December 2012

Working Reports contain information on work in progress

or pending completion.

The conclusions and viewpoints presented in the report

are those of author(s) and do not necessarily

coincide with those of Posiva.

Aapo Tanskanen

Fortum Power and Heat Oy

Working Report 2012-85

Supplementary Radiation ShieldingCalculations for the Final Disposal Facility

SUPPLEMENTARY RADIATION SHIELDING CALCULATIONS FOR THE FINAL DISPOSAL FACILITY ABSTRACT Radiation shielding analyses were made using the MCNP 5 Monte Carlo code. The study focused on the following cases: the impact of the dry spent fuel transfer on dimensioning of the ceiling of the encapsulation plant's handling cell, entrance mazes of the canister transfer corridor, radiation shielding for the weld inspection device, canister storage in the repository, radiation shielding of the canister transfer and installation vehicle, and radiation doses near a canister in the disposal hole. According to the calculation results radiation shielding standpoint has been sufficiently taken into consideration in design of the encapsulation plant and final disposal facility, and the target values for the maximum allowed dose rate levels can be achieved. In some cases further design work is required in order to achieve the target values. For example shielding of the driver's cabin of the canister transfer vehicle appears challenging. Thus, involvement of a radiation shielding specialist in the design process can be recommended. In most cases gamma radiation is the dominant dose contributor, but in specific configurations neutron radiation can become significant for the overall dose rate. This occurs for canister storages where the amount of spent fuel is large and thick concrete shields or entry mazes are used for radiation protection. Another finding of the study is that the burnup distribution of the spent fuel needs to be taken into account when assessing the yield of the neutron radiation source, because use of the assembly average burnup leads to underestimation of it. The report includes an assessment of the neutron induced activation at the encapsulation plant and the repository of spent nuclear fuel. Keywords: Final disposal of spent nuclear fuel, radiation shielding.

TÄYDENTÄVIÄ SÄTEILYSUOJAUSLASKUJA LOPPUSIJOITUSLAITOKSELLE TIIVISTELMÄ Työssä tehtiin säteilysuojauslaskuja käyttäen Monte Carlo -menetelmään perustuvaa MCNP 5 -ohjelmaa. Tutkimuksessa tarkasteltiin seuraavia tapauksia: käytetyn poltto-aineen siirtosäiliön kuivakuljetuksen vaikutus kapselointilaitoksen käsittelykammion katon mitoitukseen, kapselin siirtokäytävän labyrinttirakenteet, hitsisauman tarkastus-laitteiston säteilysuojaus, maanalainen kapselivarasto, kapselin siirto- ja asennusajo-neuvon säteilysuojaus ja säteilyannokset loppusijoitusreikään sijoitetun kapselin lähei-syydessä. Säteilysuojauslaskujen tulosten perusteella voidaan sanoa, että säteilysuojausnäkökulma on otettu riittävällä tavalla huomioon kapselointi- ja loppusijoituslaitoksen suunnitte-lussa ja tavoitteeksi asetetut annosnopeuden enimmäistasot on mahdollista saavuttaa. Joissakin tapauksissa tarvitaan lisää suunnittelutyötä tavoitearvojen saavuttamiseksi. Esimerkiksi kapselin siirto- ja asennusajoneuvon ohjaamon säteilysuojaus vaikuttaa haastavalta. Siksi säteilysuojausasiantuntijan osallistumista suunnitteluprosessiin voi-daan suositella. Useimmissa tapauksissa gammasäteily on dominoiva säteilyannoksen aiheuttaja, mutta tietyissä tapauksissa neutronisäteily voi nousta merkittäväksi annosnopeuden kannalta. Näitä ovat mm. kapselivarastot, joissa käytetyn polttoaineen määrä on suuri ja säteilysuojaukseen käytetään paksuja betonirakenteita sekä labyrinttejä. Lisäksi on syytä huomata, että neutronilähteen antoisuuden arvioinnissa tulee ottaa huomioon käytetyn polttoaineen palamajakautuma, koska keskimääräisen nippupalaman käyttö johtaa lähteen antoisuuden aliarviointiin. Raportin liitteenä on arvio neutronisäteilyn aiheuttamasta materiaalien aktivoitumisesta kapselointilaitoksella ja loppusijoitustilassa. Avainsanat: Käytetyn polttoaineen loppusijoitus, säteilysuojaus

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TABLE OF CONTENTS ABSTRACT TIIVISTELMÄ

1  INTRODUCTION ................................................................................................... 2 

2  METHODS OF ANALYSIS .................................................................................... 3 2.1  General ........................................................................................................ 3 2.2  Radiation source terms ................................................................................ 3 

3  RADIATION SHIELDING ANALYSES................................................................... 5 3.1  Dry VVER transfer cask and dimensioning of the handling cell ceiling ........ 5 

3.1.1  Dose rate in the decontamination centre .......................................... 5 3.2  The front and back end of the canister transfer corridor .............................. 6 

3.2.1  The front end of the canister transfer corridor .................................. 6 3.2.2  The back end of the canister transfer corridor .................................. 7 

3.3  Radiation shielding for the weld inspection device ....................................... 8 3.3.1  Required shield for the radiation beam ............................................. 8 3.3.2  Required shielding for the scattered radiation .................................. 9 3.3.3  Photoneutron production ................................................................ 10 

