-area of the transfer cask. fer to the CASTOR geo69 storage cask.

Installation in the truck lock after loading of the Temporary additional neutron canister into the CASTOR geo69 storage and gamma shielding (e.g. lead Storage cask cask. Removal after closure of the storage blankets, PE plates) for the lid cask and in case of any maintenance work area of the storage cask.

during storage.

Installation before vacuum drying of the canis-Shielding plate for the canister Canister ter cavity in the reactor hall. Removal before lid.

transfer to the CASTOR geo69 storage cask.

The use of temporary additional shielding is required at the lid areas of canister, transfer cask and storage cask. The lateral area of the transfer cask provides sufficient shielding during loading and unloading operations due to the integrated lead shield and the two water chambers. During canister transhipment in the truck lock, the contact area between transfer cask and storage cask is sufficiently shielded by the transfer lock. Therefore, the use of additional temporary shielding at the lateral area of transfer cask and storage cask is optional.

11.1 Ensuring that Occupational Radiation Exposures are as Low as Reasonably Achievable (ALARA)

Section 11.1, Rev. 1 Page 11.1-4

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev. 1 ()GNS 11.2 Radiation Protection Design Features Prepared Reviewed 11.2 Radiation Protection Design Features Section 11.2, Rev. 1 Page 11.2-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The development of both, the CASTOR geo69 DSS as well as of the CLU have focused on design provisions to address the considerations summarized in Subsection 11.1.2 and 11.1.3. The intent has been to combine a canister-based system with the proven CASTOR type dual-purpose cask design for transport and storage requiring no additional loading or unloading procedures after closure of the cask in the reactor building. Canister and storage cask form a double-containment system around the loaded SNF. Both closure systems are re-openable and provide sufficient activity reten-tion. The design thus combines the preferred features of canister-based and metal cask systems.

This approach reduces overall radiation levels and the need for performing operating procedures and maintenance. The following specific design features of DSS and CLU ensure a high degree of containment integrity and radiation protection:

• Two independent containment barriers of the DSS

• Reduction of streaming paths to a minimum

• DSS equipped with o Thick monolithic DCI wall and bottom of the storage cask and a thick stainless steel canister and cask lid system; o UH MW-PE rods in the wall and plates in the cask bottom and below the cask lid; o Shielding elements in the canister; to reduce the surface dose rates;

• CLU equipped with o Thick stainless steel wall and bottom lid as well as shield in the wall of the transfer cask; o Thick ******** of the transfer lock; to reduce the surface dose rates;

• Material selection and smooth surface preparation (e.g. coating, almost no crud traps) of DSS and CLU to enable easy decontamination;

• A totally passive design requiring a minimum of maintenance during handling and storage.

11.2 Radiation Protection Design Features Section 11.2, Rev. 1 Page 11.2-2

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev. 1 11.3 Occupational Exposures Prepared Reviewed 11.3 Occupational Exposures Section 11.3, Rev. 1 Page 11.3-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS This section provides the estimates of the cumulative exposure to personnel performing loading, unloading and canister transshipment operations as well as receipt of the loaded DPC in transport configuration after transport on public roads in the storage facility and preparation of the DSS for long-term storage. Furthermore, surveillance and maintenance operations using the CASTOR geo69 DSS and the CLU are discussed. This section uses the shielding evaluation provided in Chap-ter 5 to develop dose assessments for the operating procedures provided in Chapter 9. The dose rates from the CASTOR geo69 storage cask and the CLU, each housing the loaded canister, are calculated at various positions (height from ground and distance from lateral area) around the (trans-fer/storage) casks lateral and lid area to determine the cumulative dose to personnel performing the operations.

11.3.1 Dose Rate Calculations The cumulative exposure to personnel performing the procedure described in Chapter 9 is calculated for a homogeneous content loading of the cask in accordance with loading pattern TR1 (see Chap-ter 5). The exposure doses are product of the dose rates at certain locations and the corresponding dwell time of the personnel involved in the handling operations. Influences of other radiation sources, such as background radiation, are not taken into account.

MCNP [1] is used for the shielding calculation for all handling operations with relevant contribution to the cumulative dose to the staff starting from the lifting of the loaded transfer cask out of the SNF pool until the CASTOR geo69 DSS is connected to the cask monitoring system of the ISFSI. The detailed 3O-model of the cask and its inventory from Chapter 5 is taken as a basis. Radiation from the active zone of the FAs, the head and foot pieces and from the structural parts are considered as radiation sources, whereby the dose rates are calculated separately for neutrons, gamma radiation from the fuel stack, and gamma radiation from the fuel assembly nozzles and plena.

The calculation for the loading procedure is performed with and without the temporary additional shielding specified in Subsection 11.1.4.

The same dose reduction is assumed for the temporary shielding used during the operating procedures for preparation of the DSS in the ISFSI.

The shielding effect of the service stations in the reactor hall, in the truck lock and in the storage facility, which are required for the handling procedure, are neglected. Furthermore, the protection cover of the DSS is neglected conservatively and for simplicity.

11.3 Occupational Exposures Section 11.3, Rev. 1 Page 11.3-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 11.3.2 Estimated Exposures During Operations in the NPP 11.3.2.1 Estimated Exposures during Loading the Cask This subsection deals with the occupational dose during the operational procedures loading, unload-ing and canister transshipment as well as dispatch of canister and cask.

As they are basically the same, the detector positions indicated in Figure 11.3-1 apply to both, the transfer cask (CLU) and the CASTOR geo69 storage cask solely some additional points are mod-elled for the storage cask (see Figure 11.3-4). The positions are enumerated with letters from A to Z. Each position is a possible inhabitancy for personnel during the operating procedures described in Sections 9.1 and 9.2.

- F' E'

C B

F E H L

K 0

N R

Q top edge / lid A D G J M p

• z (trunnion area) s T u

, , , (middle of the fuel stack)

'V w X Figure 11.3-1: Relevant positions at the transfer/storage cask surface 11.3 Occupational Exposures Section 11.3, Rev. 1 Page 11.3-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The shielding models of the DPC and of the CLU described in Chapter 5 are modified to represent the situations during corresponding processing steps. The following four shielding configurations are modeled for the calculation of the cumulative dose to personnel:

1. Shielding configuration 1: (Figure 11.3-2)

Location: Service station next to the SNF pooL

Description:

2. Shielding configuration 2: (Figure 11.3-3)

Location: Service station in the truck lock.

Description:

Canister completely dispatched, transfer cask positioned on top of the transfer lock atop the CASTOR geo69 storage cask,

3. Shielding configuration 3: (Figure 11.3-4)

Location: Service station in the truck lock.

Description:

Loaded CASTOR geo69 storage cask without cask lid rest-ing on the concrete foundation of the truck lock, transfer cask and transfer lock removed.

4. Shielding configuration 4: (Figure 11.3-5)

Location: Service station in the truck lock.

Description:

Loaded CASTOR geo69 storage cask with cask lid resting on the concrete foundation of the truck lock.

The occupational dose evaluations for the four shielding configurations are listed in Table 11.3-1.

For each relevant sub-step of each operation, the dwell time, the number of persons involved, the position of the persons (e.g. according to Figure 11.3-1), the dose rate (without additional temporary shielding) at this position, the cumulative dose and the reduced dose with additional temporary shielding (if applicable) are listed. The dose is calculated by multiplying the dose rate at the respec-tive position with the dwell time required for the operational sub-step and the number of persons involved in the operation. Certain handling operations require the abidance of personnel at more than one position for different dwell times. Therefore, the doses accumulated at every position are listed separately.

The analysis results in a total occupational dose of approximately - per cask which can be reduced to approximately - when using additional temporary shielding. It is obvious that the application of additional temporary shielding is highly recommended. However, in this regard it is 11.3 Occupational Exposures Section 11.3, Rev. 1 Page 11.3-4

Non-Proprietary Version 1014-SR-00002 Rev. 1 Proprietary Information withheld per 10CFR 2.390 s

important to point out that the total occupational dose is distributed among several individuals not least because many handling operations demand more than one person.

It has to be mentioned that handling operations included in the procedure for loading the cask ac-cording to Section 9.1 but not relevant for the exposure calculation due to large distance from the radiating source are not considered in Table 11.3-1. An example is the installation of the blind flange (with metal gasket) in the cask lid and the check for proper installation, which is performed before the cask lid is positioned and installed on the CASTOR geo69 storage cask. The sequence of steps in Table 11.1-1 is in accordance with Section 9.1.

Figure 11.3-2: Shielding configuration 1 - Loaded transfer cask 11.3 Occupational Exposures Section 11.3, Rev. 1 Page 11.3-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Figure 11.3-3: Shielding configuration 2 - Canister transhipment 11.3 Occupational Exposures Section 11.3, Rev. 1 Page 11.3-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 11.3 Occupational Exposures Section 11.3, Rev. 1 Page 11.3-7

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Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Table 11.3-1: Occupational doses for CASTOR geo69 loading Operation Sub-steps Time Pers. Pos. Dose rate Dose Reduced dose*

[min] [mSvlh] [mSv] [mSv]

Shielding configuration 1 Crane trans-fer of the loaded trans-fer cask to the service station next to SNF pool Dispatch of Transfer to service station Removal of traverse Removal of lifting pintle I - -

canister and Removal of guide bolts transfer cask at the service Dewatering and drying of station annulus between canister and transfer cask cavity Drying of canister lid area Installation of additional temporary shielding I I I I Check for proper installation of canister lid Vacuum drying of canister cavity I I Installation of blind plug and quick connect Evacuation and helium fill-ing of canister cavity I I Installation of tightening plug with metal gasket Fastening of pressure nut

-- I -- -- --

Check for proper installation of pressure nut Leakage test of canister lid system Removal of additional tern-I porary shielding Installation of lifting pintle 11.3 Occupational Exposures Section 11.3, Rev. 1 Page 11.3-9

Non-Proprietary Version 1014-SR-00002 Rev. 1 Proprietary Information withheld per *1 OCFR 2.390

@ NS Operation Sul>steps Time Pers. Pos. Dose rate Dose Reduced dose*

[min] [mSvlh] [mSv] [mSv]

-- I -- --

Transfer of Attachment of traverse on transfer cask trunnions to the CASTOR geo69 stor- Crane transfer to truck lock I

age cask Positioning of transfer cask in transfer lock on storage cask -I Shielding configuration 2 I - -

Canister transhipment

- Shielding configuration 3 Removal of CLU compo-nents and I I I - -

canister lift-ing pintle Installation of I

retention ring in the CASTOR geo69 star-age cask I

Dispatch of Removal of sealing surface CASTOR protection geo69 stor-Installation of guide bolts age cask Positioning of cask lid (load attachment on lifting pintle)

I I Shielding configuration 4 Installation of additional temporary shielding 11.3 Occupational Exposures Section 11.3, Rev. 1 Page 11.3-10

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Operation Sub-steps Time Pers. Pos. Dose rate Dose Reduced dose*

[min] [mSvlh] [mSv] [mSv]

Installation of cask lid bolt-ing I I Check for proper installation of cask lid Removal of lifting pintle Installation of blind flange or pressure switch with metal gasket and cap screws in cask lid and check for proper installation Vacuum drying of cask cav-ity I I Helium back filling of cask cavity and adjustment of pressure I I Installation of protection cap with metal gasket and cap screws in cask lid and check for proper installation Leakage test of protection cap I I

--* I I I Leakage test of blind flange Leakage test of cask lid Removal of additional tern-I porary shielding Preparation for transport of the storage cask out of the building I I Total occupational dose*

• Reduced doses can be achieved by temporary additional shielding which is not mandatory but highly recom-mended as this reduces the occupational dose of the personnel significantly.

11.3 Occupational Exposures Section 11.3, Rev. 1 Page 11.3-11

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 11.3.2.2 Estimated Exposures during Unloading the Cask The occupational dose for the procedure for unloading the cask in accordance with Section 9.2 is expected to be smaller than the occupational dose for loading the cask. Most sub-steps of the un-loading procedure, which are relevant for occupational exposure, are consistent with sub-steps of the loading procedure, but are performed in reverse order. The four shielding situations described in Subsection 11.3.2.1 appear in reverse order during the unloading procedure.

Unless there are no reasons for intervention, unloading ofthe CASTOR geo69 storage cask is expected to take place after the intended storage period of the DSS. During long-term interim dry storage, the total activity of the loaded contents decreases exponentially, leading to lower radiation source terms compared to those specified in Section 5.2. Radiation resulting from cask activation as a consequence of neutron irradiation during storage can be neglected according to Section 2.4.

Therefore, occupational dose rates during unloading of the cask are lower than during loading of the cask.

11.3.3 Estimated Exposures during Operations in the Storage Facility 11.3.3.1 Estimated Exposures during Receipt by and Storage in the Storage Facility This subsection on the one hand deals with the occupational dose during delivery and acceptance of a transport package of the type CASTOR geo69 (equipped with impact limiters) by a storage facility after transport on public routes, taking into account the relevant inspections in the process.

This assumption is covering the delivery and acceptance of a storage cask after on-site transfer without impact limiters from the NPP to the storage facility. On the other hand, exposures to the personnel during transfer of the DPC to the storage position and during preparation of the DSS for long-term dry storage are estimated. The following three shielding configurations are dosimetrically evaluated:

1. Shielding configuration 5: (Figure 11.3-6)

Location: Reception and receiving area of the storage facility

Description:

The CASTOR geo69 transport package, i.e. cask equipped with impact limiters is positioned on the transport vehicle in horizontal orientation above the ground of concrete (in order to take scattering into account). Possible shielding impact of the transfer vehicle is completely neglected, the DPC is mod-elled freely floating above the ground with no support. Rele-vant configuration as long as impact limiters a installed.

2. Shielding configuration 6: (Figure 11.3-7)

Location: Reception and receiving area of the storage facility 11.3 Occupational Exposures Section 11.3, Rev. 1 Page 11.3-12

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Description:

CASTOR geo69 DPC positioned on the transport vehicle in horizontal orientation above the ground of concrete (in order to take scattering into account). Possible shielding impact of the transfer vehicle is completely neglected, the DPC is mod-elled freely floating above the ground with no support. Rele-vant configuration until tilting the cask into upright position.

3. Shielding configuration 7: (Figure 11.3-8)

Location: Service station of the storage facility and storage position

Description:

CASTOR geo69 DPC in vertical orientation in its final stor-age position being prepared for the long-term interim stor-age. Possible shielding impact of the protection cover is completely neglected A configuration considering the protection cover is neglected conservatively and for simplicity. The analysed positions for the respective shielding configuration are each indicated with letters. Each position is a possible inhabitancy for personnel during the operating procedures described in Section 9.3.

