Text

1014-SR-00002, Revision 0, Safety Analyses Report Dry Storage System Castor geo69. Part 2
	ML23124A283
Person / Time
Site:	07201052
Issue date:	04/21/2023
From:	; GNS Gesellschaft fur Nuklear-Service mbH
To:	; Office of Nuclear Material Safety and Safeguards
References
	T1213-CO-00018, EPID L-2021-NEW-0006 1014-SR-00002, Rev 0
	Download: ML23124A283 (1)
	v • d • e

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev. 0 @GNS 5 Shielding Evaluation 5.0 Overview 5.0 Overview Section 5.0, Rev. 0 Page 5.0-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS In this chapter, the shielding analysis of the CASTOR geo69 dual purpose cas~ (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 1.2.3 are used to distribute the SNF into of fuel with identical character-istics (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 sec-

-i tions that the homogeneous pattern TR1 with its bounding source terms covers the other two pat-terns in a sense of external dose rate.

Storage configuration of the cask, analysed in this chapter, coincides with its transport version be-sides the absence of the impact limiters in storage. An optional storage frame and a protection cover being an integral part of the DSS are conservatively not modelled and are 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-1536 [1] and NUREG-2215

[2]:

The radiation shielding features of the proposed dry storage system must be sufficient for it to meet the radiation dose requirements in 10 CFR 72.104. This is demonstrc1.ted by providing a shielding analysis of the surrounding dose rates that contribute to off-site :doses at appro-priate distances for a cask (typical array of casks in the most bounding site configuration) with bounding source terms for normal conditions of storage and anticipated occurrences (off-normal conditions).

Dry storage system contents and design features important for shielding are adequately de-scribed for evaluating 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 accident conditions for appropriate configurations of the cask and as-sumptions 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. 0 Page 5.0-2

Non-Proprietary Version 1014-SR-00002 Rev. 0 Proprietary Information withheld per 10CFR 2.390 s

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 contains 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 (identical to the source term defined in the corre-sponding SAR (transport) documentation [3])

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, a two-fold assessment of the dose at the site boundary is done:

5.0 Overview Section 5.0, Rev. 0 Page 5.0-3

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

A minimum distance to the site boundary is determined, where a most penalising array of the DSS complies with the radiation dose limit from 10 CFR 72.104

A minimum concrete wall thickness of the storage building is determined so that the radiation dose limit from 10 CFR 72.104 is met at a distance of 100 m The use of the CASTOR geo69 cask loading unit system (CLU, see section 1.2) including the trans-fer cask, the transfer lock and further equipment its contribution to the dose at the storage site boundary is neglected. A discus-sion 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 CASTOR geo69 CLU is presented 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-1536, Revision 1, Standard Review Plan for Spent Fuel Dry Storage Systems at a General License Facility, Office of Nuclear Material Safety and Safeguards, July 201 O

[2] NUREG-2215, Standard Review Plan for Spent Fuel Dry Storage Systems and Facilities, Office of Nuclear Material Safety and Safeguards, April 2020

[3] 1014-SR-00001, Rev. 0 Safety Analysis Report Type B(U)F Transport Package CASTOR geo69 Docket No.: 71-9383, 23.12.2020 5.0 Overview Section 5.0, Rev. 0 Page 5.0-4

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev.o v~GNS 5.1 Discussion and Results 5.1 Discussion and Results Section 5.1, Rev. 0 Page 5.1-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 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 effective 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 Figure 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. Different colours selected for the moderator rods are used to distinguish longer and shorter rods. 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. 0 Page 5.1-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS 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. 0 Page 5.1-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 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.

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. 0 Page 5.1-4

Non-Proprietary Version 1014-SR-00002 Rev. 0 Proprietary Information withheld per 10CFR 2.390 s

Axially the main shielding is provided by:

a cask bottom part consisting of thick DCI part ( bottom modera-tor plate, and a 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, 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. 0 Page 5.1-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @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. 0 Page 5.1-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Besides this geometry-independent mesh of detectors ********I),

The information about particular dose rates is gained from detectors positioned all around the cask.