3.4  Canister storage for the final disposal canisters in the repository .............. 11 3.4.1  Dose rate above the canister storage ............................................. 11 3.4.2  Dose rate in the maintenance gap .................................................. 13 3.4.3  Dose rate near the canister hoist .................................................... 14 3.4.4  Dose rate in the upper hall during a canister transfer by lift ........... 16 

3.5  Radiation shield of the canister transfer and installation vehicle ................ 18 3.5.1  Previous studies ............................................................................. 18 3.5.2  Axial source distribution .................................................................. 19 3.5.3  Azimuthal dose rate distribution ..................................................... 20 3.5.4  Maximum dose rate as a function of the distance from the radiation shield .............................................................................................. 22 3.5.5  Rear radiation shield ....................................................................... 23 3.5.6  Radiation shield end ....................................................................... 25 3.5.7  Radiation shielding of the driver's cabin ......................................... 26 

3.6  Disposal canister in the disposal hole ........................................................ 30 3.6.1  Dose rate in the disposal tunnel ..................................................... 30 

4  CONCLUSIONS .................................................................................................. 32 

APPENDIX I .................................................................................................................. 36  Assessment of the neutron induced activation at the encapsulation plant and the repository of spent nuclear fuel

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1 INTRODUCTION

The spent nuclear fuel from the reactors in Loviisa and Olkiluoto is stored in water pools at the plant sites. After some tens of years of interim storage the spent nuclear fuel is transported to Posiva's encapsulation plant in Olkiluoto, where the fuel bundles are dried and placed into the disposal canisters. The disposal canisters are closed by welding a copper cover using an electron beam welding device. The quality of the weld is planned to be inspected with various methods including radiographic imaging using a high-energy X-ray linear accelerator. The canisters are transported from the encapsulation plant by a lift to canister storage in the underground repository from where the canisters are fetched with a disposal canister transport vehicle to be placed in deposition holes bored in bedrock and to be sealed with bentonite. While the decay heat and radioactivity of the spent nuclear fuel has diminished substantially during intermediate storage, the spent nuclear fuel is still highly radioactive and emits both gamma and neutron radiation. Anttila (2005) has calculated the gamma and neutron source terms of the various spent nuclear fuel types. The radiation shielding analyses related to final disposal of spent nuclear fuel have been carried by Anttila (1998) and Ranta-Aho (2008), and the results have been used for design and dimensioning of the radiation shields of the final disposal facility. The radiation shielding of the encapsulation plant and the final disposal repository is based mainly on use of concrete structures and doorless entry mazes. The purpose of this study is to supplement the previous radiation shielding studies as the layout of the encapsulation plant has been updated and for some systems the detailed design information has only recently become available. These include the weld inspection device, the canister storage for spent fuel canisters in the final disposal repository, and the radiation shield of the canister transfer and installation vehicle. The report can not be considered a complete, standalone report, but the reader is advised to refer to the previous study by Ranta-Aho (2008). Thus, the report may appear fragmentary as it is simply a collection of the results of the separate radiation shielding studies. It should be noted that the calculation results are presented in an order that is in accordance with the final disposal process steps.

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2 METHODS OF ANALYSIS

2.1 General

Monte Carlo calculations with MCNP 5-1.40 code were carried out in order to assess the dose rates caused by gamma and neutron radiation of the spent nuclear fuel. The source terms of the shielding analyses were the same as those used by Ranta-Aho (2008). They are based on the ORIGEN-S calculations of Anttila (2005), where it was assumed that the burnup of the spent fuel was 60 MWd/kgU and it had been stored for 20 years after reactor use. Most of the shielding analyses were carried out assuming BWR type fuel as it was found by Ranta-Aho (2008) that the BWR type final disposal canister in most cases determines he required radiation shielding structures. The shielding structures were assumed to be made of unreinforced concrete (material M200 described by Ranta-Aho 2008), which is a conservative assumption. In the shielding calculations the material composition of the spent nuclear fuel was assumed to be that of the fresh nuclear fuel. This simplification has only a small effect on the gamma shielding as the spent nuclear fuel is mostly uranium (some 95 weight-%) even after reactor use. The dose rates were calculated using the ICRP74 gamma and neutron flux-to-dose conversion coefficients (ICRP, 1996). Variance reduction techniques were used extensively in order to increase the efficiency of the Monte Carlo calculations.

2.2 Radiation source terms

In gamma radiation shielding calculations the main source nuclides were Ba-137m (daughter nuclide of Cs-137) and Eu-154. The gamma source spectrum used in calculations was similar to that used by Ranta-Aho (2008). The dominant neutron source of a spent fuel after 20 years cooling is spontaneous fission of Cm-244 that was modelled to have an energy spectrum described by the Watt fission spectrum. The fission neutrons can in principle cause secondary fissions in the fuel region. The disposal canister is a subcritical system, where the primary neutrons originating from the spontaneous fission of Cm-244 may cause fission reactions and secondary neutrons. The material composition of the spent nuclear fuel was conservatively assumed to be that of the fresh nuclear fuel, which leads to exaggeration of the secondary fission neutrons. There is an option in MCNP that the secondary fission neutrons can be prevented, and some additional calculations were made using this option. The results showed that for the disposal canister geometry with the fresh fuel assumption the secondary neutrons account for some 20 % of the total dose rate. The finding is in line with the calculation results of Anttila (1996). The reactivity of the spent nuclear fuel is lower than that of fresh fuel, and thus the results obtained assuming fresh fuel are conservative. Neutrons can also generate secondary gamma radiation when high energy neutrons hit heavy target nuclei. The dose rate caused by secondary gamma radiation was calculated in some cases. According to the calculation results, the secondary gamma radiation is not a major radiation shielding issue for spent nuclear fuel. The axial burnup distribution has an effect on the axial distribution of the radiation source. Because the dominant gamma emitter of the 20 year cooled spent nuclear fuel is a fission product, the axial gamma source distribution is similar to the burnup