The occupational dose evaluations for the three shielding configurations are listed Table 11.3-2. For each relevant sub-step of each operation, the dwell time, the number of persons involved, the posi-tion of the persons, the dose rate (without temporary additional shielding) at this position, the cumu-lative dose and the reduced cumulative dose with additional temporary shielding (if applicable) are listed. The dose is calculated by multiplying the dose rate at the respective position with the dwell time required for the operational sub-step and the number of persons involved in the operation. Cer-tain handling operations require the abidance of personnel at more than one position for different dwell times. Therefore, the doses accumulated at every position are listed separately.

The analysis results in a total occupational dose of approximately per cask which can be reduced to approximately when using additional temporary shielding. It is obvious that the application of additional temporary shielding is highly recommended. However, in this regard it is important to point out that the total occupational dose is distributed among several individuals not least because many handling operations demand more than one person.

It has to be mentioned that handling operations included in the procedure according to Section 9.3 but not relevant for the exposure calculation due to large distance from the radiating source are not considered in Table 11.3-2. The sequence of steps in Table 11.3-2 is in accordance with Section 9.3.

11.3 Occupational Exposures Section 11.3, Rev. 1 Page 11.3-13

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 11.3-6: Shielding configuration 5 - CASTOR geo69 in transport configuration (top view on the transport package (incl. impact limiters) in horizontal ori-entation.)

11.3 Occupational Exposures Section 11.3, Rev. 1 Page 11.3-14

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 11.3-7: Shielding configuration 6 - CASTOR geo69 storage cask in horizontal ori-entation (left: side view, right: top view) 11.3 Occupational Exposures Section 11.3, Rev. 1 Page 11.3-15

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Figure 11.3-8: Shielding configuration 7 - CASTOR geo69 storage cask in vertical orien-tation (side view) 11.3 Occupational Exposures Section 11.3, Rev. 1 Page 11.3-16

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 (@)GNS Table 11.3-2: Occupational doses During Receipt by and Storage in the Storage Facility Operation Sub-steps Time Pers. Pos. Dose rate Dose Reduced dose*

[min] [mSv/h] [mSv] [mSv]

Inspections Shielding Configuration 5 during re-ceipt of Removal of preservation of lid impact limiter screws transport package Removal of preservation of bottom impact limiter screws Removal of lid impact limiter screws Removal of bottom impact limiter screws Load attachment and re-moval of lid impact limiter Load attachment and re-moval of bottom impact lim-iter Shielding Configuration 6 Removal of Load attachment and re-moval of Contamination measure-ment on cask Dose rate measurement on cask Visual inspection of outer surface coating Visual inspection of cask lid system Visual inspection of trun-nions Visual inspection of tilting studs and wear protection incl. preservation Visual inspection of closure plate and screws incl.

preservation Leak-tightness test of clo-sure plate Installation of Load attachment on trun-the pressure nions switch Tilting of the storage cask 11.3 Occupational Exposures Section 11.3, Rev. 1 Page 11.3-17

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@ NS Operation Sub-steps Time Pers. Pos. Dose rate Dose Reduced dose*

[min] [mSvlh] [mSv] [mSv]

Shielding Configuration 7 Vertical crane transfer to service station Installation temporary addi-tional shielding*

Removal of protection cap incl. screws and metal gas-ket Controlled pressure normal-isation in cask cavity via quick connect Removal of blind flange incl.

screws and metal gasket Visual inspection of pres-sure switch, sealing sur-faces of pressure switch and cask lid, metal gasket and screws Installation of pressure switch without metal gasket and check for proper lid in-stallation Removal of the pressure switch Installation of pressure switch with metal gasket and check for proper lid in-stallation Functional test of pressure switch Helium backfilling of storage cask via quick connect Leakage test of metal gas-ket in pressure switch Visual inspection of protec-tion cap, sealing surfaces of protection cap and cask lid, metal gasket and screws Installation of protection cap without metal gasket and check for proper lid installa-tion Removal of the protection cap Installation of protection cap with metal gasket and check for proper lid installa-tion 11.3 Occupational Exposures Section 11.3, Rev. 1 Page 11.3-18

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Operation Sub-steps ***.Time' Pers:* *POS.* Doseiate I Dose Reduced

• • • *' dose*

[min] ,. [mSvlh] [mSv], [mSv].

Leakage test of metal gas-ket in protection cap Removal of temporary addi-tional shielding*

Set-up of the Load attachment on trun-CASTOR nions geo69 DSS Removal of service platform and vertical crane transfer to designated storage posi-tion Removal of traverse Installation of protection cover Installation of cable conduit in protection cover Connection to the cask monitoring system Total occupational dose*

• Reduced doses can be achieved by temporary additional shielding which is not mandatory but highly recommended as this reduces the occupational dose of the personnel significantly.

11.3.3.2 Estimated Exposures during Surveillance and Maintenance An estimation of the occupational exposure required for security surveillance and maintenance dur-ing long-term interim dry storage of the CASTOR geo69 DSS is necessarily site specific.

Security surveillance time is based on a daily security patrol around the controlled area boundary of the storage facility. According to the results of the shielding evaluation presented in Section 5.1, the CASTOR geo69 DSS complies with 10 CFR 72.104 and the maximum permitted annual dose equivalent of 0.25 mSv at the controlled area boundary for normal operations and anticipated occur-rences, assuming 100 % occupancy (8766 hours0.101 days <br />2.435 hours <br />0.0145 weeks <br />0.00334 months <br />) and a bounding array of There-fore, the occupational annual dose due to security surveillance is only a small fraction of this limit value. A duration o f

• for a daily security patrol around the controlled area boundary leads to a maximum annual dose equivalent of approximately This complies with the occupational dose limits specified in 10 CFR 20, Subpart C.

Occupational exposure for maintenance operations is reduced to a minimum due to the passive design of the CASTOR geo69 DSS. Only minimum maintenance will be required over the lifetime of the DSS. An ongoing maintenance program, including specified maintenance procedures and intervals, will be defined and incorporated into the CASTOR geo69 DSS operations manual, prior 11.3 Occupational Exposures Section 11.3, Rev. 1 Page 11.3-19

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 s to the delivery and first use of the DSS. The maintenance procedures and intervals shall comply with the requirements of 10 CFR 20, Subpart C.

Maintenance operations on the CLU components are most likely performed while the CLU is not in use and thus without any radiation exposure from radioactive contents. Radiation exposure for maintenance of CLU components is negligible since the transfer cask is decontaminated after load-ing in the SNF pool and the exposure time (neutron irradiation during loading of the cask) is too short to lead to significant radiation resulting from activation of CLU components.

List of References

[1] C.J. Werner (ed.), MCNP User's Manual - Code Version 6.2, LA-UR-17-29981, 2017 11.3 Occupational Exposures Section 11.3, Rev. 1 Page 11.3-20

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev. 1 QGNS 11.4 Exposures at or Beyond the Controlled Area Boundary Prepared Reviewed 11.4 Exposures at or Beyond the Controlled Area Boundary Section 11.4, Rev. 1 Page 11.4-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 This section only considers contributions to the annual dose at the controlled area boundary origi-nating for the DSS in the ISFSI. As it is shown in Chapter 5, contributions from loading operations inside the NPP are negligible.

11.4.1 Controlled Area Boundary Dose for Normal Conditions of Storage 10 CFR 72.104 limits the annual dose equivalent to any real individual at or beyond the controlled area boundary to a maximum of 0.25 mSv to the whole body, 0. 75 mSv to the thyroid, and 0.25 mSv for any other critical organ. This includes contributions from all uranium fuel cycle operations in the region, including CLU operations. Radiation exposure in connection with handling operations involv-ing the CLU contributes to the controlled are boundary dose for NCS.

- Dose rates and doses at the controlled area boundary result from the direct neutron and gamma radiation stemming from the loaded CASTOR geo69 DSS. The structural integrity and the redun-dant containments are not impaired during normal, off-normal and accident conditions of storage, as demonstrated in Chapter 7. Therefore, no radioactive material is released from the DSS that could contribute to the controlled area boundary dose.

It is not feasible to predict bounding controlled area boundary dose rates on a generic basis since radiation from plant and other sources; the location and the layout of the storage facility and the number and configuration of storage casks are necessarily site-specific. In order to compare the performance of the CASTOR geo69 DSS with the regulatory requirements, a bounding array of storage casks is analysed in Chapter 5. Users are required to perform a site specific dose analysis for their particular situation in accordance with 10 CFR 72.212. The analysis must account for size, configuration and FA specifics of the storage installation and any other radiation from uranium fuel

- cycle operations within the region.

Section 5.1 presents dose rates and annual doses at the storage site for an array as a function of distance. 100 % occupancy (8766 hours0.101 days <br />2.435 hours <br />0.0145 weeks <br />0.00334 months <br />) is conservatively assumed.

10 CFR 72.106 specifies that the minimum distance from the storage facility to the nearest boundary of the controlled area must be at least 100 m. For open-air storage without additional shielding pro-vided by a storage building, the minimum distance needed to meet the annual dose requirements of 10 CFR 72.104 is approx. -away from the centre of the long side of the DSS array specified in Section 5.1.

11.4 Exposures at or Beyond the Controlled Area Boundary Section 11 .4, Rev. 1 Page 11.4-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 These results demonstrate the compliance of the CASTOR geo69 DSS design with the require-ments listed in Subsection 11.1.2. Users are required to perform a site-specific analysis to demon-strate compliance with 10 CFR 72 104 and 10 CFR 20. Neither the distances nor the array configu-rations specified in this subsection become part of the technical specifications of the DSS or the ISFSI.

11.4.2 Controlled Area Boundary Dose for Off-normal Conditions of Storage As demonstrated in the shielding evaluation in Chapter 5, there are no factors influencing the shield-ing performance of the storage cask under off-normal conditions of storage. None of the off-normal conditions analysed have an impact on the shielding analysis. The only significant difference be-tween off-normal conditions and NCS is that 10 % fuel rod failure is assumed instead of 3 %. Fuel rod failure under normal or off-normal conditions of storage does not lead to an increase of the dose rate. Therefore, the dose at the controlled area boundary from direct radiation for off-normal condi-tions is equal to that for NCS.

11.4.3 Controlled Area Boundary Dose for Accident Conditions of Storage 10 CFR 72.106 specifies the maximum allowed doses to any individual at the controlled area bound-ary from any design basis accident. In addition, it is specified that the distance to the controlled area boundary must be at least 100 m.

- For accident conditions of storage, it is conservatively assumed that all moderator material in the J DSS is lost. These assumption is bounding for all design basis accidents specified in Subsec-tion 2.2.3. The Jjose at a distance of 100 m remains safely under the maximum total effec-tive dose equivalent of 0.05 Sv specified in 10 CFR 72.106, as demonstrated in Section 5.1. The DSS thus fulfils the shielding requirements for ACS at the controlled area boundary for the worst case shielding consequence.

11.4 Exposures at or Beyond the Controlled Area Boundary Section 11.4, Rev. 1 Page 11.4-3

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev.O @GNS 11.5 Appendix Prepared Reviewed 11.5 Appendix Section 11.5, Rev. 0 Page 11.5-1

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11.5 Appendix Section 11.5, Rev. 0 Page 11.5-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev.1 GNS 12 Accident Analyses 12.0 Overview Prepared Reviewed 12.0 Overview Section 12.0, Rev. 1 Page 12.0-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS This chapter presents the evaluation of the CASTOR geo69 DSS and the CLU for the effects of off-normal and accident conditions of storage. The design basis off-normal and accident conditions of storage, including events resulting from mechanistic and non-mechanistic causes as well as those caused by natural phenomena, are identified in Section 2.2. For each postulated event, the event cause, means of detection, consequences, and corrective action are discussed and evaluat-ed. As applicable, the evaluation of consequences includes structural, thermal, shielding, criticality, containment, and radiation protection evaluations for the effects of each design event.

The structural, thermal, shielding, criticality, and containment features and performance of the CASTOR geo69 DSS are discussed in Chapter 3, 4, 5, 6 and 7, respectively. The evaluations provided in this chapter are based on the design features and evaluations described therein.

Off-normal events for the DSS and the CLU are discussed in Chapter 12.1. According to Chap-ter 2, accident conditions during canister handling via CLU are not to be assumed. The evaluations provided in Chapter 12.2 thus exclude the CLU components.

12.0 Overview Section 12.0, Rev. 1 Page 12.0-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 12.1 Off-Normal Conditions Prepared Reviewed 12.1 Off-Normal Conditions Section 12.1, Rev.1 Page 12.1-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Off-normal conditions, as defined in accordance with ANSI/ANS-57.9 [1], are those conditions, which, although not occurring regularly, are expected to occur no more than once a year. In this section, expected off-normal conditions according to Section 2.2 are considered.

The results of the evaluations demonstrate that the CASTOR geo69 DSS and the CLU both with-stands the effects of off-normal conditions without affecting any of its safety functions. The follow-ing subsections present the evaluations of DSS and CLU for the design basis off-normal condi-tions. Based on these evaluations, it is concluded that off-normal conditions do not affect the safe operation of CASTOR geo69 DSS and CLU. Compliance with the requirements of 10 CFR 72.122, 10 CFR 72.104(a) and 10 CFR 20 is demonstrated.

The following off-normal conditions regarding the handling and storage operations of DSS and CLU are discussed in the subsequent Subsections:

• Off-normal pressure,

• Off-normal temperature;

• Off-normal handling of storage cask,

• Malfunction of the pressure monitoring system and

• Off-normal handling of CLU during short-term operations 12.1.1 Off-normal Pressure 12.1.1.1 Postulated Cause of Event The off-normal pressure in the canister bounds the cumulative effects of the maximum fill gas vol-ume, off-normal environmental temperatures, the maximum SNF heat load, and an assumed 10 %

of the fuel rods ruptured with 100 % of the fill gas and 15 % of the fission gases released due to a cladding breach in accordance with NUREG-2224 [2] and is thus slightly increased compared to the NCS pressure.

After FA loading, the canister is drained, dried, and backfilled with helium to assure long-term fuel cladding integrity during dry storage. Therefore, the probability of failure of intact fuel rods during dry storage is low. Nonetheless, the event is postulated and evaluated.

The off-normal pressure in the storage cask is slightly higher compared to the NCS pressure ac-cording to Section 7.3.

The off-normal pressure event regarding canister or storage cask does not distinguish handling with the DSS, the CLU or DSS storage operations.

12.1 Off-Nom1al Conditions Section 12.1, Rev. 1 Page 12.1-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 12.1.1.2 Detection of Event Storage cask and canister of the CASTOR geo69 DSS are both designed to withstand off-normal internal pressure without any effects on their ability to meet the safety requirements. There is no requirement for detection of off-normal pressure. Therefore, no monitoring is required.

12.1.1.3 Analysis of Effects and Consequences The off-normal internal pressure of the canister and the corresponding boundary conditions are reported in Section 7.3. The applied pressure values for storage cask and canister are summarized in the Appendix 3-1 in Section 3.10.