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 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 The corre-sponding dose rate evaluations 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 accident conditions of storage there is a certain probability to completely lose the neutron moderator as a result of fire. It is shown, how-ever, 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 assumed.

The code system MCNP6 in version 2.0 [2] is used to calculate the dose rates.

The bounding source terms are not calculated anew, they are taken over from the SAR (transport)

[3], as the CASTOR geo69 is supposed to transport and store the very same contents. Furthermore, the shielding configurations for the transportation and storage do not differ considerably, it is only the presence of the impact limiter for the former one.

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 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 understand, where the boundary of the restricted area should be established, a bounding array of storage casks is investigated.

5.1 Discussion and Results Section 5.1, Rev. 0 Page 5.1-7

Non-Proprietary Version 1014-SR-00002 Rev. 0 Proprietary Information withheld per 10CFR 2.390

@ NS Figure 5.1-6 Bounding array of storage casks With respect to the configuration flexibility of the storage facility, 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 requirements of 10 CFR 72.104 al-ready at a distance of 100 m, and whether the requirements of 10 CFR 72.106 are fulfilled.

Apparently, none of the off-normal conditions have any impact on the shielding analysis. The only significant difference is that under off-normal conditions 1O % 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 Subpart C and D specify additional requirements for occupational dose limits and radia-tion 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) is of particular interest.

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 to skin or any extremity may not exceed 0.5 Sv (50 rem). The minimum distance from the spent 5.1 Discussion and Results Section 5.1, Rev. 0 Page 5.1-8

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS 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 (TR1) 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 (for details see section 5.4) 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.

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 shall generate the highest dose rate at the site bound-ary.

5.1 Discussion and Results Section 5.1, Rev. 0 Page 5.1-9

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 1 0CFR 2.390 Rev. 0 @GNS The maximum dose rate at the cask surface under normal (and off-normal) conditions is dominated by The total dose rate e.g. cold cask loaded accord-ing to TR1 sums up as follows: 1.47 mSv/h = - - - - - - - - - - - - -

w i t h the con-tributions from primary gamma radiation, neutrons including secondary gamma radiation, and hard-ware 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 2.435 hours 0.0145 weeks 0.00334 months 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 ). As one sees, the distant annual doses are not very different with a marginal excess a t

The neutron radiation share as well as the combined contribution of primary, secondary and 6°Co gamma radia-tion 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 at*

of controlled area on its long side. Almost

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

An alternative solution 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 5.1-9).

After an accident, when all moderator material is assumed to be lost, it is also assumed that no building is left around 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 at-distance (see Figure 5.1-10).

5.1 Discussion and Results Section 5.1, Rev. 0 Page 5.1-10

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev.a @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. 0 Page 5.1-11

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @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 / top) bottom / top) 1.47 1.55 0.042 0.041 TR1

( 6.29) ( 0.324)

TR2 1.00 I 1.05 - 0.050 I 0.050 -

( 7.02) - ( 0.370) -

TR3 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 fron;i 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)

TR2 0.133 I 0.127 - 0.017 I 0.017 -

( 2.57) - ( 0.129) -

TR3 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 tor 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) 0.093 0.090 0.009 0.009 TR1

( 1.22) ( 0.072)

TR2 0.076 I 0.071 - 0.013 I 0.011 -

( 1.35) - ( 0.083)

TR3 0.078 I 0.073 - 0.013 I

( 0.079) 0.011 -

( 1.28) -

5.1 Discussion and Results Section 5.1, Rev. 0 Page 5.1-12

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Figure 5.1-9 Annual dose from the storage building as a function of distance Figure 5.1-10 Annual dose from the unshielded storage cask array under accident storage conditions 5.1 Discussion and Results Section 5.1, Rev. 0 Page 5.1-13