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distribution. However, the build-up of the Cm-244 nuclide as a function of burnup is a nonlinear process, and thus, the axial neutron source distribution has a bigger variance than the burnup distribution. Furthermore, estimation of the neutron source using the bundle average burnup leads to underestimation of the neutron source, and the real neutron source can be as much as 30 % larger than the one obtained assuming a uniform burnup profile. The issue is discussed in detail in Section 3.5.1. Even though in general neutron irradiation may induce activity in exposed materials, the neutron flux densities at the encapsulation plant and in the underground repository are so low that the neutron-induced radiation sources are insignificant from the radiation protection point of view. This issue is discussed in Appendix I.

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3 RADIATION SHIELDING ANALYSES

3.1 Dry VVER transfer cask and dimensioning of the handling cell ceiling

Previously the VVER type fuel from the Loviisa NPP has been planned to be transported with a transport cask that fits 84 fuel bundles and is filled with water (i.e. wet transfer). Now an alternative option is being studied where the fuel is transported in a dry cask. As the water fill of the transfer cask serves as a radiation shield, there is a need to check that the concrete ceiling of the handling cell of the encapsulation plant is sufficiently thick so that that the dose rate in the decontamination centre does not exceed 2.5 µSv/h. The dimensioning of the handling cell wall is based on MCNP calculations where the source term is an EPR fuel bundle. The analyses were made for four different concrete types and also the thickness of the wall was varied from 110 to 130 cm. According to MCNP calculations of Ranta-Aho (2008) a 120 cm thick concrete wall decreases the gamma and neutron radiation dose rate by a factor of 2.4E8 and 2.0E5, respectively. Using the dose rates calculated by Ranta-Aho (2008) on top of the VVER assembly (60 MWd/kgU burnup, 20 years cooling), it can be assessed that the dose rates caused by 84 VVER fuel bundles of this type would be less than 0.35 µSv/h for primary gamma radiation and 0.11 µSv/h for neutrons. The summary dose rate (0.46 µSv/h) is well below the 2.5 µSv/h level. Furthermore, because the distance between the top of the transfer cask and the handling cell ceiling is several meters the dose rates in the decontamination centre are further alleviated.

3.1.1 Dose rate in the decontamination centre

A MCNP model of the handling cell with a dry transfer cask connected to it was made. The source term was modelled as a single VVER fuel bundle in the centre of a steel cylinder. The fuel was conservatively assumed to have a uniform 60 MWd/kgU burnup and 20 years cooling after reactor use. The source term photon and neutron sources were set to correspond to those of 84 similar fuel assemblies. The handling cell was assumed to have a 120 cm thick ceiling made of M200 type concrete (Ranta-Aho 2008). The calculated dose rates corresponding to the primary gamma radiation, neutrons and secondary gamma radiation are reported in Table 1. The calculated total dose rate is less than 0.2 µSv/h, and there is no need to modify the thickness of the handling cell ceiling in order to enable the dry transport option. Table 1. Dose rates in the decontamination centre when a dry VVER transport cask is connected to the handling cell.

Dose Component Dose Rate (µSv/h)

Primary gamma 0.138 ± 0.008Neutron 0.034 ± 0.001Secondary gamma 0.003 ± 0.000Total 0.174

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3.2 The front and back end of the canister transfer corridor

The final disposal canister moves along the canister transfer corridor during the encapsulation process. The thickness of the transfer corridor walls, floor and ceilings were confirmed to provide sufficient radiation shielding by Ranta-aho (2008). The entry mazes of the transfer corridor were also studied, but because of the revised layout of the encapsulation plant there was a need to re-evaluate the radiation shielding situation in the front and back end of the canister transfer corridor. The shielding analyses were made using a simplified model of the VVER type disposal canister. This canister type was selected because in the front end of the corridor the canister top is open and the shortest canister type causes maximum dose rates in the entry maze. The same source term was used for studying the back end of the corridor, even though the maximum dose rates would be caused by the BWR type canister. In the MCNP model the contents of the disposal canister was homogenized which results in conservative dose rate estimates on the canister surface. The calculated gamma dose rate on surface of the canister was 240 mSv/h which is 45 % higher than the dose rate calculated by Ranta-aho (2008) using a detailed VVER canister model. For neutron radiation the dose rate on the surface of the canister was of the same order as that obtained by Ranta-aho (2008).

3.2.1 The front end of the canister transfer corridor

In Figure 1 is shown the MCNP geometry model used to study the adequacy of the front-end entry maze as a radiation shield. The canister transfer corridor is a high room, while the adjacent room is divided in two floors. The slab between the two levels was not included in the model, which is a conservative simplification. The disposal canister (red circle) is located under the handling cell docking station in the canister transfer corridor. There is no need to transfer canisters filled with spent fuel closer to the front end of the corridor. The dose rates were calculated on both floors at positions marked with "D" and "E" in Figure 1. The calculated dose rates are shown in Table 2. The results show that the dose rates in the front-end entry maze remain clearly below 25 µSv/h.

Figure 1. The MCNP geometry model of the front end of the canister transfer corridor. The red arrow shows the path of the canister in the corridor.