Structural The structural evaluation of the DSS for off-normal internal pressure conditions is discussed in Section 3.5. Internal overpressure acts in hot temperature conditions, while external overpressure is considered in cold temperature conditions. The stress results show that the DSS structure can withstand the applied loadings due to off-normal pressure during all handling and storage opera-tions.

Thermal The temperatures under off-normal pressure conditions are evaluated in Section 4.5 and for a fail-ure of 10 % of the fuel rods after 20 years of storage in Section 4.8. The evaluation of the results show that all calculated maximum temperatures of the DSS components and the content are far below the maximum admissible values with large safety margins. The analysis are also covering the short-term operations with the CLU and the on-site transfer.

Shielding Off-normal pressure conditions have no effect on the shielding performance of the DSS and CLU during all handling and storage operations.

Criticality Off-normal pressure conditions have no effect on the criticality control of the content inside the can-ister during all handling and storage operations.

Containment Off-normal pressure conditions have no effect on the containment function of canister and storage cask during all handling and storage operations. As discussed in the structural evaluation above, all stresses remain within allowable values, assuring containment boundary integrity.

12.1 Off-Normal Conditions Section 12.1, Rev. 1 Page 12.1-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Radiation Protection Since there is no degradation in shielding or containment capabilities as discussed above, off-normal pressure conditions have no effect on occupational or public exposures.

12.1.1.4 Corrective Actions No corrective action is required for off-normal pressure conditions.

12.1.1.5 Radiological Impact No radiological impact has to be assumed because the integrity of containment and shielding is not compromised.

12.1.2 Off-normal Temperature 12.1.2.1 Postulated Cause of Event Off-normal temperatures conditions are a consequence of extreme seasonal variations of the envi-ronmental temperatures. Both the on-site transfer of the storage cask and the interim storage of the DSS take place outdoors and unsheltered and are thus directly influenced by environmental temperatures and solar insolation. Off-normal temperature during CLU handling operations inside the reactor building is not a credible event. Moreover, it is assumed that they are covered by off-normal temperature evaluations for the DSS.

12.1.2.2 Detection of Event The CASTOR geo69 DSS is designed to withstand the off-normal environmental temperatures without any effects on its ability to maintain safe storage conditions. There is no requirement for detection of off-normal environmental temperatures.

12.1.2.3 Analysis of Effects and Consequences Structural The structural evaluation of the canister for off-normal conditions of storage in hot and cold condi-tion is provided in Chapter 3. The off-normal conditions of storage provides for the DSS compo-nents a thermal equilibrium at their respective maximum temperatures for hot conditions and at

-29 °C for cold conditions. Thermal stresses occur due to differential thermal expansion. The stress results indicate that the DSS structure withstands the applied loadings due to off-normal temperatures. The evaluations are covering the on-site transfer.

12.1 Off-Normal Conditions Section 12.1, Rev. 1 Page 12.1-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Thermal For the hot case an off-normal ambient temperature o f - is used in conjunction with solar inso-lation according to 10 CFR 71.71 (see Section 4.5). For the cold case -40°C and solar flux of zero, the NCS evaluations are bounding. It is demonstrated in Section 4.5 that the CASTOR geo69 DSS fulfils all requirements for off-normal temperature conditions with regard to thermal aspects.

Off-normal temperatures as a consequence of 10 % of the fuel rods failure with release of 100 % of the fill gas 15 % of the fission gases after 20 years of storage are also calculated. The maximum temperatures of the DSS components and the content are far below the maximum admissible val-ues with large safety margins.

It is demonstrated that the evaluations cover also the on-site transfer.

Shielding Off-normal temperature conditions have no effect on the shielding performance of the DSS during all handling and storage operations.

Criticality Off-normal temperature conditions have no effect on the criticality control of the content inside the canister during all handling and storage operations.

Containment Off-normal temperature conditions have no effect on the containment function of the canister and storage cask during all handling and storage operations. As discussed in the structural evaluation above, all stresses remain within allowable values, assuring containment boundary integrity.

Radiation Protection Since there is no degradation in shielding or containment capabilities as discussed above, off-normal temperature conditions have no effect on occupational or public exposures.

12.1.2.4 Corrective Actions No corrective action is required for off-normal temperature conditions.

12.1.2.5 Radiological Impact No radiological impact has to be assumed because the integrity of containment and shielding is not compromised.

12.1 Off-Normal Conditions Section 12.1, Rev. 1 Page 12.1-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 12.1.3 Off-normal Handling of Storage Cask 12.1.3.1 Postulated Cause of Event During the routine handling operations as discussed in Section 2.2, several off-normal handling events may be caused by mishandling or simple negligence of operators, equipment malfunction and failure or loss of power for a limited duration. Although such events are very unlikely, off-normal handling events are discussed in the following subsections.

12.1.3.2 Detection of Event Handling of the storage cask at the NPP, during the on-site transfer or at the ISFSI site is per-formed under the supervision of qualified personnel. Therefore, the attending personnel immedi-ately detects off-normal handling events. There is no requirement for detection of off-normal han-dling events.

12.1.3.3 Analysis of Effects and Consequences Off-normal handling conditions can lead to unintended mechanical loads. Due to the inertia of the storage cask, mishandling via crane (e.g. abrupt crane stop or impermissible crane movement) may lead to a slight contact between the storage cask and the service platform or other surround-ing structures of the NPP or the ISFSI. A light bumping of the handled storage cask with another CASTOR geo69 DSS located at the ISFSI is also a credible event. A temporary loss of power can lead to an abrupt stop of crane movement. Furthermore, unplanned braking during the on-site transfer cannot be completely ruled out. Drop events resulting from handling accidents during han-dling operations performed on the storage cask are considered non-credible events since the CASTOR geo69 DSS is designed for handling via single-failure proof handling devices.

Structural The trunnions of the storage cask are special lifting devices under consideration of the provisions of ANSI N14.6 [3]. Therefore, the loads induced during off-normal handling via crane are not ex-pected to cause inadmissible stresses in the trunnions. Collisions with surrounding structures or with other storage casks located at the ISFSI are bounded by the accidents analysed in Sec-tion 3.6. The tip-over of the storage cask and the tornado missile impacts analysed lead to higher mechanical loads compared to off-normal handling events due to the limited crane movement speed. Loads resulting from unplanned breaking during on-site transfer are of minor importance, since the transfer takes place at walking pace.

12.1 Off-Normal Conditions Section 12.1, Rev. 1 Page 12.1-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Thermal Off-normal handling conditions have no effect on the heat removal performance of the storage cask.

Shielding Off-normal handling conditions have no effect on the shielding performance of the storage cask.

Criticality Off-normal handling of the loaded storage cask has no effect on the criticality control of the SNF in the canister.

Containment Off-normal handling conditions have no effect on the containment function of the storage cask.

Radiation Protection Since there is no degradation in shielding or containment capabilities as discussed above, off-normal handling conditions have no effect on occupational or public exposures.

12.1.3.4 Corrective Actions Collisions that occur as a consequence of off-normal handling operations require an interruption of the performed handling operation and an inspection of the involved components for possible dam-age. In the event of a power failure, the power supply must be restored.

12.1.3.5 Radiological Impact No radiological impact has to be assumed because the integrity of containment and shielding is not compromised.

12.1.4 Malfunction of the Pressure Monitoring System 12.1.4.1 Postulated Cause of Event During the NCS as discussed in Section 2.2, a malfunction of the pressure monitoring system may occur resulting from either an equipment failure (especially the pressure switch) or a loss of power for a limited duration.

12.1.4.2 Detection of Event A malfunction of the pressure monitoring system will lead to a pressure switch alert as discussed in Section 1.2.

12.1 Off-Normal Conditions Section 12.1, Rev. 1 Page 12.1-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 12.1.4.3 Analysis of Effects and consequences Structural A malfunction of the pressure monitoring system has no effect on the structural performance of the storage cask.

Thermal A malfunction of the pressure monitoring system has no effect on the heat removal performance of the storage cask.

Shielding A malfunction of the pressure monitoring system has no effect on the shielding performance of the storage cask.

Criticality A malfunction of the pressure monitoring system has no effect on the criticality control of the SNF inside the storage cask.

Containment A malfunction of the pressure monitoring system has no effect on the containment function of the storage cask.

Radiation Protection Since there is no degradation in shielding or containment capabilities as discussed above, a mal-function of the pressure switch has no effect on occupational or public exposures.

12.1.4.4 Corrective Actions In case of a defect, the pressure switch needs to be replaced. In the event of a power failure, the power supply must be restored.

12.1.4.5 Radiological Impact No radiological impact has to be assumed because the integrity of containment and shielding is not compromised.

12.1 Off-Normal Conditions Section 12.1, Rev. 1 Page 12.1-8

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 12.1.5 Off-normal Handling of CLU (short-term operations) 12.1.5.1 Postulated Cause of Event During the routine handling operations as discussed in Section 2.2, several off-normal handling events may be caused by mishandling or simple negligence of operators, equipment malfunction or failure, or loss of power for a limited duration. Although _such off-normal handling events are very unlikely, they are discussed in the following subsections.

12.1.5.2 Detection of Event CLU handling operations in the NPP are performed under the supervision of qualified personnel.

Therefore, the attending personnel immediately detects off-normal handling events. There is no requirement for detection of off-normal handling events.

12.1.5.3 Analysis of Effects and consequences Off-normal handling conditions can lead to unintended mechanical loads. Due to the inertia of the canister and the transfer cask, mishandling via crane (e.g. abrupt crane stop or impermissible crane movement) may lead to a slight contact between the transfer cask and e.g. the service plat-form or any other surrounding structure of the reactor building, or between the canister and parts of the CLU when setting up the canister transhipment configuration. Moreover, a temporary loss of power can lead to an abrupt stop of crane movement or to the temporary failure of the dewatering, drying or He-equipment during canister dispatch. A malfunction of the transfer lock may lead to the jamming of the canister or bottom lid of transfer cask during canister transhipment.

Structural The LAP on the canister lid and the trunnions of the transfer cask are special lifting devices under consideration of the provisions of ANSI N14.6 [3]. Therefore, the loads induced during off-normal handling via crane are not expected to cause inadmissible stresses in the LAP of canister or trans-fer cask. Due to the limited crane movement velocity, minor collisions of the canister during tran-shipment and the related loads induced into the canister are covered by the ACS load case tipping of the storage cask (with canister inside) in Chapter 3 and Section 12.2. During canister tranship-ment, jamming of the transfer cask bottom lid in the transfer lock and the canister in the transfer cask may occur due to e.g. a malfunction of the transfer lock. Since the inner surface of the trans-fer cask is smooth and does not exhibit edges or undercuts, the canister can always be lifted up-wards in case of a transfer lock malfunction.

Collisions between the canister and the transfer cask during canister transhipment are not ex-pected to cause significant damage to the transfer cask due to the ** thick steel liner in the 12.1 Off-Normal Conditions Section 12.1, Rev. 1 Page 12.1-9

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 transfer cask. Since the crane movement speed and the space for the canister to move in horizon-tal orientation is limited, considerable impact forces cannot occur. With regard to the outer shell of the transfer cask, the bottom ring is a protruding component compared to the cylindrical part of the transfer cask body. Therefore, collisions due to an oscillation movement of the transfer cask during handling via crane will only occur between the bottom ring and surrounding structures in the NPP.

Since the bottom ring is a massive steel component and the crane movement speed is very limited, a damage of the water shielding of the transfer cask is considered a non-credible event.

Thermal The environmental temperatures during off-normal handling condition are bounded by NCS, since the pool water and the reactor hall temperatures are chosen conservatively high. Off-normal events during handling of the CLU have no impact on the heat removal capability. The temporary failure of the dewatering, drying or He-equipment during canister dispatch because of power loss can lead to an increase of temperature in the canister as described in Section 4.7. During vacuum drying and after helium backfilling, the calculated maximum temperatures of the radioactive content and the components are far below the maximum admissible values with large safety margins. This applies for an unlimited time in final steady state after heating up to thermal equilibrium.

Shielding Off-normal handling conditions have no effect on the shielding performance of the canister. With regard to the transfer cask, the degradation of the shielding performance due to water leakage from the outer water chamber after a collision with the surrounding structures of the NPP is con-sidered a non-credible event.

Criticality Off-normal handling conditions have no effect on the criticality control of the canister.

Containment Off-normal handling conditions have no effect on the containment function of the canister. As dis-cussed in the structural evaluation above, impermissible stresses are not expected, assuring con-tainment boundary integrity.

Radiation Protection There is no degradation in the shielding or containment capabilities of the canister. Degradation of the shieling capabilities of the transfer cask can only occur as a consequence of a collision and subsequent leakage of water from the outer water chamber. However, the loss of a considerable amount of water from the water chamber of the transfer cask can be prevented by immediate cor-rective actions.

12.1 Off-Normal Conditions Section 12.1, Rev. 1 Page 12.1-10

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 12.1.5.4 Corrective Actions Collisions that occur as a consequence of off-normal handling operations require an interruption of the performed handling operation and an inspection of the involved components for possible dam-age. Handling operations may be continued after the integrity of all components has been verified.

In case of a malfunction of the transfer lock during canister transshipment, the canister shall be lifted up to allow for the inspection and repair of the transfer lock. If necessary, canister and trans-fer cask shall be moved back to the service platform to improve the accessibility of the transfer lock for repair.

Since inadmissible temperatures during canister dispatch cannot be reached as calculated in Sec-tion 4.7, corrective actions are not required during a power outage. After the power outage, canis-ter dispatch can be continued.

12.1.5.5 Radiological Impact No radiological impact has to be assumed because the integrity of containment and shielding is not compromised.

List of References

[1] ANSI/ANS-57.9-1992 Design Criteria For An Independent Spent Fuel Storage Installation (Dry Type)

American National Standards Institute

[2] NUREG-2224, November 2020 Dry Storage and Transportation of High Burnup Spent Nuclear Fuel - Final Report U.S. Nuclear Regulatory Commission

[3] ANSI N14.6 (1993)

Radioactive Materials - Special Lifting Devices for Shipping Containers Weighing 10 000 Pounds (4500 kg) or More 12.1 Off-Normal Conditions Section 12.1, Rev. 1 Page 12.1-11

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 12.2 Accidents Prepared Reviewed 12.2 Accidents Section 12.2, Rev. 1 Page 12.2-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Accidents, in accordance with ANSI/ANS-57.9 [1], are either infrequent events that could reasona-bly be expected to occur during the lifetime of the CASTOR geo69 DSS or events postulated be-cause their consequences may affect the public health and safety. Section 2.2 defines the design basis accidents considered. By analysing for these design basis events, safety margins inherently provided in the CASTOR geo69 DSS design can be quantified.

The results of the evaluations performed herein demonstrate that the CASTOR geo69 DSS with-stands the effects of all credible and hypothetical accident conditions and natural phenomena with-out affecting its safety function. The following subsections present the evaluation of the following design basis accident and natural phenomena:

• Fire

• Tip-over

• Earthquake

• Tornado

• Flood

• Explosion

• Lightning

• Burialunderdebris

• 100 % fuel rod failure Handling accidents are not assumed for the DSS and hence not evaluated as DSS handling re-quires the usage of single failure crane installations.