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @)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 and the storage building 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] 1014-SR-00001, Rev. 0 Sat ety Analysis Report Type B(U)F Transport Package CASTOR geo69 Docket No.: 71-9383, 23.12.2020 5.1 Discussion and Results Section 5.1, Rev. 0 Page5.1-14

___J

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev.a 5.2 Source Specification Prepared Reviewed 5.2 Source Specification Section 5.2, Rev. O Page 5.2-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS The procedure how the bounding source terms for each loading pattern TR1 to TR3 are determined is described in SAR (transport) [1] in detail. For a sake of completeness, the most important proper-ties of the radioactive contents are summarised here.

Each basket position is characterised by the maximum allowed decay heat power (see section 1.2.3).

Within one chosen position group (see Figure 5.0-1) all the position have identical requirements. The minimum cooling times required to reach a certain decay heat for a particular loading pattern are presented in Table 5.2-1. Some field are left vacant, because the cooling times are smaller than the minimum cooling times for this particular SNF type (see section 1.2.3).

Table 5.2-1 Minimum cooling times (in years) needed to reach certain decay heat 5.2.1 Gamma Source The gamma-radiation fuel is analysed in seven energy groups (see Table 5.2-2), 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 significantly 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.

5.2 Source Specification Section 5.2, Rev. 0 Page 5.2-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 Table 5.2-2 Gamma energy structure Energy group, i = 1 2 3 4 5 6 7 Average group energy, MeV 0.575 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 Besides gamma particles stemming from the fuel pellet stack, there is also a gamma radiation from activated hardware: the end fittings and plenum springs. The primary source of activity in the non-59 fuel regions of a SNF arise from the activation of Co to 6°Co. The activities of 6°Co are determined during burn-up and depletion calculation together with the calculations for the primary gamma radi-ation. The flux scaling factors of 0.1 for the top end fittings and of 0.2 for the bottom end fittings and

- plenum springs are applied according to PNL-6906 [2].

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 terms are reported in Table 5.2-3 for TR1, in Table 5.2-4 for TR2, and in Table 5.2-5 for TR3 decoded after position groups. The sources are normalised to one megagram of heavy metal to account for the differences in the mass of the SNF including the mass in the shielding model.

Table 5.2-3 Bounding gamma source term for TR1 5.2 Source Specification Section 5.2, Rev. 0 Page 5.2-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @)GNS Table 5.2-4 Bounding gamma source term for TR2 Table 5.2-5 Bounding gamma source term for TR3 The total source strength in each gamma energy group is calculated by summation of the 24 axial nodal values stemming from the conservative axial burn-up profile discussed above. The corre-sponding 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.

5.2 Source Specification Section 5.2, Rev. 0 Page 5.2-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Figure 5.2-1 Axial gamma source strength distributions 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). The bound-ing source terms used for the shielding analysis are reported in Table 5.2-6 for TR1, in Table 5.2-7 for TR2, and in Table 5.2-8 for TR3.

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 [3]). While the energy spectrum 244 from the spontaneous fission could be nicely described by the Watt spectrum from Cm (parame-1 ters a= 0.902523 MeV, b = 3.72033 Mev- , see [3]), 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 5.2-9).

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

5.2 Source Specification Section 5.2, Rev. 0 Page 5.2-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Figure 5.2-2 Axial neutron source strength distributions Table 5.2-6 Bounding neutron source for TR1 Table 5.2-7 Bounding neutron source for TR2 Table 5.2-8 Bounding neutron source for TR3 5.2 Source Specification Section 5.2, Rev. 0 Page 5.2-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Table 5.2-9 Bounding (a,n)-reaction neutron source as a function of energy The largest total neutron source per storage cask stems from TR2 and accounts to neu-trons/s and corresponds to complete loading with the bounding source.