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Table 2. Calculated dose rates in the front-end entry maze of the canister transfer corridor. The dose rate calculation points are marked in Figure 1.

Lower floor level +1.90

Upper floor level +6.10

Dose rate calculation point

D (µSv/h)

E (µSv/h)

D (µSv/h)

E (µSv/h)

Gamma 6.8 2.2 7.7 2.5 Neutron 1.9 0.75 2.1 0.84 Total 8.7 2.9 9.8 3.3

3.2.2 The back end of the canister transfer corridor

In Figure 2 is shown the MCNP geometry model used to study the radiation shielding adequacy of the canister transfer corridor back-end shielding structures. Two different cases were studied: in the first case the disposal canister is situated in front of the elevator door (red circle in Figure 2), in the second case the canister is in the transfer corridor at position marked with "C" in Figure 2. The dose rates were calculated in the entry maze and at the transfer vehicle docking station, the points marked with "A" and "B" in Figure 2, respectively. The calculated dose rates are shown in Table 3. According to the calculation results the dose rates in the entry maze ("A") are small in both cases. The dose rate at the transfer vehicle docking station ("B") is elevated only in case the canister is at position "C". However, the calculated dose rate of 64 µSv/h allows a short-term attendance for maintenance of the vehicle.

Figure 2. The MCNP model of the back end of the canister transfer corridor. The red arrow shows the path of the canister in the corridor.

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Table 3. Calculated dose rates at the selected positions of the back-end of the canister transfer corridor. The dose rate calculation points are marked in Figure 2.

Canister location Canister in front of the elevator door

(red circle in Figure 2)

Canister in the corridor (mark "C" in Figure 2)

Dose rate calculation point

B (µSv/h)

A (µSv/h)

B (µSv/h)

A (µSv/h)

Gamma 56.2 0.28 1.85 0.36 Neutron 7.7 0.13 0.87 0.18 Total 64 0.41 2.7 0.54

3.3 Radiation shielding for the weld inspection device

A high energy X-ray linear accelerator is planned to be used for inspection of the disposal canister lid weld (Sandelin 2010). The current plans are based on use of the Varian Mi-9 Linatron or an equivalent radiation source. The acceleration voltage is 9 MV meaning that the electrons that hit the target produce X-rays that have a kinetic energy of 9 MeV. According to Müller et al. (2006) the accelerator produces a beam of photons whose energy can be up 8 MeV, but the median source photon energy is about 3 MeV. According to Varian Medical Systems (2007) the maximum dose rate of Mi-9 Linatron is of the order of 30 Gy/min (1800 Gy/h). There is a STUK's regulatory guide ST 5.6 on radiation safety in industrial radiography. As the weld inspection device is located in a separate room it can be considered as a closed installation. According to the regulatory guide ST 5.6 the dose rate outside of the weld inspection room shall be smaller than 7.5 µSv/h.

3.3.1 Required shield for the radiation beam

According to Varian Medical Systems (2007, Table 3-2 on page 10) a typical half value layer in concrete for the Varian Mi-9 Linatron is 11 cm. Requiring that the dose rate outside of the weld inspection room shall be smaller than 7.5 µSv/h it can be estimated that 3.1 meters of concrete is needed in order to attenuate the intensive photon beam. The estimate was checked with a MCNP model. The MCNP model consisted of a monodirectional point source hitting a concrete block. The photon energy spectrum determined by Müller et al. (2006) was used and the source photon source was adjusted so that the dose rate before the concrete block was set to 1800 Gy/h. In Figure 3 is shown the dose rate along the photon beam in the primary radiation shield made of concrete. According to the calculation results the concrete block should be at least 2.9 meters thick in order to ensure that the dose rate on the outer surface of the weld inspection room wall is smaller than 7.5 µSv/h. The result is in accordance with the simple estimate.

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Figure 3. The calculated dose rate along the photon beam inside the concrete block. In Figure 4 is shown the calculated dose rate distribution inside the concrete radiation shield. The y-axis of the figure corresponds to the radiation beam while the x-axis is the radial distance from the beam. If the weld inspection room walls had a thickness of 80 cm, the radius of the primary radiation shield would not have to be much bigger than the beam itself. However, there is a need to consider also the scattering radiation.

Figure 4. Calculated dose rate distribution inside the primary radiation shield.

3.3.2 Required shielding for the scattered radiation

During the weld inspection with linear accelerator the photon beam hits the copper target and causes scattered radiation. The dose rate distribution caused by scattered radiation was studied with a MCNP model where the photon beam was set to hit a

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cylindrical copper target whose radius and length were 0.5 and 10 cm, respectively. The resulting dose rate distribution is shown in Figure 5. The figure shows that the scattering radiation peaks forward at relatively small scattering angles. The forward-directed scattering radiation can in principle be taken care of by increasing the size of the primary radiation shield. Especially the height of the shield may not be sufficient. The shields on the side of the target (walls of the room) are not planned to be as thick as the primary radiation shield. According to Figure 5 the dose rate caused by scattered radiation on the side of the copper target at a distance of 1 m can be as high 1 mSv/h. Thus, there is a need to attenuate the scattered radiation by a factor of 133 in order to reach the limiting value of 7.5 µSv/h. The 80 cm thick concrete walls, floor and ceiling of the weld inspection room provide sufficient radiation shielding against the scattered radiation.