Since all handling operations with the CLU are limited to take place inside the reactor building and by usage of single failure crane installations neither natural phenomena nor accidents are to be considered credible events for the CLU.

The evaluations demonstrate that the requirements of 10 CFR 72.122 are satisfied, and that the corresponding radiation doses satisfy the requirements of 10 CFR 72.106(b) and 10 CFR 20.

12.2.1 Fire Accident 12.2.1.1 Cause of Fire Accident The possibility of a fire accident at or near the storage site is considered to be extremely remote due to the absence of significant combustible materials. The only credible source for a fire is the fuel tank of the transport vehicle, which is used to move the storage cask to the storage position at 12.2 Accidents Section 12.2, Rev. 1 Page 12.2-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS the ISFSI. The fire accident is conservatively postulated to be the result of the spillage and ignition of 200 liters of combustible transporter fuel either during on-site transfer or during long-term interim dry storage.

12.2.1.2 Fire Accident Analysis To demonstrate the fuel cladding and containment boundary integrity under an exposure to a hypo-thetical short duration fire event during vertical storage, a fire accident analysis of the loaded CASTOR geo69 DSS is performed. During the postulated fire accident, the DSS is completely engulfed by an 800 °C hot fire. The surface of the DSS receives an incident radiation and forced convection heat flux from the fire. The same fire is applied to the storage cask during the horizontal on-site transfer.

Structural The structural integrity and the redundant containments of the CASTOR geo69 DSS are not im-paired by the fire accident (see Section 3.6 for ACS and Section 3.8 for the on-site transfer). The resulting thermal stresses are within the allowable values.

Thermal The fire duration is approximately 5 minutes according to Section 4.6 (ACS) and Section 4.7 (on-site transfer). Due to the thermal inertia of the DSS, many components reach the maximum tem-perature during the subsequent cooling phase. The evaluation of the results show that all calculat-ed maximum temperatures of DSS components and content are far below the maximum admissi-ble values for ACS with large safety margins.

Shielding

- The assumed loss of all moderator material in the CASTOR geo69 DSS results in an increase in the radiation dose rates. However, the shielding evaluation presented in Chapter 5 demonstrates that the requirements of 10 CFR 72.106 are not exceeded at any time.

Criticality The fire accident has no effect on the criticality control of the content inside the canister.

Containment The fire accident has no effect on the containment function of the DSS. Both containment barriers remain leak-tight. The internal pressure in the canister and the cask cavity is reported in Sec-tion 7.4. The temperatures of the components of the containment boundary do not exceed the short-term allowable temperature limits (see Section 4.3).

12.2 Accidents Section 12.2, Rev. 1 Page 12.2-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Radiation Protection There is no degradation in containment of the CASTOR geo69 DSS, as discussed above. In-crease in the dose rate due to the loss of moderator material is evaluated in Chapter 5.

12.2.1.3 Fire Accident Dose Calculations The complete loss of the CASTOR geo69 DSS neutron moderator (UHMW PE rods and plates) material is assumed in the shielding analysis for the post-accident analysis of the loaded storage cask in Chapter 5 and bounds the determined fire accident consequences. The loaded CASTOR geo69 storage cask following a fire accident meets the accident dose rate requirement of 10 CFR 72.106(b). The 30 days-dose remains safely under the dose limit o f * * * * *

• dis-tance 12.2.1.4 Fire Accident Corrective Actions Upon detection of a fire adjacent to a loaded CASTOR geo69 DSS, the ISFSI operator shall take the appropriate immediate actions necessary to extinguish the fire. Firefighting personnel should take appropriate radiological precautions. Following the termination of the fire, a visual and radio-logical inspection of the equipment shall be performed. As appropriate, temporary shielding shall be installed around the DSS.

If damage to the CASTOR geo69 DSS as the result of a fire event is widespread and/or as radio-logical conditions require, the canister shall be removed from the storage cask in accordance with the procedure specified in Section 9.2. However, the thermal analysis described herein demon-strates that the DSS components remain below the accident temperature limits. The DSS may be returned to service if there is no significant increase in the measured dose rates and if the visual

- inspection is satisfactory.

12.2.2 Tip-Over 12.2.2.1 Cause of Tip-Over The analysis of the CASTOR geo69 DSS has shown that the cask does not tip over as a result of the postulated and analysed design basis accidents. It is highly unlikely that the storage cask will tip-over during handling operations because of the required use of lifting devices designed in ac-cordance with ANSI N14.6 [2] and NUREG-0612 [3] as specified in Section 2.0. Tip-over (roll off) of the horizontally positioned storage cask from the transfer-vehicle during on-site transfer is also unlikely. The tip-over accident is stipulated as a non-mechanistic accident.

12.2 Accidents Section 12.2, Rev. 1 Page 12.2-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 (_@)GNS 12.2.2.2 Tip-Over Analysis The tip-over accident analysis evaluates the effects of the loaded DSS tipping-over onto a rein-forced concrete pad. The tip-over analysis is provided in Chapter 3.

Structural A storage cask drop analysis with the longitudinal axis horizontal (side drop) together with a stor-age cask drop analysis with the longitudinal axis vertical (bottom end drop) are assumed as a bounding accident combination for the non-mechanistic tip-over analysis in Section 3.6. The struc-tural evaluation demonstrates that under the calculated - loading the stresses are within the allowable values. The analysis also bounds the storage cask roll off from the on-site transfer vehi-cle (Section 3.8).

Thermal The thermal analysis of the CASTOR geo69 DSS is based on the vertical storage configuration.

The thermal consequences after the tip-over, when the DSS is in the horizontal orientation, are bounded by the thermal analysis for fuel rod failure for ACS impact in Section 4.6, which considers a maximum amount of fission gas and leads to maximum temperatures in the canister. The evalua-tion of all results for ACS impact (scenario I and scenario II as specified in Section 4.6) show that the calculated maximum temperatures of DSS components and content are far below the maxi-mum admissible values with large safety margins. The analysis also bounds the storage cask roll off from the on-site transfer vehicle (Section 4. 7).

Shielding The bottom of the storage cask, which is normally facing the ground, will face the horizon after the tip-over. This orientation does not lead to dose rates exceeding the total effective dose limits speci-e fied in 10 CFR 72.106. Containment and shielding material are not impaired by the storage cask tip-over. The storage cask roll off from the transfer vehicle does also not impact the shielding per-formance.

Criticality The tip-over accident has no effect on the criticality control of the DSS. The displacement of the loaded FA as a result of the horizontal orientation of the DSS after tip-over is bounded by the mod-el for criticality evaluation in Chapter 6. The same is valid for the storage cask roll off from the transfer vehicle.

12.2 Accidents Section 12.2, Rev. 1 Page 12.2-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Containment The tip-over/roll off accident has no effect on the containment function of the DSS. As discussed in the structural evaluation above, all stresses remain within allowable values, which suggests con-tainment boundary integrity.

Radiation Protection There is no degradation in containment of the CASTOR geo69 DSS as a result of the tip-over/roll off accident. Based on a minimum distance to the controlled area boundary of ****l the 30 days-dose at the controlled area boundary (see Section 5.1) does not exceed the 10 CFR 72.106 dose requirements, even for a complete loss of the moderator material.

12.2.2.3 Tip-Over Dose Calculations

- I The analysis of the tip-over/roll off accident has shown that the DSS containment barriers will not be compromised and, therefore, there will be no release of radioactivity or increase in site-boundary dose rates from release of radioactivity.

The tip-over/roll off accident could cause localized minor damage to the outer shell of the storage cask (e.g. cooling fins). However, due to the massive design of the DCI cask body, containment and shielding material are not impaired. There will be no significant increase in the on-site or the boundary dose rate as a result of the localized damage.

12.2.2.4 Tip-Over Corrective Actions Following a tip-over/roll off accident, the ISFSI operator shall first perform a radiological and visual inspection to determine the extent of the damage to the DSS. Special handling procedures, includ-ing the use of temporary shielding, will be developed to restore the vertical orientation of the DSS.

Likewise, the CASTOR geo69 DSS shall be thoroughly inspected and a determination shall be made if repairs are required and will enable the DSS to return to service. Subsequent to the re-pairs, the equipment shall be inspected and appropriate tests shall be performed to certify the DSS for service. If the equipment cannot be repaired and returned to service, the equipment shall be disposed of in accordance with the appropriate regulations.

12.2.3 Earthquake 12.2.3.1 Cause of Earthquake It is possible that, during the on-site transfer or the intended storage period of the CASTOR geo69 DSS, the ISFSI may experiences an earthquake.

12.2 Accidents Section 12.2, Rev. 1 Page 12.2-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 12.2.3.2 Earthquake Analysis The earthquake accident analysis evaluates the effects of a seismic event on the loaded CASTOR geo69 DSS. The objective is to determine the stability limits of the DSS. Based on a static stability criteria, it is shown in Chapter 3 that the CASTOR geo69 DSS is qualified to seis-mic activity less than or equal to the values specified in Section 13.2. The analyses in Chapter 3 show that the CASTOR geo69 DSS will not tip over under these conditions and that incipient slid-ing occurs before incipient tipping. The space between neighbouring storage casks prevents pos-sible collisions.

Some storage sites may have design basis earthquakes that exceed the seismic activity limits specified in Section 13.2. These high-seismic sites require an additional evaluation of the DSS design with regard to the structural consequences of such an earthquake. Additional safety measures (e.g. anchoring of the DSS on the storage pad) must be implemented to prevent a tip-over of the DSS as a result of the earthquake.

Structural The sole structural effect of the earthquake is an inertial loading of less than

• This loading is bounded by the non-mechanistic tip-over analysis presented in Chapter 3 by a covering cask drop analysis. The calculated steady state deceleration of the CASTOR geo69 DSS is - The stress results indicate that the DSS structure can withstand the applied impact loadings.

When a seismic event occurs during the on-site transfer of the storage cask, the design of the load restraint shall ensure that the storage cask detaches from the transfer vehicle, providing that that storage cask can roll off the vehicle unimpededly and intact. The roll off is also covered by the non-mechanistic tip-over analysis.

Thermal The earthquake has no effect on the thermal performance of the DSS.

Shielding The earthquake has no effect on the shielding performance of the DSS.

Criticality The earthquake has no effect on the criticality control features of the DSS.

Containment The earthquake has no effect on the containment function of the DSS.

12.2 Accidents Section 12.2, Rev. 1 Page 12.2-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Radiation Protection Since there is no degradation in shielding or containment capabilities as discussed above, there is no effect on occupational or public exposures as a result of the earthquake.

12.2.3.3 Earthquake Dose Calculation The structural analysis of the earthquake accident shows that the loaded DSS will not tip over as a result of the specified seismic activity, hence, there is no increase in radiation dose rates or release of radioactivity to be expected.

12.2.3.4 Earthquake Corrective Actions Following the earthquake accident, the ISFSI operator shall perform a visual and radiological in-spection of the DSS to determine if any of the storage casks have slipped or tipped-over. In the unlikely event and against all expectations of a tip-over, the corrective actions shall be in accord-ance with Subsection 12.2.2.4.

12.2.4 Tornado 12.2.4.1 Cause of Tornado The CASTOR geo69 DSS may be stored at an unsheltered ISFSI site that is subject to environ-mental conditions. Therefore, it is possible that the DSS will experience the extreme environmental conditions of a tornado during storage. Additionally, the storage cask may experience the environ-mental conditions of a tornado during on-site transfer.

12.2.4.2 Tornado Analysis The tornado accident has two effects on the CASTOR geo69 DSS. The tornado winds and/or tor-nado missile attempt to tip-over the loaded storage cask. The pressure loading of the high velocity winds and/or the impact of the large tornado missiles act to apply an overturning moment. The second effect is tornado missiles propelled by high velocity winds, which attempt to penetrate the storage cask.

Structural Section 3.6 provides the analysis of the tornado loading which attempts to tip-over the DSS and the analysis of the effects of the different types of tornado missiles. These analyses show that the loaded DSS does not tip-over as a result of the tornado winds and/or tornado missiles. The tornado wind does not lead to a sliding of the loaded DSS. In addition, the analyses provided in Section 3.6 12.2 Accidents Section 12.2, Rev. 1 Page 12.2-8

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 also shows that the tornado missiles do not penetrate the storage cask. The tornado missile im-pacts analysed do not lead to a punch through of the storage cask.

When a tornado event occurs during the on-site transfer of the storage cask, the design of the load restraint shall ensure that the storage cask detaches from the transfer vehicle providing that that storage cask can roll off the vehicle unimpededly and intact. The roll off is covered by the non-mechanistic tip-over analysis.

Thermal The tornado has no effect on the thermal performance of the DSS.

Shielding The tornado has no effect on the shielding performance of the DSS.

Criticality The tornado has no effect on the criticality control features of the DSS.

Containment The tornado has no effect on the containment function of the DSS. Since the tornado missiles do not lead to a punch through of the storage cask, the containment of the storage cask remains in-tact.

Radiation Protection Since there is no degradation in shielding or containment capabilities as discussed above, there is no effect on occupational or public exposures as a result of a tornado.

12.2.4.3 Tornado Dose Calculation The structural analysis of tornado wind and tornado missile impact shows that the loaded DSS will not tip-over as a result of the specified tornado and not punch through of the storage cask occurs.

Hence, there is no increase in radiation dose rates or release of radioactivity to be expected. A tornado missile may cause localized damage in the cask body of the storage cask. However, the damage will have a negligible effect on the site boundary dose.

12.2.4.4 Tornado Corrective Actions Following the tornado accident, the ISFSI operator shall perform a visual and radiological inspec-tion of the DSS to determine if any of the storage casks have slipped or tipped-over. In the unlikely event and against all expectations of a tip-over, the corrective actions shall be in accordance with Subsection 12.2.2.4. Damage sustained by the storage cask shall be inspected and repaired.

12.2 Accidents Section 12.2, Rev. 1 Page 12.2-9

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 12.2.5 Flood 12.2.5.1 Cause of Flood The CASTOR geo69 DSS may be stored at an unsheltered ISFSI site. Therefore, it is possible for the storage area to be flooded. The potential sources for the floodwater could be unusually high water from a river or stream, a dam break, a seismic event, or high tides from tornados. On-site transfers are not permitted during a flood.

12.2.5.2 Flood Analysis A flood with a given velocity will apply a uniformly distributed pressure onto the DSS over a pro-jected area in flow direction.

Structural The flood accident affects the CASTOR geo69 DSS by applying an overturning moment, which attempts to tip-over the loaded DSS. Section 3.6 provides the analysis of the floodwater applying an overturning moment. The results analysis show that the loaded DSS does not tip over if the flood velocity does not exceed 5 mis. Incipient sliding of the DSS also does not occur.