List of References

[1] 1014-SR-00001, Rev. 0 Safety Analysis Report Type B{U)F Transport Package CASTOR geo69 Docket No.: 71-9383, 23.12.2020

[2] 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

[3] 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. 0 Page 5.2-7

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

'e 5.3 Model Specification Section 5.3, Rev. o Page 5.3-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev.a 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. 0 Page 5.3-2

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev.0 @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 d,ose rate limit of 0.02 mSv/h (2 mrem/h) according to 10 CFR 20.1301.

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 % (1 0 %) 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 1O %-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. 0 Page 5.3-3

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @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 section 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. 0 Page 5.3-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 1 OCFR 2.390 Rev. 0 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, Atrium-1 OA design is selected for the shielding model. The main reason for this is the lattice configuration. With the large water channel and part length rods the self-shielding effects are minimised. Additionally, this fuel design has the thinnest cladding and fuel channel thickness among other SNF.

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 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. 0 Page 5.3-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 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 5.3.1.4 Canister 5.3 Model Specification Section 5.3, Rev. 0 Page 5.3-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS 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, SA-479M 414) or DCI (lower tilting studs), the corresponding flattenings of the cask body possess minimum dimensions.

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 Model Specification Section 5.3, Rev. 0 Page5.3-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 1 OCFR 2.390 Rev. 0 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.

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 5.3 Model Specification Section 5.3, Rev. 0 Page 5.3-8

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

The design basis for the material data is given in chapter 8.

5.3 Model Specification Section 5.3, Rev. 0 Page 5.3-9

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Table 5.3-1 Material properties in the shielding model 5.3 Model Specification Section 5.3, Rev. 0 Page 5.3-10

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev.a @GNS Table 5.3-1 Material properties in the shielding model (cont.}

5.3 Model Specification Section 5.3, Rev. 0 Page 5.3-11

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @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] 1014-SR-00001, Rev. 0 Safety Analysis Report Type B(U)F Transport Package CASTOR geo69 Docket No.: 71-9383, 23.12.2020 5.3 Model Specification Section 5.3, Rev. 0 Page 5.3-12

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev. O @GNS 5.4 Shielding Evaluation Name, Function Date Signature Prepared Reviewed 5.4 Shleldlng Evaluatlon Section 5.4, Rev. O Page 5.4-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 1 OCFR 2.390 Rev. 0 @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/8-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 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 TR1.

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. 0 Page 5.4-2

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

@ NS Table 5.4-1 Message digest overview Calculation Series Model Message Digest, mdS Standalone storage cask Array of storage casks 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. 0 Page 5.4-3

Non-Proprietary Version.

1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS 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 mSv/h/(y/cm *s) mSv/h/(y/cm *s) 0.01 3.96E-05 1.4 2.51 E-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 6.31 E-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 1.0SE-05 5.75 6.37E-05 0.5 1.17E-05 6.25 6.74E-05 0.55 1.27E-05 6.75 7.11 E-05 0.6 i.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 mSv/h/(n/cm2 *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 iO 1.47E-03 14 2.08E-03 20 2.27E-03 5.4 Shielding Evaluation Section 5.4, Rev. 0 Page 5.4-4

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @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, the irregularities vanish (see Figure 5.4-4, right).

5.4 Shielding Evaluation Section 5.4, Rev. 0 Page 5.4-5

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Table 5.4~4 Maximum external dose rates for the cold storage cask 5.4 Shielding Evaluation Section 5.4, Rev. 0 Page5.4-6

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 (@)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. 0 Page 5.4-7

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS 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. 0 Page 5.4-10

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS 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. 0 Page 5.4-8

Non-Proprietary Version 1014-SR-00002 Rev. 0 Proprietary Information withheld per 10CFR 2.390

@ NS 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).

For the far distances, the cold case is bounding. For that reason, all the storage site calculations are performed with the cold cask model. The dose rate distributions from the hot storage cask are fairly similar to those from the cold one, therefore they are not pre-sented here.