3.3.3 Photoneutron production

Some of the photons produced by Varian Mi-9 Linatron have a sufficient energy to produce photoneutrons. The threshold energy and cross section for the production of photoneutrons depend on the nuclide being hit by the energetic photons (IAEA 2000). The target of the linatron, which absorbs high energy electrons and produces gamma radiation, is typically made of tungsten for which the threshold energy for production of photoneutrons is 6.19 MeV. For copper the threshold energy is so high that the final disposal canister will not emit photoneutrons. The production of photoneutrons was not estimated in this study as the standard version of the MCNP program is not suitable for that. The production of photoneutrons can be prevented by avoiding the use of materials with low threshold energy for photoneutron production such as lead, tantalum, tungsten or uranium. In general the production of photoneutrons is likely to be low and the radiation shielding of the weld inspection room should be adequate. However, the photoneutrons may affect the service life of the electronic instrumentation of the weld inspection room. Neutron activation is likely to be measurable but insignificant for radiation protection.

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Figure 5. The dose rate pattern caused by scattered radiation when the photon beam hits a copper cylinder.

3.4 Canister storage for the final disposal canisters in the repository

The underground repository is planned to have canister storage for the final disposal canisters. The planned canister storage capacity is 30 canisters. The canister storage is planned to be an elongated concrete structure inside a rock excavated tunnel. A maintenance gap is planned between the canister storage and the tunnel wall. The length and width of the tunnel are about 60 and 10 meters, respectively. Taking into account the maintenance gap and the canister storage wall, the inner width of the canister storage is of the order of 6 meters. The inner height of the canister storage varies; the minimum height 7.39 m was used in the shielding calculations. The canister storage has a connection to the canister lift and the disposal canisters are stored in the far end of the elongated storage room. The canister storage hosts some additional functions and thus, there is a need to ensure the possibility to work in the front part of the canister storage without time limitations. Thus, the dose rate above the canister storage and in the front part of the canister storage room should be below 3 µSv/h. This is achieved with a doorless entry maze that needs to be sufficiently spacious considering the transfer of the spent fuel canisters with the automated guided vehicle. There is no need for the plant personnel to permanently stay in the maintenance gap, but it is reasonable to require that the radiation dose rate in the maintenance gap shall stay below 25 µSv/h.

3.4.1 Dose rate above the canister storage

The dose rate above the canister storage was calculated with a MCNP model including a single EPR type final disposal canister, 80 cm thick ceiling and wall made of M200 type concrete (Ranta-Aho, 2008). The EPR type canister was selected because it is the tallest

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canister type. The MCNP model, shown in Figure 6, was made to represent an infinitely long canister storage, where the canisters are placed every 1.6 meters on both sides of the storage room, by setting the three model boundaries reflective. A homogenized model of the EPR fuel bundles was used. Axial burnup distribution was not taken into account, i.e. source terms were assumed axially uniform, which is a conservative assumption.

Figure 6. MCNP model used for calculation of the dose rate in the room above the canister storage. The dose rates in the maintenance gap between the canister storage wall and the rock tunnel were calculated using the MCNP variance reduction techniques. Secondary gamma radiation doses were not calculated as they were considered negligible. The calculated gamma and neutron dose rates in the room above the canister storage are shown in Table 4. As expected the dose rates were found small. It is interesting that the dose rate in the room above the canister storage is mainly caused by neutron radiation. This occurs because concrete is a more efficient radiation shield against gamma radiation than against neutron radiation, and because neutrons from several disposal canisters contribute to the dose rate above the canister storage. Table 4. Dose rates in the room above the canister storage.

Dose component Dose rate (µSv/h)

Primary gamma 0.102 ± 0.010Neutron 0.218 ± 0.015Total 0.320

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3.4.2 Dose rate in the maintenance gap

The dose rates in the maintenance gap between the canister storage wall and the rock tunnel were evaluated using a MCNP model shown in Figure 7. The model included a single BWR type final disposal canister, 80 cm thick concrete ceiling and wall made of M200 type concrete and three reflective boundaries. The BWR type canister was selected because it has the highest dose rate on the surface of the canister (Ranta-Aho 2008). The model represents a infinitely long canister storage, where the canisters are placed every 1.6 meters on both sides of the storage room. A homogenized model of the BWR fuel bundles was used. Axial burnup distribution was not taken into account, i.e. source terms were assumed axially uniform. This leads to some underestimation of the calculated dose rates. This is partly caused by the fact that the real source term is more concentrated in the middle part of the fuel column than assumed. However, this effect diminishes as the distance between source and the point of interest increases or there is a thick radiation shield in between. In case of neutrons the assumption of uniform axial burnup distribution leads to an additional underestimation, because the neutron source increases exponentially as a function of burnup. The issue is discussed in more detail in Section 3.5.1.

Figure 7. The MCNP model used for evaluation of the dose in the maintenance gap. Because three model boundaries are reflective, the model represents an infinitely long canister storage, where the canisters are placed every 1.6 meters on both sides of the storage room. The dose rates in the maintenance gap between the canister storage wall and the rock tunnel were calculated using the MCNP variance reduction techniques. Secondary gamma radiation doses were not calculated as they were considered minuscule. The calculated dose rates are shown in Table 5. The maximum dose rate in the maintenance gap is reasonable considering the need to monitor the condition of the area.

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Table 5. Dose rates in the maintenance gap between the canister storage wall and the rock tunnel.