Thermal The flood has no effect on the thermal performance of the DSS.

Shielding The flood has no effect on the shielding performance of the DSS.

Criticality The flood has no effect on the criticality control features of the DSS. The criticality analysis is unaf-fected because under the flooding condition water does not enter the storage cask or the canister cavity.

Containment The flood has no effect on the containment function of the DSS.

Radiation Protection Since there is no degradation in shielding or confinement capabilities as discussed above, there is no effect on occupational or public exposures because of this event. Flood Dose Calculation Since the flood accident produces no leakage of radioactive material and no reduction in shielding effectiveness, there are no adverse radiological consequences. The floodwater provides additional shielding that reduce radiation doses.

12.2 Accidents Section 12.2, Rev. 1 Page 12.2-10

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @)GNS 12.2.5.3 Flood Corrective Actions As shown in the structural analysis of the flood accident, the CASTOR geo69 DSS sustains no damage due to the flood accident. At the completion of the flood, exposed surfaces may need de-bris and adherent foreign matter removal.

12.2.6 Explosion 12.2.6.1 Cause of Explosion Explosions within the boundary of the facility is rather improbable since there are no explosive ma-terials within the site boundary. The only reasonable source for an explosion is a fuel tank of a transport vehicle. Explosions may also occur because of an industrial accident in the vicinity of the Dry storage facility or along the on-site transfer path of the DSS. As the fuel available for the ex-plosion is limited in quantity, the effects of an explosion on the DSS are minimal.

12.2.6.2 Explosion Analysis Any credible explosion accident is bounded by the accident external pressure analysed as a result of the enhanced water immersion test, because explosive materials are not stored within close proximity to the DSS. The DSS can withstand the effects of substantial accident external pressures without damage.

Structural The structural evaluations for the DSS accident condition external pressure are presented in Sec-tion 3.6 and demonstrate that all stresses are within allowable values.

Thermal The explosion has no effect on the thermal performance of the DSS.

Shielding The explosion has no effect on the shielding performance of the DSS.

Criticality The explosion has no effect on the criticality control features of the DSS.

Containment The explosion has no effect on the containment function of the DSS.

12.2 Accidents Section 12.2, Rev. 1 Page 12.2-11

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Radiation Protection Since there is no degradation in shielding or confinement capabilities as discussed above, there is no effect on occupational or public exposures because of this event.

12.2.6.3 Explosion Dose Calculation The bounding external pressure load has no effect on the CASTOR geo69 DSS. Therefore, no effect on the shielding, criticality, thermal or confinement capabilities of the DSS is experienced because of the explosion pressure load. The effects of explosion-generated projectiles on the DSS are bounded by the analysis of tornado-generated missiles.

12.2.6.4 Explosion Corrective Actions Following an explosion, the ISFSI operator shall perform a visual and radiological inspection of the DSS. Damage sustained by the storage cask due to explosion-generated projectiles shall be in-spected and repaired.

12.2.7 Lightning 12.2.7.1 Cause of Lightning The CASTOR geo69 DSS is permitted to be stored at an unsheltered ISFSI site. There is the po-tential for lightning to strike the DSS.

12.2.7.2 Lightning Analysis The CASTOR geo69 DSS consists of a large metal cask stored at an unsheltered ISFSI. When the DSS is hit by lightning, the lightning will discharge through the DCI cask body of the storage cask to the ground. Lightning strikes have high currents, but their duration is short (i.e., less than a second). The cask body consists of conductive ductile cast iron and will provide a direct path to the ground. The canister is unaffected by the lightning.

Structural The lightning has no effect on the structural integrity of the DSS.

Thermal The lightning has no effect on the thermal performance of the DSS.

Shielding The lightning has no effect on the shielding performance of the DSS.

12.2 Accidents Section 12.2, Rev. 1 Page 12.2-12

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Criticality The lightning has no effect on the criticality control features of the DSS.

Containment The lightning has no effect on the containment function of the DSS.

Radiation Protection Since there is no degradation in shielding or confinement capabilities as discussed above, there is no effect on occupational or public exposures because of this event.

12.2.7.3 Lightning Dose Calculation The effect of a lightning strike has no effect on the confinement boundary or shielding materials of the DSS. Therefore, no further analysis is necessary.

12.2.7.4 Lightning Corrective Actions The CASTOR geo69 DSS will not sustain any damage from the lightning accident. There is no surveillance or corrective action required.

12.2.8 Burial under Debris 12.2.8.1 Cause of Burial under Debris Since the CASTOR geo69 DSS is placed outdoors on an unsheltered reinforced concrete storage pad at an ISFSI site, burial under debris may be a result of storms, mud slides or flooding. Debris from a collapsing building is not a credible event since there are no structures over the DSS. There is no credible mechanism for the DSS become completely buried under debris. However, for con-servatism, complete burial under debris is considered.

12.2.8.2 Burial under Debris Analysis The structural loads due to burial under debris are covered by other load cases evaluated in Sec-tion 3.6. The thermal effects of burial under debris on the DSS are evaluated in Section 4.6. Burial under debris affects thermal performance because the debris acts as an insulator and heat sink.

The evaluation for burial under debris comprises an initial steady state and a subsequent transient burial phase, when the heat removal at the entire outer cask surface conservatively is completely eliminated. For burial under debris, a failure of 100 % of the fuel rods in conjunction with a fraction of fission gas release from the fuel rods of 15 % is assumed.

12.2 Accidents Section 12.2, Rev. 1 Page 12.2-13

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Structural The burial under debris has no effect on the structural integrity of the DSS.

Thermal The initial conditions before the burial accident correspond to the steady state for NCS described in Section 4.4. A complete burial of the DSS for example by debris or sludge is considered. As a hy-pothetic limit case, it is conservatively assumed that the heat removal is completely eliminated at the entire outer surface, which leads to a continuous increase of temperature. Because of the high thermal inertia of the loaded DSS, the temperatures only rise very slowly at a nearly constant rate of about I For the design-relevant components, the time until the maximum admissi-ble temperature is reached is I (see Section 4.6).

Shielding The burial under debris has no effect on the shielding performance of the DSS.

Criticality The burial under debris has no effect on the criticality control features of the DSS.

Containment The burial under debris has no effect on the containment function of the DSS.

Radiation Protection Since there is no degradation in shielding or confinement capabilities as discussed above, there is no effect on occupational or public exposures because of this event.

Based on this evaluation, it is concluded that the burial under debris accident does not affect the safe operation of the CASTOR geo69 DSS, if the debris is removed within 3 days. The 3-days minimum inspection interval ensures that a burial under debris condition will be detected long be-fore the allowable burial time is reached.

12.2.8.3 Burial under Debris Dose Calculation The shielding is enhanced while the CASTOR geo69 DSS is buried. The elevated temperatures will not cause the breach of the confinement system and the short-term fuel cladding temperature limit is not exceeded. Therefore, there is no radiological impact.

12.2.8.4 Burial under Debris Corrective Actions Analysis of the burial under debris accident in Section 4.6 shows that the fuel cladding peak tem-peratures are not exceeded even for an extended duration of several days of burial. Upon detec-12.2 Accidents Section 12.2, Rev. 1 Page 12.2-14

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 tion of the burial under debris accident, the ISFSI operator shall assign personnel to remove the debris with mechanical and manual means as necessary. After uncovering the DSS, the storage cask shall be visually and radiologically inspected for any damage. The site's emergency action plan shall include provisions for the performance of these corrective actions.

12.2.9 100 % Fuel Rod Failure 12.2.9.1 Cause of 100 % Fuel Rod Failure This accident event postulates that all fuel rods rupture and that the appropriate quantities of fis-sion product gases and fill gas are released from the fuel rods into the canister cavity. Through all credible ACS, the CASTOR geo69 DSS maintains the SNF in an inert environment while main-

- taining the peak fuel cladding temperature below the required temperature limits, thereby providing assurance of fuel cladding integrity. There is no credible cause for 100 % fuel rod failure. This ac-cident is postulated in accordance with NUREG-2224 [4].

12.2.9.2 100 % Fuel Rod Failure Analysis According to NUREG-2224, the two separate cases accident fire conditions (HAG-fire) and acci-dent impact conditions (HAG-impact) are to be analysed. The fraction of fission gas released is 15 % for HAG-fire and 35 % for HAG-impact, which includes an extra 20 % fraction of the pellet-retained fission gases that might be released during a drop impact. The amount of fuel rod filling gas released is calculated in the appendix of Chapter 7.

Structural The pressurisation of the containment as a result of 100 % fuel rod failure is analyzed in Sec-tion 7.4. The maximum absolute pressure is *** in the canister (HAG-impact) and in the storage cask (HAG-fire). The applied pressure values in Chapter 3 bound the pressure cal-culated assuming 100% fuel rod rupture. Structural integrity of the canister and the storage cask are neither impaired for HAG-fire nor for HAG-impact.

Thermal The 100 % fuel rod failure has no effect on the thermal performance of the DSS. The maximum temperatures of the DSS components and the content are far below the maximum admissible val-ues with large safety margins (see Section 4.8).

12.2 Accidents Section 12.2, Rev. 1 Page 12.2-15

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Shielding The 100 % fuel rod failure has no effect on the shielding performance of the DSS. According to Section 5.1, the ACS dose rates without fuel damage is bounding and generates the highest dose rate at the site boundary.

Criticality The 100 % fuel rod failure has no effect on the criticality control features of the DSS.

Containment The 100 % fuel rod failure has no effect on the containment function of the DSS. Storage cask and canister containment are not impaired for HAC-fire and HAC-impact (see Chapter 3) and both con-tainment barriers remain leak-tight.

Radiation Protection Since there is no degradation in shielding or containment capabilities as discussed above, there is no effect on occupational or public exposures as a result of 100 % fuel rod failure.

12.2.9.3 100 % Fuel Rod Failure Dose Calculations The storage cask and canister containment boundaries maintain their integrity. There is no effect on the shielding effectiveness. Fuel rod failure under ACS leads to a local increase of the dose rate at the DSS surface according to Section 5.1. However, this effect vanishes with increasing dis-tance from the DSS.

12.2.9.4 100 % Fuel Rod Failure Corrective Actions As shown in the ":inalysis for 100 % fuel rod failure, neither storage cask nor canister containment boundaries are damaged. No corrective actions are required.

12.2 Accidents Section 12.2, Rev. 1 Page 12.2-16

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 List of References

[1] ANSI/ANS-57.9-1992 Design Criteria For An Independent Spent Fuel Storage Installation (Dry Type)

American National Standards Institute

• [2] ANSI N14.6 - 1993 Radioactive Materials - Special Lifting Devices for Shipping Containers Weighing 10000 Pounds (4500 kg) or More

[3] NUREG-0612, July 1980 Control of Heavy Loads at Nuclear Power Plants U.S. Nuclear Regulatory Commission, Office for Nuclear Reactor Regulation

[4] NUREG-2224, November 2020 Dry Storage and Transportation of High Burnup Spent Nuclear Fuel - Final Report U.S. Nuclear Regulatory Commission 12.2 Accidents Section 12.2, Rev. 1 Page 12.2-17

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev.0 @GNS 12.3 Appendix Prepared Reviewed 12.3 Appendix Section 12.3, Rev. 0 Page 12.3-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS With intent no items.

12.3 Appendix Section 12.3, Rev. 0 Page 12.3-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 13 Operating Controls and Limits 13.0 Overview Prepared Reviewed 13.0 Overview Section 13.0, Rev. 1 Page 13.0-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS The CASTOR geo69 DSS provides passive long-term interim dry storage of SNF inside a canister with a re-openable closure system. Canister loading under water and handling inside the reactor building is conducted via the CLU. This chapter defines the operating controls and limits (i.e. tech-nical specifications) including their supporting bases for loading of a CASTOR geo69 DSS by usage of the CLU, for storage of the DSS and for handling operations. The bases applicable to the technical specifications are provided in Appendix 13-1 in Section 13.3 in compliance with the standard format and content of NUREG-1745 [2].

List of References

[2] NUREG-1745 (June 2001)

Standard Format and Content for Technical Specifications for 10 CFR Part 72 Cask Certifi-cates of Compliance U.S. Nuclear Regulatory Commission 13.0 Overview Section 13.0, Rev. 1 Page 13.0-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 13.1 Proposed Operating Controls and Limits Prepared Reviewed 13.1 Proposed Operating Controls and Limits Section 13.1, Rev. 1 Page 13.1-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 13.1.1 Content of Operating Controls and Limits This chapter establishes the commitments regarding the CASTOR geo69 DSS and the CLU, as applicable, and their use. Other 10 CFR 72 and 10 CFR 20 [1] requirements in addition to the tech-nical specifications may apply. The licensee shall meet the conditions for a general license holder found in 10 CFR 72.212 prior to loading SNF into the CASTOR geo69 DSS. The general license conditions governed by 10 CFR 72 are not repeated in this chapter. Licensees are required to com-ply with all commitments and requirements.

13.1.2 Bases for Operating Controls and Limits The operating controls and limits are primarily established to maintain subcriticality, containment boundary and FA cladding integrity, shielding and radiation protection, heat removal capability, and structural integrity under normal, off-normal and accident conditions as well as during handling op-erations. Table 13.1-1 gives an overview of the conditions to be controlled, the required technical specifications and the chapters where the basis for a technical specification is discussed. It referees to the respective Chapters of Appendix 13-1 in Section 13.3, which provides proposals for the Ap-pendixes of the Coe regarding authorized contents in Chapter 2, technical specification in Chapter 3, design features in Chapter 4 and administrative controls in Chapter 5.

13.1.3 Training Program A training program shall be developed under the licensee's systematic approach to training to ensure the correct and safe operations of the CASTOR geo69 DSS during both, FA (un-)loading, canister dispatch and transfer into the storage cask via the CLU as well as during on-site transfer and long-term interim storage. Training modules shall include comprehensive instructions for the operation and maintenance of the DSS and CLU. A dry run training exercise on all operation procedures de-scribed in Chapter 9 like (un-)loading, closure (canister and storage cask), dewatering, drying, helium backfilling, handling, canister transshipment and (in-site) transfer of canister, transfer cask, storage cask and DSS shall be conducted by the licensee prior to the first hot use of the system to load spent fuel assemblies.