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 have been adjusted accordingly by selecting maximum source terms among all possible combinations of fuel assemblies after 20 years (see Table 5.2-1 and [6] for de-tails). The redistribution of the fuel does not affect the shell dose rate in a negative way. The 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.

5.4 Shielding Evaluation Section 5.4, Rev. 0 Page 5.4-9

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 1 0CFR 2.390 Rev. 0 @GNS 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

- 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. 0 Page 5.4-11

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev.a (@)GNS Table 5.4-7 Maximum external dose rates for design basis accident 5.4 Shielding Evaluation Section 5.4, Rev. 0 Page 5.4-12

Non-Proprietary Version 1014-SR-00002 Rev. 0 Proprietary Information withheld per 10CFR 2.390 s

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. 0 Page 5.4-13

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS 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 2.435 hours 0.0145 weeks 0.00334 months is assumed.

(see Table 5.4-10). The neutron dose equivalent is mainly due to the spon-taneous fission, the major gamma radiation contribution comes from the fuel gammas at energies between Table 5.4-1 0 Relative contributions to the annual dose at 100 m from the storage cask 5.4 Shielding Evaluation Section 5.4, Rev. 0 Page 5.4-14

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 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 For this reason, the distance .to the nearest boundary of the controlled area has to be larger. Alternatively, a deployment of a storage building made of e.g. concrete is possible.

Assuming a bounding array of - 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 away 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. For easier understanding the line representing a dose rate limit 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 of 100 m, a building ***********************is required around the entire array of the storage casks (see Figure 5.1-9). For this scenario, the annual do*se 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 all-storage casks have lost their neutron moderators completely. The upright arrangement as discussed in section 0 is retained. For the duration of acci-dent conditions 30 days are implied. The resulting annual dose equivalents ar~ presented in Figure 5.1-10. It can be clearly seen that the limiting total effective dose equivalent of 50 mSv according to 10 CFR 72.106 is safely complied with.

5.4 Shielding Evaluation Section 5.4, Rev. 0 Page 5.4-15

Non-Proprietary Version 1014-SR-00002 Rev. 0 Proprietary Information withheld per 10CFR 2.390

@ NS 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 Insti-tute. American national standard neutron and gamma-ray flux-to-dose-rate factors, La Grange Park, 111.: 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] 1014-SR-00001, Rev. O Safety Analysis Report Type B(U)F Transport Package CASTOR geo69 Docket No.: 71-9383, 23.12.2020 5.4 Shielding Evaluation Section 5.4, Rev. 0 Page 5.4-16

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 1014-SR-00002 Rev. 0 5.5 Appendix Prepared Reviewed 5.5 Appendix Section 5.5, Rev. o Page 5.5-1

Non-Proprietary Version 1014-SR-00002 Proprietary Information withheld per 10CFR 2.390 Rev. 0 @GNS Appendix 5-1: Shielding Model for Cask Loading Unit Appendix 5-2: Sample Input File for MCNP 5.5 Appendix Section 5.5, Rev. 0 Page 5.5-2

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 Appendix 5-1 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, while 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 5.5-1. The cask itself is con-servatively modelled without transfer cask lid. The basket is drained, both water chambers are filled Conservatively, the contents of the canister are assumed to be loaded according to the uniform loading pattern (TR1 ). Typical dose rate distributions on the shell surface of the transfer cask and in 1 m from the cask are displayed in Figure 5.5-2.

- 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 5.5-3) 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 1 of 3

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 Figure 5.5-1 Shielding model of the transfer cask Figure 5.5-2 Dose rate distributions (in mSv/h) on the shell surface (left) and in 1 m distance from the transfer cask (right)

Page 2 of 3

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 Figure 5.5-3 Shielding model of the canister transfer Page 3 of 3

Non-Proprietary Version Proprietary Information withheld per 10CFR 2.390 Appendix 5-2 Sample Input File for MCNP

Page 1 of 23