Dose component Dose rate (µSv/h)

Primary gamma 1.388 ± 0.092Neutron 3.029 ± 0.110Total 4.417

3.4.3 Dose rate near the canister hoist

There is a need to divide the tunnel where the canister storage is located in two parts separated with an entrance maze that ensures that the dose rate is low in the room where the canister hoist is located. Moreover, a low dose rate level in the maintenance room is required for the automated guided vehicles. The entrance maze was studied on a conceptual level with the aim to provide a viable solution for the structural design. The entrance maze needs to be relatively broad so that malfunctioning of an automated guided vehicle in the maze can be clarified with another automated guided vehicle. The MCNP model used for investigation of the shielding effectiveness of the entrance maze is shown in Figure 8. The left boundary of the model is reflective, which effectively doubles the source term to match the planned canister storage capacity of 30 disposal canisters. The canister storage was assumed to contain BWR type disposal canisters filled with BWR fuel whose burnup was 60 MWd/kgU and that had been stored for 20 years after reactor use. The assumption is very conservative as in reality disposal canisters would contain spent nuclear fuel with various interim storage length because the total decay heat power of each disposal canister is limited. Furthermore, the most radioactive fuel bundles can be placed in the central positions of the BWR type disposal canisters which significantly decreases the dose rates on the surface of the disposal canisters.

Figure 8. The MCNP model used for investigation of the shielding effectiveness of the entrance is maze. The MCNP mesh tally was used to study the gamma and neutron dose rate in the maze geometry. The MCNP variance reduction techniques were used in order to increase the sampling of particle tracks in the maze structure. Figures 9, 10, and 11 show the gamma, neutron and total dose rate in the horizontal plane that corresponds to the axial middle of the spent fuel. According to the MCNP results the total dose rate in the

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canister storage is higher than 100 mSv/h due to the intense gamma radiation source. The results show that the entrance maze is an effective radiation shield against gamma radiation. However, the maze does not prevent leakage of neutron radiation as effectively, and neutrons are actually the dominant dose contributors near the canister hoist. The total dose rate falls below 2.5 µSv/h a few meters from the maze, which can be considered an acceptable dose rate level near the canister hoist. The same amount of time was allocated for both the gamma and neutron transport calculation with MCNP. The results show, that while the obtained neutron dose rate distribution is relatively smooth, more calculation time would be needed in order to obtain a smooth gamma dose rate distribution. However, this would not have an effect on the conclusions of the study as the neutrons cause most of the dose rate near the canister hoist.

Figure 9. Gamma dose rate in the canister storage entrance maze. The axial level corresponds to the axial middle of the spent nuclear fuel.

Figure 10. Neutron dose rate in the canister storage entrance maze. The axial level corresponds to the axial middle of the spent nuclear fuel.

Figure 11. Total dose rate in the canister storage entrance maze. The axial level corresponds to the axial middle of the spent nuclear fuel.

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3.4.4 Dose rate in the upper hall during a canister transfer by lift

The cruciform hall above the canister storage is used for several purposes. Firstly, there is a connection to the canister lift that is used for transportation of bentonite blocks into the repository. Secondly, the level is connected to the repository tunnels and there is a spent fuel canister loading station that is used for hoisting the spent fuel canister into the radiation shield of the canister transfer vehicle. In addition to the canister loading station two other openings between the canister storage and the hall are planned: a hatch is required for transportation of equipment between the levels and a staircase is planned to connect the two levels. During the transport of a disposal canister from the encapsulation plant into the canister storage using the canister lift the disposal canister passes the level above the canister storage and causes elevated dose rate there. A radiation shielding wall made of concrete is being planned in order to ensure that the dose rate in the upper hall stay low. The effectiveness of the radiation shielding wall was studied using a simplified MCNP model shown in Figure 12. The radiation shielding wall, that is planned to consist of two parts, was modelled as a block whose one side is cut. The canister lift was assumed to carry a BWR type disposal canister loaded with 12 spent nuclear fuel bundles whose assembly average burnups were 60 MWd/kgU and that had been stored for 20 years after reactor use. The radiation shielding wall prevents the direct radiation from reaching the centre of the hall. However, the scattered gamma and neutron radiation raise the dose rate level in the upper hall when there is a disposal canister in the canister lift on the same level.

Figure 12. The MCNP model used for investigation of the effectiveness of the radiation shield wall in the upper level. The MCNP mesh tally was used to calculate the gamma and neutron dose rate distributions in the upper hall. The dose rates were calculated for the axial middle of the spent fuel bundle. The MCNP variance reduction techniques were used in order to increase the sampling of particle tracks in the region of interest. Figures 13, 14, and 15 show the gamma, neutron and total dose rates in the horizontal plane that corresponds to the axial middle of the spent fuel bundle in the canister lift. The results show that the dose rate at some 25 meters distance from the canister lift is of the order 25 µSv/h. The dose rate is mainly due to the gamma radiation. This is an acceptable dose rate level provided that employees can not enter area of high dose rate when there is a disposal

17

canister in the canister lift. However, the canister passes the level within less than 20 seconds and thus, even relatively high dose rate dose not cause a significant dose as the situation is temporary.

Figure 13. Gamma dose rate in the upper hall when a disposal canister passes the level. The axial level corresponds to the axial middle of the spent fuel bundles.

Figure 14. Neutron dose rate in the upper hall when a disposal canister passes the level. The axial level corresponds to the axial middle of the spent fuel bundles.

Figure 15. Total dose rate in the upper hall when a disposal canister passes the level. The axial level corresponds to the axial middle of the spent fuel bundles.