13.1 Proposed Operating Controls and Limits Section 13.1, Rev. 1 Page 13.1-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS Table 13.1-1: Conditions to be controlled in the CASTOR geo69 DSS and CLU Condition to be SAR Technical Required technical specification controlled Chapter Specification*

SNF content

• Type and condition of the FA prior to loading 1, 2 2

• Fuel channel cross section

• Fuel channel pitch Subcriticality

• Minimum neutron absorber 10 8 loading 6 4.1.1

• Maximum amount of fissile material (maximum ini-tial fuel enrichment)

• Requirements on containment components of cask Containment and canister 4.1.2 boundary and FA

• Inert atmosphere in cask and canister (quantity of 7 3.1 cladding integrity helium and residual moisture)

• Helium leak rate limits

• Requirements on shielding components of storage cask, canister and CLU Shielding and radi-

• Transfer cask water chamber filling 3.2 ation protection

• CLU sequence of operations 5, 11 4.6

• Surface contamination limits

• Minimum cooling time of FA prior to loading

• Supplemental shielding

• Requirements on heat removal components of star-age cask, canister and CLU Heat removal

• Maximum heat designed to be dissipated 4 4.5

• Storage cask array and spacing limits

• Minimum cooling time of FA prior to loading

• Requirements on structural components of storage cask, canister and CLU 4.2 Structural integrity

• Fabrication and design codes 2, 3 4.3

• Design basis earthquake, tornado, flood, etc . 3.4

• Handling and lifting devices 4.6

• ISFSI concrete pad

• Respective Chapter of the Technical Specification Basis 1014-TR-00077 Rev. 0 in Appendix 13-1 at-tached to Section 13.3 of this SAR List of References

[1] Title 10 CFR Part 20 Standards for Protection Against Radiation U.S. Nuclear Regulatory Commission 13.1 Proposed Operating Controls and Limits Section 13.1, Rev. 1 Page 13.1-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 13.2 Development of Operating Controls and Limits Prepared Reviewed 13.2 Development of Operating Controls and Limits Section 13.2, Rev. 1 Page 13.2-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 @GNS 13.2.1 Functional and Operating Limits, Monitoring Instruments, and Limiting Control Settings The controls and limits specified in this subsection apply to operating parameters and conditions that are observable, detectable, and/or measurable. The CASTOR geo69 DSS and the CLU are pas-sively cooled. During long-term interim dry storage, the DSS is equipped with a pressure switch that is connected to a pressure monitoring system (see Section 1.2). Other monitoring instruments are not required. The transfer cask is essentially passive during normal canister handling operations regarding its safety functions (structural integrity, shielding and heat removal). However, remote handling is foreseen for flooding and unflooding the cavity and the water chambers of the transfer cask and for opening and closure of the bottom lid of the transfer cask via the transfer lock.

13.2.2 Limiting Conditions for Operation Limiting conditions for operation specify the minimum capability or level of performance that is re-quired to assure that CASTOR geo69 DSS and CLU can entirely fulfil their safety functions.

13.2.2.1 Equipment The CASTOR geo69 DSS, the CLU and their respective components have been analysed for spec-ified normal, off-normal, and accident conditions of storage, including extreme environmental condi-tions as well as for handling operations. Analysis in this SAR have shown that no credible condition or event prevents the CASTOR geo69 DSS or the CLU from meeting their safety functions. As a result, there is no threat to public health and safety from any postulated condition or analysed event.

When all equipment is loaded, tested, and placed into storage in accordance with the procedures specified in this SAR, no failures of the DSS and CLU to perform their safety function are expected to occur.

13.2.2.2 Technical Conditions and Characteristics Contents shall be limited to SNF as described in Section 2.1. A change of the fuel parameters as listed in Section 2.1 requires NRC approval. The loading plan shall ensure that the maximum per-mitted total heat power is not exceeded. Therefore, the content shall comply with the design features for loading of contents as summarized in Section 1.2. The height of the FA shoes shall be adjusted to the FA length to maintain the axial FA position under all credible conditions.

The CASTOR geo69 DSS shall be stored with a minimum centre-to-centre spacing between neigh-bouring storage casks (pitch) of - to ensure sufficient heat removal.

13.2 Development of Operating Controls and Limits Section 13.2, Rev. 1 Page 13.2-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 13.2.3 Surveillance Requirements The evaluations in this SAR show that CASTOR geo69 DSS and CLU both fulfil their safety func-tions, provided that the authorized contents as well as the technical specifications and surveillance requirements for loading, unloading and storage operations proposed in the technical specifications basis (Appendix 13-1 in Section 13.3) are satisfied.

13.2.4 Design Features 13.2.4.1 Design Features Significant to Safety Design features significant to safety are those, which would have a significant effect on safety if altered or modified. These features require design, fabrication and operational controls. The follow-

- ing design features have a significant impact on subcriticality:

J

• Distance between neighbouring FA positions: ***(thickness of fuel basket sheets) 10 B loading in the criticality control material: ~

• Maximum heavy metal mass: (fully loaded DSS) with these design features ensures subcriticality for the content described in Subsection 1.2.3.

13.2.4.2 Codes and Standards Section 2.0 provides information on the governing codes and standards for SSCs of the CASTOR geo69 DSS and CLU design. It further provides a list of code alternatives applied to the DSS and CLU design. A detailed listing of component items, including the material specification, is given in Section 1.2 and the appendixes in Section 1.5. The applicable codes and standards for materials are given in Section 8.1. Alternative materials are specified in the respective parts lists. No other materials are permitted for the SSCs of the CASTOR geo69 DSS and CLU.

13.2 Development of Operating Controls and Limits Section 13.2, Rev. 1 Page 13.2-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 13.2.4.3 Structural Performance The CASTOR geo69 DSS is designed to withstand the effects of an earthquake at the storage site without tipping. The design earthquake ground motion on the top surface of the storage facility pad shall not exceed the following combination of horizontal peak acceleration (in each of the two or-thogonal acceleration directions) and vertical peak acceleration:

Horizontal acceleration: -

Vertical acceleration: -

The use of the CASTOR DSS is limited to sites that are bounded by these peak acceleration values.

Site-specific design basis earthquakes exceeding the listed permissible peak acceleration values require an additional evaluation of the DSS design with regard to the structural consequences of such an earthquake. Additional safety measures (e.g. anchoring of the DSS on the storage pad) shall be implemented to ensure that no tip-over will result from the design basis earthquake.

A maximum permitted lifting height (above ground) is not established for the CASTOR geo69 DSS and CLU. Storage cask, transfer cask and canister shall be handled via single-failure proof handling devices, which satisfy the enhanced safety criteria of NUREG-0612 [1] and are designed in accord-ance with ANSI N14.6 [2].

List of References

[1] NUREG-0612, July 1980 Control of Heavy Loads at Nuclear Power Plants U.S. Nuclear Regulatory Commission, Office for Nuclear Reactor Regulation

[2] ANSI N14.6 - 1993 Radioactive Materials - Special Lifting Devices for Shipping Containers Weighing 10 000 Pounds (4500 kg) or More

• 13.2 Development of Operating Controls and Limits Section 13.2, Rev. 1 Page 13.2-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 13.3 Appendix Prepared Reviewed 13.3 Appendix Section 13.3, Rev. 1 Page 13.3-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 1 Appendix 13-1: 1014-TR-00077 Rev. 0 Technical Specification Basis Dry Storage System CASTOR geo69 DSS 13.3 Appendix Section 13.3, Rev. 1 Page 13.3-2

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.39 Technical Specification Basis Dry Storage System CASTOR geo69 Document Type Technical Report Document No. 1014-TR-00077 Revision 0 Name (Function) Date Signature Prepared Reviewed Approved a;

0:::

N 0

0 u.

This document may not be cited. reproduced in whole or in part, or made available to third parties without the prior written consent of GNS Gesellschaft fur Nuklear-Service mbH, Essen.

All rights reserved by GNS.

This document contains business and trade secrets of GNS.

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0 Status of Revision Revision Date Author Reason for change 0 13.12.2022

- First issue CASTOR is a registered trade mark.

Status of Revision Page 2 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0

@GNS Table of Contents Status of Revision 2 Table of Contents 3 1 Use and Application 5 1.1 Definitions 5 1.2 Logical Connectors 7 1.3 Completion Time 9 1.4 Frequency 12 2 Approved Contents 14

- 3 3.1 3.1.1 Limiting Condition for Operation (LCO) Applicability & Surveillance Requirement (SR) Applicability DSS and Fuel Integrity Water Temperature 15 17 17 3.1.2 Canister Cavity Vacuum Drying 18 3.1.3 Canister Helium Backfill Pressure 19 3.1.4 Canister standard helium leakage rate 20 3.1.5 Storage Cask Cavity Vacuum Drying 21 3.1.6 Storage Cask Helium Backfill Pressure 22 3.1.7 Storage Cask Cavity Pressure Monitoring 23 3.1.8 Cask standard helium leakage*rate 25 3.1.9 Storage Cask Minimum Handling Temperature 26 3.2 DSS Radiation Protection 27 3.2.1 Contamination control 27 3.2.2 External dose rate control 28 4 Design Features 30 4.1 Design Features Significant to Safety 30 4.1.1 Criticality Control 30 4.1.2 Materials 30 4.2 Codes and Standards 30 4.2.1 Alternatives to Codes, Standards, and Criteria 30 4.2.2 Construction/Fabrication Alternatives to Codes, Standards, and Criteria 31 4.3 Structural Performance 31 4.3.1 Earthquake Loads 31 4.3.2 Design g-Loads 31 Table of Contents Page 3 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev.a 4.4 Storage Cask Handling and On-Site Transfer 32 4.5 Storage Pad 32 4.6 ISFSI Specific Parameters and Analyses 32 5 Administrative Controls 33 5.1 Administrative Programs 33 5.1.1 Radioactive Effluent Control Program 33 5.1.2 Cask Loading, Unloading, and Preparation Program 33 5.1.3 ISFSI Operations Program 33 List of References 34 Table of Contents Page 4 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0 1 Use and Application This report defines the Technical Specification Bases for the CASTOR geo69 Dry Storage Sys-tem for spent nuclear fuel in accordance with the requirements of NU REG 1745 [1 ].

1.1 Definitions The defined terms of this section appear in capitalized type and are applicable throughout these technical specifications and bases.

Term Definition ACTIONS ACTIONS shall be that part of a specification that prescribes re-quired actions to be taken under designated conditions within spec-ified completion times.

CANISTER The CANISTER is the inner containment barrier of the DRY STORAGE SYSTEM. It consists of a welded shell and bottom as well as a bolted lid which is sealed by a metallic gasket. Inside of the CANISTER the basket is assembled to receive the spent nu-clear fuel assemblies.

DRY STORAGE SYSTM DRY STORAGE SYSTEMs (DSS) are approved for the storage of spent nuclear fuel assemblies at the ISFSI. The CASTOR geo69 DRY STORAGE SYSTEM consists of the STORAGE CASK, its in-tegral CANISTER and a protection cover laid on the lid-end of the STORAGE CASK.

INTACT FUEL INTACT FUEL ASSEMBLY is a fuel assembly without known or ASSEMBLY suspected cladding defects greater than pinhole leak or a hairline crack and which can be handled by normal means. Partial fuel as-semblies, that is fuel assemblies from which fuel rods are missing, shall not be classified as intact fuel assemblies unless dummy fuel rods are used to displace an amount of water greater than or equal to that displaced by the original fuel rod(s).

LOADING LOADING OPERATIONS include all licensed activities on a OPERATIONS STORAGE CASK, TRANSFER CASK or CANISTER while it is be-ing loaded with fuel assemblies. LOADING OPERATIONS begin when the first fuel assembly is placed in the CANISTER and end when the STORAGE CASK is suspended from or secured on the transporter/transport vehicle.

STORAGE CASK The STORAGE CASK is the outer containment barrier of the DRY STORAGE SYSTEM which consists of a monolithic cask body made of ductile cask iron and a bolted cask lid which is sealed by a metallic gasket.

1 Use and Application Page 5 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0 STORAGE STORAGE OPERATIONS include all licensed activities that are OPERATIONS performed at the Independent Spent Fuel Storage Installation (ISFSI) while a DSS containing spent fuel is sitting on a storage pad within the ISFSI.

TRANSFER CASK The TRANSFER CASK is designed to contain the CANISTER dur-ing and after loading of spent nuclear fuel assemblies and to trans-fer the CANISTER to the STORAGE CASK.

TRANSFER TRANSFER OPERATIONS include all licensed activities per-OPERATIONS formed on the TRANSFER CASK loaded with the CANISTER con-taining one or more fuel assemblies. TRANSFER OPERATIONS begin after the CANISTER is loaded into the TRANSFER CASK and end after the CANISTER has been removed from the TRANSFER CASK.

ON-SITE TRANSFER ON-SITE TRANSFER OPERATIONS include all licensed activities OPERATIONS performed on the STORAGE CASK loaded with one or more fuel assemblies when it is being moved to and from the ISFSI. ON-SITE TRANSFER OPERATIONS begin when the STORAGE CASK is first suspended from or secured on the transfer vehicle and end when the STORAGE CASK is at its destination and no longer se-cured or suspended from the transfer vehicle.

UNLOADING UNLOADING OPERATIONS include all licensed activities on a OPERATIONS STORAGE CASK to be unloaded of the contained fuel assemblies.

UNLOADING OPERATIONS begin when the STORAGE CASK is no longer suspended from or sect.ired on the transporter or transfer vehicle and end when the last fuel assembly is removed from the STORAGE CASK.

1 Use and Application Page 6 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0 1.2 Logical Connectors PURPOSE The purpose of this section is to explain the meaning of logical con-nectors.

Logical connectors are used in Technical Specifications (TS) to dis-criminate between, and yet connect, discrete conditions, required ac-tions, completion times, surveillances, and frequencies. The only log-ical connectors that appear in TS are AND and OR. The physical ar-rangement of these connectors constitutes logical conventions with specific meanings.

BACKGROUND Several levels of logic may be used to state required actions. These levels are identified by the placement (or nesting) of the logical con-nectors and by the number assigned to each required action. The first level of logic is identified by the first digit of the number assigned to a required action and the placement of the logical connector in the first level of nesting (i.e. left justified with the number of the required ac-tion). The successive levels of logic are identified by additional digits of the required action number and by successive indentions of the log-ical connectors.

When logical connectors are used to state a condition, completion time, surveillance, or frequency, only the first level of logic is used, and the logical connector is left justified with the statement of the con-dition, completion time, surveillance, or frequency.

1 Use and Application Page 7 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0

@GNS EXAMPLES The following examples illustrate the use of logical connectors.

EXAMPLE 1.2-1 ACTIONS CONDITION REQUIRED ACT/Of:' COMPLETION TIME A. LCO not met. A.1 Verify ...

AND A.2 Restore ...

In this example, the logical connector AND is used to indicate that when in condition A, both required actions A.1 and A.2 must be com-pleted.

EXAMPLE 1.2-2 ACTIONS

'COl!IDIT/ON REQUIRED ACTION COMPLETION

.. TIME A. LCO not met. A.1 Stop ...

OR A.2.1 Verify ...

AND A.2.2.1 Reduce ...

OR A.2.2.2 Perform ...

OR A.3 Remove ...