18

3.5 Radiation shield of the canister transfer and installation vehicle

The final disposal canister is transported from the canister storage to the final disposal tunnel to be positioned into a deposition hole. The transport vehicle's radiation shield is a tube that lies in a horizontal position on top of a carriage in the vehicle's rear part. The shield is planned to consist of a 5 cm thick inner shielding tube made of borated (5 weight-%) polyethylene and a 15 cm thick outer cast iron shielding tube. The back end of the radiation shield is called the rear radiation shield, while the front end that is closest to the driver's cabin is called the radiation shield end. The design target of the shield design is to ensure that the dose rate on surface of the radiation shield is always less than 2 mSv/h (Kirkkomäki et al. 2005). As the maximum dose rate on the surface of the disposal canister is according to Ranta-Aho (2008) some 200 mSv/h, the radiation shield should provide a shielding factor of the order of 100. The canister transfer and installation vehicle is planned to manned, and in this study a target value of 2.5 µSv/h was used for the dose rate level in the driver's cabin.

3.5.1 Previous studies

The dose rates on surface of the radiation shield have been calculated in the past by Anttila (Salo 1997). However, those analyses were made assuming different material compositions (inner shield made of unborated polyethylene, and outer shield made of steel) and the maximum burnup of only 45 MWd/kgU. According to the results of Anttila (Salo 1997) the maximum dose rate on the surface of the radiation shield was some 0.5 mSv/h. Some 60 % of the dose rate was caused by neutrons and the rest by gamma radiation. The increase in the assumed maximum burnup increases the radiation source term. As the dominant gamma source nuclide (Cs-137) of the spent nuclear fuel at 20 years cooling times is a fission product, it can be assumed that the gamma source term behaves linearly as a function of burnup. The neutron source, however, increases exponentially as a function of burnup (Würz 1991)

Bn BUS (1),

where B is a constant of the order of 4.5 - 4.8 according to Anttila (1996). Using the formula it can be estimated that the increase in burnup from 45 MWd/kgU to 60 MWd/kgU results in some 300 % increase in the neutron source. It should be noted that the burnup limit of 60 MWd/kgU is set for the bundle average burnup, while in reality the bundle has an axial burnup distribution. The effect of the axial burnup distribution on the neutron source is shown in Figure 16. Often the effect of the axial burnup on the neutron source distribution is omitted and a uniform axial neutron source distribution is assumed. This leads to underestimation of the neutron dose rates on the sides of the fuel assembly and overestimation of the neutron dose rates on top and bottom of the fuel bundle. It should also be noted that the overall neutron source calculated using a realistic axial burnup distribution can be almost 30 % larger than the one obtained assuming a uniform burnup distribution.

19

0

0.5

1

1.5

2

2.5

1 2 3 4 5 6 7 8 9 10

Node

Bu

rnu

p d

istr

ibu

tio

n o

r n

eu

tro

n s

ou

rce

no

rma

lize

d w

ith

n

eu

tro

n s

ou

rce

ca

lcu

late

d a

ss

um

ing

un

ifo

rm b

urn

up

BurnupNeutron Source

Figure 16. The effect of the axial burnup distribution on the neutron source. Anttila (Salo 1997) also investigated the effect of the radiation shield order on the dose rates and found the shield to be effective even if the order of the shielding materials was altered. Unfortunately, the documentation of the work does not clarify whether the secondary gamma radiation caused by neutrons was taken into account in the analyses. Anttila also made some calculations assuming boron in polyethylene. According to the calculations of Anttila 1 weight percent of boron decreased the neutron dose on the outer surface by some 20 %.

3.5.2 Axial source distribution

In this study the geometry of the models of the disposal canister and its radiation source were similar to those used by Ranta-Aho (2008). Calculations were made using a model of the BWR canister where the fuel bundles were homogenized. The analyses were carried out assuming that the spent nuclear fuel has a uniform burnup distribution. For shielding analyses the uniform burnup assumption leads to underestimation of the dose rates on the outer surface of the radiation shield in the axial midplane of the shield, while the top and bottom dose rates are exaggerated. Assuming that the maximum axial burnup is some 20 % higher than the average bundle burnup, it can be assessed that the axial midplane gamma source is at largest some 20 % higher than the average. Using Formula 1 it can be calculated that 20 % increase in burnup implies a neutron source that is some 2.4 times higher than the neutron source corresponding to the average bundle burnup. This number was used for scaling of the neutron dose rates calculated using a uniform axial neutron source distribution. Similarly the calculated gamma dose rates were scaled by a factor of 1.2 based on the assumptions that the maximum axial

20

burnup is some 20 % higher than the average bundle burnup, and that the gamma source depends linearly on burnup.

3.5.3 Azimuthal dose rate distribution

In Figure 17 is shown the calculated horizontal dose rate caused by primary gamma radiation in the axial midplane of the fuel. In the central region of the canister (white colour) the dose rate is higher than 1 Sv/h. The figure shows that the dose rate on the outer surface of the radiation shield varies azimuthally. Similarly in Figure 18 is shown the horizontal neutron dose rate that is much more uniform azimuthally. The azimuthal dose rate variations were studied further by calculating the gamma and neutron dose rates as a function of the azimuthal angle. The results of these studies are shown in Figure 19. According to the calculation results the maximum azimuthal primary gamma dose rate on the shield surface is 1.41 times larger than the average. Similarly the maximum azimuthal neutron dose rate on the shield surface is 1.11 times larger than the average. These numbers were used for scaling of the calculated average neutron and gamma dose rates on the surface of the canister.