This example represents a more complicated use of logical connect-ors. Required Actions A.1, A.2 and A.3 are alternative choices, only one of which must be performed as indicated by the use of the logical connector OR and the left justified placement. Any one of these three ACTIONS may by chosen. If A.2 is chosen, then both A.2.1 and A.2.2 must be performed as indicated by the logical connector AND. Re-quired action A.2.2 is met by performing A.2.2.1 or A.2.2.2. The in-tended position of the logical connector OR indicates that A.2.2.1 and A.2.2.2 are alternative choices, only one of which must be performed.

1 Use and Application Page 8 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS 1.3 Completion Time PURPOSE The purpose of this section is to establish the completion time conven-tion and to provide guidance for its use.

BACKGROUND Limiting Conditions for Operation (LCOs) specify the lowest functional capability or performance levels of equipment required for safe opera-tion of the facility. The ACTIONS associated with an LCO state condi-tions that typically describe the ways in which the requirements of the LCO can fail to be met. Specified with each stated condition are re-quired actions(s) and completion times.

DESCRIPTION The completion time is the amount of time allowed for completing a re-quired action. It is referenced to the time of discovery of a situation (e.g.

equipment or variable not within limits) that requires entering an ACTIONS condition unless otherwise specified, providing the cask sys-tem is in a specified condition state in the applicability of the LCO. Re-quired Actions must be completed prior to the expiration of the specified completion time. An ACTIONS condition remains in effect and the re-quired actions apply until the condition no longer exists or the cask sys-tem is not within the LCO applicability.

Once a condition has been entered, subsequent subsystems, compo-nents, or variables expressed in the condition, discovered to be not within limits, will not result in separate entry into the condition unless otherwise stated. The required actions of the condition continue to apply to each additional failure, with completion times based on initial entry into the condition.

EXAMPLES The following examples illustrate the use of Completion Times with different types of Conditions and changing Conditions.

EXAMPLE 1.3-1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. Required Action B.1 Perform 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and associated Action B.1.

completion time AND not met.

B.2 Perform 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Action 8.2.

Condition B has two required actions. Each required action has its own separate completion time. Each completion time is referenced to the time that condition B is entered.

1 Use and Application Page 9 of 34

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-TR-00077 Rev. 0 s

The required actions of condition 8 are to complete action 8.1 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> AND complete action 8.2 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. A total of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> is allowed for completing action 8.1 and a total of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> (not 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />) is allowed for completing action 8.2 from the time that condi-tion 8 was entered. If action 8.1 is completed within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, the time allowed for completing action 8.2 is the next 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> because the total time allowed for completing action 8.2 is 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

EXAMPLE 1.3-2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One system not A.1 Restore system to 7 days within limit within limit.

B. Required Action B.1 Complete Action 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and associated 8.1.

completion time AND not met.

8.2 Complete Action 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> B.2.

When a system is determined not to meet the LCO, Condition A is entered. If the system is not restored within 7 days, Condition 8 is also entered and the Completion Time clocks for Required Actions 8.1 and 8.2 start. If the system is restored after Condition 8 is entered, Conditions A and 8 are exited, and therefore, the Required Actions of Condition 8 may be terminated EXAMPLE 1.3-3 ACTIONS

---

Note------------------------------------

Separate Condition entry is allowed for each component.

CONDITION REQUIRED ACTION COMPLETION TIME A. One system not A.1 Restore compli- 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> within limit ance with LCO B. Required 8.1 Complete 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Action and Action 8.1.

associated com- AND pletion time not met. 8.2 Complete 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> Action 8.2.

1 Use and Application Page 10 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0 G S The Note above the ACTIONS table is a method of modifying how the Completion Time is tracked. If this method of modifying how the Completion Time is tracked was applicable only to a specific Condition, the Note would appear in that Condition rather than at the top of the ACTIONS Table.

The Note allows Condition A to be entered separately for each component and Completion Times tracked on a per component basis.

When a component is determined to not meet the LCO, Condition A is entered and its Completion Time starts. If subsequent components are determined to not meet the LCO, Condition A is entered for each component and separate Completion Times start and are tracked for each component.

IMMEDIATE When "Immediately" is used as a Completion Time, the Required COMPLETION TIME Action should be pursued without delay and in a controlled manner.

1 Use and Application Page 11 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0 1.4 Frequency PURPOSE The purpose of this section is to define the proper use and application of Frequency requirements.

DESCRIPTION Each Surveillance Requirement (SR) has a specified Frequency in which the Surveillance must be met in order to meet the associated Limiting Condition for Operation (LCO). An understanding of the correct application of the specified Frequency is necessary for compliance with the SR.

The "specified Frequency" is referred to throughout this section and each of the Specifications of Section 3.0, Surveillance Requirement (SR) Applicability. The "specified Frequency" consists of the require-ments of the Frequency column of each SR.

Situations where a Surveillance could be required (i.e., its Frequency could expire), but where it is not possible or not desired that it be per-formed until sometime after the associated LCO is within its Applicabil-ity, represent potential SR 3.0.4 conflicts. To avoid these conflicts, the SR (i.e., the Surveillance or the Frequency) is stated such that it is only "required" when it can be and should be performed. With an SR satis-fied, SR 3.0.4 imposes no restriction.

EXAMPLES The following examples illustrate the various ways that Frequencies are specified.

EXAMPLE 1.4-1 SURVEILLANCE REQUIREMENTS SURVEILLANCE r FREQUENCY Verify pressure within limit 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> Example 1.4-1 contains the type of SR most often encountered in the Technical Specifications (TS). The Frequency specifies an interval (12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />) during which the associated Surveillance must be performed at least one time. Performance of the Surveillance initiates the subse-quent interval. Although the Frequency is stated as 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, an exten-sion of the time interval to 1.25 times the interval specified in the Fre-quency is allowed by SR 3.0.2 for operational flexibility. The measure-ment of this interval continues at all times, even when the SR is not required to be met per SR 3.0.1 (such as when the equipment or varia-bles are outside specified limits, or the facility is outside the Applicability of the LCO). If the interval specified by SR 3.0.2 is exceeded while the facility is in a condition specified in the Applicability of the LCO, the LCO is not met in accordance with SR 3.0.1. If the interval as specified by SR 3.0.2 is exceeded while the facility is not in a condition specified in 1 Use and Application Page 12 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS the Applicability of the LCO for which performance of the SR is required, the Surveillance must be performed within the Frequency requirements of SR 3.0.2 prior to entry into the specified condition. Failure to do so would result in a violation of SR 3.0.4 EXAMPLE 1.4-2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY Verify flow is within limits Once within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior to start-ing activity AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter Example 1.4-2 has two Frequencies. The first is a one time perfor-mance Frequency, and the second is of the type shown in Example 1.4-1. The logical connector "AND" indicates that both Frequency re-quirements must be met. Each time the example activity is to be per-formed, the Surveillance must be performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> prior to starting the activity.

The use of "once" indicates a single performance will satisfy the speci-fied Frequency (assuming no other Frequencies are connected by "AND"). This type of Frequency does not qualify for the 25% extension allowed by SR 3.0.2.

"Thereafter" indicates future performances must be established per SR 3.0.2, but only after a specified condition is first met (i.e., the "once" performance in this example). If the specified activity is cancelled or not performed, the measurement of both intervals stops. New intervals start upon preparing to restart the specified activity.

1 Use and Application Page 13 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0 2 Approved Contents The spent nuclear fuel to be stored in the CASTOR geo69 cask shall meet the following require-ments:

A. Fuel shall be unconsolidated INTACT FUEL ASSEMBLIES.

B. Fuel shall be limited to fuel with Zircaloy cladding.

C. Fuel shall be limited to the following fuel types with the following specifications:

D. The maximum decay heat per DSS shall be limited to 18.5 kW while the maximum decay heat per FA position can differ between - - - - - - * [see Chapter 4.1.2 of the SAR].

E. Maximum heavy metal mass:

2 Approved Contents Page 14 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0 2.1 Limiting Condition for Operation (LCO) Applicability & Surveillance Re-quirement (SR) Applicability Limiting Conditions for Operation (LCO) Applicability LCO 3.0.1 LCOs shall be met during specified conditions in the applicability, except as provided in LCO 3.0.2.

LCO 3.0.2 Upon discover of a failure to meet an LCO, the required actions of the asso-ciated conditions shall be met.

If the LCO is met or is no longer applicable prior to expiration of the specified completion time(s), completion of the required actions(s) is not required, unless otherwise stated.

f-------+-----

LCO 3.0.3 Not applicable.

LCO 3.0.4 When an LCO is not met, entry into a specified condition in the Applicability shall not be made except when the associated ACTIONS to be entered permit continued operation in the specified condition in the Applicability for an unlim-ited period of time. This Specification shall not prevent changes in specified conditions in the Applicability that are required to comply with ACTIONS or that are related to the unloading of an DSS.

LCO 3.0.5 Not applicable.

2 Approved Contents Page 15 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0 Surveillance Requirements (SR) Applicability SR 3.0.1 SRs shall be met during the specified conditions in the Applicability for indi-vidual LCOs, unless otherwise stated in the SR. Failure to meet a Surveil-lance, whether such failure is experienced during the performance of the Sur-veillance or between performances of the Surveillance, shall be failure to meet the LCO. Failure to perform a Surveillance within the specified Fre-quency shall be failure to meet the LCO except as provided in SR 3.0.3. Sur-veillances do not have to be performed on equipment or variables outside specified limits.

SR 3.0.2 The specified Frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met.

For Frequencies specified as "once," the above interval extension does not apply. If a Completion Time requires periodic performance on a "once per... "

basis, the above Frequency extension applies to each performance after the initial performance.

Exceptions to this Specification are stated in the individual Specifications.

SR 3.0.3 If it is discovered that a Surveillance was not performed within its specified Frequency, then compliance with the requirement to declare the LCO not met may be delayed, from the time of discovery, up to 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or up to the limit of the specified Frequency, whichever is less. This delay period is permitted to allow performance of the Surveillance.

If the Surveillance is not performed within the delay period, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered. When the Surveillance is performed within the delay period and the Surveillance is not met, the LCO must immediately be declared not met, and the applicable Condition(s) must be entered.

SR 3.0.4 Entry into a specified condition in the Applicability of an LCO shall not be made unless the LCO's Surveillances have been met within their specified Frequency. This provision shall not prevent entry into specified conditions in the Applicability that are required to comply with Actions or that are related to the unloading of an DSS.

2 Approved Contents Page 16 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS 2.2 D55 and Fuel Integrity 2.2.1 Water Temperature LCO 3.1.1 The time after the canister lid is mounted on the loaded CANISTER and the dewatering process is started shall not exceed

• hours.

APPLICABILITY: During LOADING OPERATIONS ACTIONS:

CONDITION REQUIRED ACTION COMPLETION TIME A. Loaded CANISTER in A.1 Lift up the transfer cask until the Immediately SNF pool; CANISTER CANISTER lid is accessible. De-lid is mounted. contaminate the canister lid.

AND A.2 Install the multi-equipment (incl. *hours drainage lance) on the CANISTER lid and start dewatering.

B. Dewatering of the B.1 Lower the CAN !STER back into Immediately CANISTER cannot be the FA storage pool and subse-initiated within the asso- quently remove the CANISTER lid ciated completion time. to cool the loaded FA.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.1 Verify the time after mounting the canister lid and before Once, prior to dewatering process is started is :,;

• hours. SR 3.0.4 does TRANSFER opera-not apply. tions 2 Approved Contents Page 17 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0 2.2.2 Canister Cavity Vacuum Drying LCO 3.1.2 The CANISTER cavity vacuum drying pressure rise shall be sustained after isolation from the vac-uum drying system.

APPLICABILITY: During LOADING OPERATIONS ACTIONS:

CONDITION REQUIRED ACTION COMPLETION TIME.

A. Loaded CANISTER is A.1 Establish CANISTER cavity drying dewatered, CANISTER criteria within limits.

lid is mounted.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.2 Verify the CANISTER cavity vacuum drying criteria has Once, prior to been fulfilled TRANSFER OPERATIONS 2 Approved Contents Page 18 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev.a @GNS 2.2.3 Canister Helium Backfill Pressure LCO 3.1.3 The CANISTER cavity shall be backfilled with helium to a pressure of APPLICABILITY: During LOADING OPERATIONS ACTIONS:

CONDITION . , ~EQU/REDACT/0,N COMPLETION TIME A. Loaded CANISTER is A.1 Achieve or maintain a nominal he- Immediately dried, CANISTER lid is lium environment in the mounted CANISTER AND A.2 Establish CANISTER cavity he-lium backfill pressure within limits.

SURVEILLANCE REQUIREMENTS SURVEILLANCE* .. FREQUENCY SR 3.1.3 Verify that the CANISTER cavity pressure is Once, prior to TRANSFER OPERATIONS 2 Approved Contents Page 19 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0

@GNS 2.2.4 Canister standard helium leakage rate LCO 3.1.4 The standard helium leakage rate for the CANISTER closure system shall not exceed 1

• 1o-oa Pa m3/s.

APPLICABILITY: During LOADING OPERATIONS ACTIONS:

CONDITION REQUIRED ACTION COMPLETION TIME A. Loaded CANISTER is A.1 Establish CANISTER helium leak dried, CANISTER lid is rate s 1

• 1o-oa Pa m3/s mounted and sealed B. CANSITER helium leak B.1 Visual inspection of sealing sur-rate cannot be achieved faces and change of metal gasket AND B.2 Establish CANISTER helium leak rate within limit.

C. Required Action B.2 not C.1 Lower the CANISTER back into met. the FA storage pool and subse-quently remove the CANISTER lid to cool the loaded FA.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4 Verify CANISTER helium leak rate is within limit. Once, prior to TRANSFER OPERATIONS 2 Approved Contents Page 20 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0

@GNS 2.2.5 Storage Cask Cavity Vacuum Drying LCO 3.1.5 The STORAGE CASK cavity vacuum drying pressure rise shall be sus-tained after isolation from the vacuum drying system.

APPLICABILITY: During LOADING OPERATIONS ACTIONS:

CONDITION REQUIRED ACTION COMPLETION TIME A. Loaded CANISTER is A.1 Establish STORAGE CASK cavity loaded into the drying criteria within limits.

STORAGE CASK, STORAGE CASK lid is mounted SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY f-------------

SR 3.1.5 Verify the STORAGE CASK cavity vacuum drying criteria Once, prior to ON-has been fulfilled SITE TRANSFER OPERATIONS 2 Approved Contents Page 21 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0 2.2.6 Storage Cask Helium Backfill Pressure LCO 3.1.6 The STORAGE CASK cavity shall be backfilled with helium to a maximum pressure APPLICABILITY: During LOADING and STORAGE Operations ACTIONS:

• . CONDITION: *REQUIRED ACT/ON COMPLETION TIME A. Loaded CANISTER is A.1 Achieve or maintain a nominal he-loaded into the lium environment in the STORAGE CASK, STORAGE CASK STORAGE CASK lid is AND mounted A.2 Establish STORAGE CASK cavity helium backfill pressure within lim-its.

SURVEILLANCE REQUIREMENTS

. . .. SURVEILLANCE .. FREQUENCY SR 3.1.6 Verify STORAGE CASK pressure is within limit. Once, prior to ON-SITE TRANSFER OPERATIONS and Constantly, via pres-sure monitoring sys-tern during STORAGE OPERATION 2 Approved Contents Page 22 of 34

Non-Proprietary Version 1014-TR-00077 Rev. 0 Proprietary Information withheld per 10CFR 2.390 s

2.2.7 Storage Cask Cavity Pressure Monitoring LCO 3.1.7 Pressures in STORAGE CASK and CANISTER shall both be maintained at the specified filling pressures. Thus, the cask lid is equipped with a pressure switch (see Section 1.2 of the SAR) that detects a pressure drop in the storage cask. It is connected to the pressure monitoring system of the stor-age facility. An alert if either a defect of the pressure switch or a leakage of a containment boundary of the CANISTER or the STORAGE CASK occurs.

APPLICABILITY: During STORAGE OPERATIONS ACTIONS:

Condition . Required Action Completion Time A Pressure switch signals A.1 Inform the supervisory authority to Immediately a pressure drop. coordinate the following procedure.

AND A.2 Remove the protection cover from The time schedule for the STORAGE CASK. the procedures follow-ing the indicated pres-AND sure drop in the star-A.3 Check leak tightness of the age cask shall be co-STORAGE CASK lid system. ordinated with the su-pervisory authority.

B Leakage in the B.1 Restore leak-tightness of As soon as possible STORAGE CASK lid STORAGE CASK lid system. Re-system. store pressure and helium atmos-phere in the STORAGE CASK.

C STORAGE CASK lid C.1 Check pressure in the reference As soon as possible system is leak-tight in chamber of the pressure switch to accordance with ANSI determine whether there is a leak N14.5. in the pressure switch membrane or in the containment boundary of the CANISTER.

AND C.2.1 Replace the defect pressure As soon as possible switch with a new exemplar.

OR C.2.2 Coordinate further procedures with the supervisory authority.

As soon as possible The storage cask internal pres-sure may be determined as a decision support.

2 Approved Contents Page 23 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0

@GNS SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY f------------

SR 3.1.7 Verify the successful commissioning of the new pressure Prior to continuous switch. storage 2 Approved Contents Page 24 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS 2.2.8 Cask standard helium leakage rate LCO 3.1.8 The combined helium leakage rate for the STORAGE CASK closure sys-tem shall not exceed 1

• 1o-s Pa m3/s.

• APPLICABILITY: During LOADING OPERATIONS ACTIONS:

CONDITION REQUIRED ACTION COMPLETION TIME A. Loaded and sealed A.1 Establish STORAGE CASK CANISTER is loaded standard helium leakage rate :;:;

into the STORAG 1

• 1o-spa m3/s CASK, STORAGE CASK lid is mounted and sealed B. STORAGE CASK B.1 Visual inspection of sealing sur-standard helium leak- faces and change of metal gasket age rate cannot be AND achieved within associ-ated completion time B.2 Establish STORAGE CASK he-lium leakage rate within limit.

C. Required Action B.2 not C.1 Unload the CANISTER from the met. STORAGE CASK.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.8 Verify STORAGE CASK helium leakage rate is within limit. Once, prior to ON-SITE TRANSFER 2 Approved Contents Page 25 of 34

Non-Proprietary Version 1014-TR-00077 Rev. 0 Proprietary Information withheld per 10CFR 2.390 s

2.2.9 Storage Cask Minimum Handling Temperature LCO 3.1.9 The loaded STORAGE CASK shall not be handled if the outer surface of the cask is below -29 °C.

APPLICABILITY: During ONSITE TRANSFER and STORAGE OPERATIONS ACTIONS:

CONDITION REQUIRED ACTION . COMPLETION TIME A. STORAGE Cask sur- A.1 Stop handling operations and Immediately face temperature be- place STORAGE CASK in safe low limit. position SURVEILLANCE REQUIREMENTS

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Note------------------------------------

This surveillance does not need to be performed if ambient temperature is known to be above limit SURVEILLANCE FREQUENCY SR 3.1.9 Verify STORAGE CASK surface temperature is above Once, immediately limit. before handling aper-ations.

2 Approved Contents Page 26 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0 2.3 DSS Radiation Protection 2.3.1 Contamination control LCO 3.2.1 The non-fixed contamination level on the STORAGE CASK surface shall be in accordance with the requirements of 49 CFR 173.443 APPLICABILITY: During LOADING OPERATIONS ACTIONS:

CONDITION REQUIRED ACTION COMPLETION TIME.

A. STORAGE CASK is A.1 Perform measurement of contam-loaded and sealed ination on STORAGE CASK sur-face B. STORAGE CASK sur- 8.1 Decontaminate STORAGE CASK face contamination is surface not within limits AND 8.2 Perform measurement of contam-ination on STORAGE CASK sur-face SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.1 Verify STORAGE CASK surface contamination is within Once, prior to ON-limit. SITE TRANSFER 2 Approved Contents Page 27 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0 2.3.2 External dose rate control LCO 3.2.2 The radiation level outside of the STORAGE CASK shall be in accordance with the requirements of 10 CFR 71.43.

APPLICABILITY: During LOADING OPERATIONS ACTIONS:

CONDITION REQUIRED ACTION COMPLETION TIME A. STORAGE CASK is A.1 Perform measurement of external loaded and sealed radiation levels B. STORAGE CASK exter- 8.1 Install temporary additional shield- Immediately nal radiation levels are ing not within limits AND B.2 Establish whether a loaded FA As soon as possible deviates from the loading plan by review of records, etc.

C. 8.2 results in misloaded C.1 Analysis: Determination by calcu- As soon as possible FA lation of the ISFSI offsite or occu-pational dose rates and compare with regulatory limits or the re-quirements of the COC or SAR, respectively.

D. Calculated ISFSI offsite D.1 Storage operation may continue or occupational dose OR rates are within regula-tory limits or the require- E.1 ments of the COC or SAR E. Calculated ISFSI offsite E.1 Unloading of the CANISTER from As soon as possible or occupational dose the STORAGE CASK via rates exceed regulatory TRANSFER CASK and transfer to limits or the require- the SNF pool, recooling of the FA ments of the COC or and unloading SAR F. B.2 results in correct F.1=C.1 loading 2 Approved Contents Page 28 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.2.2 Verify external DSS radiation levels are within limit. Once, prior to ON-SITE TRANSFER and before each handling during STORAGE OPERATIONS 2 Approved Contents Page 29 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0 3 Design Features 3.1 Design Features Significant to Safety 3.1.1 Criticality Control

1. Fuel cell spacing:

2. 10 8 loading in the Baral neutron absorbers:

3. Maximum heavy metal mass:

3.1.2 Materials

1. Protective Coatings used inside and outside the storage and transfer cask (Subsec-tion 8.2.6)

2. Neutron Absorbers and Shields (Subsection 8.2.2.5, Subsection 10.1.6.2)

3. Containment Boundary Seals (Subsection 8.2.5) 3.2 Codes and Standards The 2017 edition of the ASME Boiler & Pressure Vessel Code is the governing code for the struc-tural design of the CASTOR geo69 DSS and CLU to the most practical extend. The components important to safety of the DSS are designed, fabricated and inspected in accordance with, Sec-tion Ill, Division 3, [2] whereas the components important to safety of the CLU are designed, fab-ricated and inspected according to Section Ill, Division 1, Subsection NF [3] as class 1 support.

Exceptions are the load attachment points (trunnions, etc.) which are designed according to ANSI N14.6 [4] and NUREG-0612 [5]. Table 2.0-3 and Table 2.0-4 of the SAR lists details regarding

- the applicable governing codes for material procurement, design, fabrication and examination of the DSS and the CLU, respectively.

3.2.1 Alternatives to Codes, Standards, and Criteria Table 2.0-5 of the SAR lists alternative codes and exemptions to the ASME code for the DSS design.

3 Design Features Page 30 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0

@)GNS 3.2.2 Construction/Fabrication Alternatives to Codes, Standards, and Criteria Proposed alternatives to ASME Code including alternatives allowed by Table 2.0-5 of the SAR may be used when authorized by the Director of the Office of Nuclear Material Safety and Safe-guards or designee. The request for such alternatives should demonstrate that:

1. The proposed alternatives would provide an acceptable level of quality and safety, or

2. Compliance with the specified requirements of the ASME Code would result in hardship or unusual difficulty without a compensating increase of the level of quality and safety.

Request for alternatives shall be submitted in accordance with 10 CFR 72.4.

3.3 Structural Performance 3.3.1 Earthquake Loads The DSS earthquake load on the top surface of the ISFSI storage pad shall not exceed:

Horizontal peak acceleraticin"in each of .Corresponding vertical peak acce!eratiori .*

the two orthogonaldirections 3.3.2 Design g-Loads Since all handling operations with the storage cask or CLU require single failure proof crane in-stallations (ANSI N14.6 and NUREG-0612) and all load attachment points (trunnions, tilting suds) are respectively designed, cask drops are not credible events. Furthermore it is demonstrated that the DSS will not tip over caused by a seismic event, flood or tornado. However, the analyses of storage side and bottom end storage cask drops are evaluated bounding a non-mechanistic tipover analysis acc. to NU REG 2215, Subsection 4.5.3.3.1. The fission product barrier design g-loads due to the postulated storage cask drops shall not exceed:

I

• Barrier End Drop* Side Drop*

DSS

• bounding a non-mechanistic tipover 3 Design Features Page 31 of 34

Non-Proprietary Version 1014-TR-00077 Rev. 0 Proprietary Information withheld per 10CFR 2.390 s

3.4 Storage Cask Handling and On-Site Transfer Storage cask, transfer cask and canister shall be handled exclusively via single-failure proof han-dling devices, which satisfy the enhanced safety criteria of NUREG-0612 and are designed in accordance with ANSI N14.6.

For on-site transfer of the DSS between the reactor facility and the storage pad (operation outside a Part 50 facility) the following parameters need verification by the DSS user:

1. Maximum drop height of the storage cask from the transfer vehicle onto the ground during on-site transfer,

2. Maximum on-site transfer velocity is walking speed. No unpredictable accelerations and decelerations due to well-known road condition and exclusion of other traffic participants.

3.5 Storage Pad Each DSS shall be spaced - apart, center to center, from another. This minimum spacing will ensure the proper dissipation of radiant heat energy from an array of casks as assumed in the CASTOR geo69 SAR.

The condition of the storage pad shall ensure a friction coefficient of - in relation to the DSS contact area.

3.6 ISFSI Specific Parameters and Analyses ISFSI specific parameters and analyses that shall need verification by the DSS user are, as a minimum, as follows:

1. Tornado maximum wind speed: 103 m/h (equal to 230 mph)

2. Flood levels up to full submersion of DSS) and drag forces up to

3. Average daily ambient temperatures: ~ -29 °C minimum, s 38 °C maximum

4. The potential of fires and explosions shall be addressed, based on site specific consider-ations. Fires and explosions shall be bounded by the DSS design bases parameters of 200 litres of fuel (in the tank of the transport vehicle).

5. Supplemental Shielding: In cases where engineered features (i.e. berms, shield walls) are used to ensure that the requirements of 10 CFR 72.104(a) are met, such features are to be considered Important to Safety and must be evaluated to determine the applicable Quality Assurance Category.

3 Design Features Page 32 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS 4 Administrative Controls 4.1 Administrative Programs The following programs shall be established, implemented, and maintained.

4.1.1 Radioactive Effluent Control Program

1. Leak testing in accordance with ANSI N 14.5 to ensure a standard helium leakage rate s 1* 1o-s Pam 3/s for all containment boundaries of the DSS (LCO 3.2.3 and LCO 3.2.6).

2. Surface contamination measurement according to 49 CFR 173.443 prior to removal of the storage cask from the Part 50 structure. (LCO 3.2.7) 4.1.2 Cask Loading, Unloading, and Preparation Program To ensure the requirements of the CASTOR geo69 DSS operation and maintenance instruction are fulfilled the following criteria shall be implemented in a program:

1. Vacuum drying 2: lnertization pressure of HE-atmosphere

3. Leak testing in accordance with ANSI N 14.5

4. Surface dose rate measurements at defined positions specified by the license holder

5. Ambient respectively DSS surface temperature measurement to ensure DSS surface tem-perature is above -40 °C.

6. Verification of SNF pool water characteristics:

4.1.3 ISFSI Operations Program A program shall be implemented to verify the following criteria which are used in the CASTOR geo69 DSS safety evaluation:

1. Center to center spacing between each DSS ~ -

2. Minimum friction coefficient between DSS and ISFSI pad -

3. Operation of the DSS pressure monitoring system 4 Administrative Controls Page 33 of 34

Non-Proprietary Version 1014-TR-00077 Proprietary Information withheld per 10CFR 2.390 Rev. 0

@GNS List of References

[1] NUREG 1745 Standard Format and Content for Technical Specifications for 10 CFR Part 72 Cask Certif-icates of Compliance

[2] ASME Boiler and Pressure Vessel Code Section Ill - Rules for Construction of Nuclear Facility Components Division 3 - Containment Systems for Transportation and Storage of Spent Nuclear Fuel and High-Level Radioactive Material 2017 Edition

[3] ASME Boiler and Pressure Vessel Code Section Ill - Rules for Construction of Nuclear Facility Components Division 1, Subsection NF - Supports 2017 Edition

[4] ANSI N14.6 -1993 Radioactive Materials - Special Lifting Devices for Shipping Containers Weighing 10000 Pounds (4500 kg) or More

[5] NUREG-0612, July 1980 Control of Heavy Loads at Nuclear Power Plants U.S. Nuclear Regulatory Commission, Office for Nuclear Reactor Regulation List of References Page 34 of 34

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 f::0\GNS Rev.0

'<JI 14 Quality Assurance 14.0 Overview Prepared Reviewed 14.0 Overview Section 14.0, Rev. 0 Page 14.0-1

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev.0 All activities related to the design, fabrication and deployment of the CASTOR geo69 cask are performed under GNS Quality Assurance Program (OAP). An associated Quality Assurance Pro-gram Description (GNS-QAPD-001) is submitted to the NRC for approval under Docket-No. 71-0967.

The GNS OAP mainly consists of:

1. The Quality Assurance Manual (QAM II)

2. Subordinated Quality Assurance Procedures (QAM 11-P)

3. Project specific Quality Project Manuals (QPM).

- Activities performed by suppliers or subcontractors on behalf of GNS are governed by the suppliers

/ subcontractor approved quality assurance program or GNS QAP is extended to the supplier /

subcontractor. The amount and type of QA oversight depends on the importance to safety of the item or service to be procured and is based on the Graded Approach, which is described in GNS OAP.

14.0 Overview Section 14.0, Rev. 0 Page 14.0-2