Figure 17. Horizontal distribution of the dose rate caused by primary gamma radiation in the axial midplane of the fuel. In the central region of the canister (white colour) the dose rate is higher than 1 Sv/h.

21

Figure 18. Horizontal distribution of the neutron dose rate in the axial midplane of the fuel.

0

50

100

150

200

250

300

350

400

450

0 45 90 135 180 225 270 315 360

Azimuthal angle in degrees

Gam

ma

Do

se R

ate

(µS

v/h

)

Primary gammas

Neutrons

Figure 19. Azimuthal distribution of the primary gamma and neutron dose rates on the outer surface of the radiation shield.

22

3.5.4 Maximum dose rate as a function of the distance from the radiation shield

The MCNP geometry model was used to calculate the azimuthally averaged primary gamma, neutron, and secondary gamma dose rates as a function of the distance from the radiation shield of the canister transfer and installation vehicle. In order to estimate the maximum dose rates the calculated azimuthally averaged dose rates were multiplied with the factors described in Sections 3.5.2 and 3.5.3. The multiplication factors used for adjustment of the secondary gamma dose rate were the same as those used for correction of the neutron dose rates. The results are shown in Figure 20 and Table 6. According to the calculation results the maximum total dose rate on the surface of the radiation shield of the canister transfer and installation vehicle is of the order of 1.5 mSv/h that is below the target value of 2 mSv/h. The dominant dose rate component is the neutron radiation that accounts for over two thirds of the total dose rate. The maximum total dose rate at 2 meters from the shield surface was found to be 0.23 mSv/h. It should be noted that the model did not include any other structural parts of the transfer vehicle than the radiation shield itself. Also the disposal tunnel walls and floor were not taken into account. The inclusion of the tunnel walls in the model would result in up to some tens of percents increases in the photon dose rates near the surfaces. The increase in neutron dose rates could be even larger. The vehicle structures would cause similar effects but they would also function as additional radiation shields. The dose rate caused secondary gamma radiation was found insignificant when compared with the dose rates caused by primary radiation.

1

10

100

1000

0 1 2 3 4 5 6 7 8 9 10

Distance from Shield Surface (m)

Do

se R

ate

(µS

v/h

)

Primary Gamma

Neutron

Secondary Gamma

Total

Figure 20. Maximum gamma and neutron dose rates as functions of the distance from the radiation shield surface.

23

Table 6. The maximum gamma and neutron dose rates as functions of the distance from the surface of the radiation shield of the canister transfer and installation vehicle.

Distance from the Shield Surface

(m)

Primary Gamma Dose Rate

(µSv/h)

Neutron Dose Rate

(µSv/h)

Sec. Gamma Dose Rate

(µSv/h)

Total Dose Rate

(µSv/h) 0 341.4 571.1 7.8 920.3 1 118.2 164.0 2.5 284.8 2 66.3 81.8 1.3 149.4 3 41.7 47.6 0.8 90.1 4 28.6 30.7 0.5 59.8 5 20.4 21.3 0.4 42.1 6 15.4 15.5 0.3 31.2 7 12.1 11.8 0.2 24.1 8 9.6 9.3 0.2 19.0 9 7.7 7.5 0.1 15.3 10 6.3 6.2 0.1 12.6

3.5.5 Rear radiation shield

According to the calculations by Ranta-Aho (2008) the gamma and neutron dose rates on the surface of the bottom of the disposal canister are substantially smaller than those of the canister side. Therefore, a thinner rear radiation shield is sufficient for achievement of the targeted 2 mSv/h dose rate. Furthermore, the results of Ranta-Aho (2008) were obtained assuming a uniform axial burnup, and therefore especially the calculated neutron dose rate was exaggerated. However, the dose rates calculated by Ranta-Aho (2008) are the average dose rates for the bottom of the disposal canister, while in reality the dose rate on the surface of the bottom of the disposal canister varies. After some preliminary calculations it was found that a radiation shield consisting of 5 cm of borated polyethylene and 10 cm of cast iron is sufficient to ensure that the dose rate on the surface of the rear radiation shield stays below 2 mSv/h. The final MCNP calculation was carried out assuming a uniform axial burnup distribution. The distributions of the gamma and neutron dose rates on the surface of the rear radiation shield are shown in Figures 21 and 22. The calculated maximum gamma, neutron and secondary gamma dose rates are shown in Table 7.

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Figure 21. The distribution of the gamma dose rate (mSv/h) on the surface of the rear radiation shield.

Figure 22. The distribution of the neutron dose rate (mSv/h) on the surface of the rear radiation shield.

Table 7. Maximum dose rates on the surface of the rear radiation shield assuming a shield consisting of 5 cm of borated polyethylene and 10 cm of cast iron.

Dose Component Dose Rate(mSv/h)

Primary gamma 0.492 Neutron 0.215 Secondary gamma 0.007 Total 0.714

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3.5.6 Radiation shield end

The radiation shield end attaches to the radiation shield tube's forward end. The design of the radiation shield end is very complex as it houses the equipment for lifting of the disposal canister. As the dose rate on the top of the disposal canister is of the same order of magnitude as that of the bottom of the disposal canister it can be stated that the radiation shield of the radiation shield end should provide roughly the same amount of shielding as the rear radiation shield. Thus, a radiation shield consisting of some 5 cm of borated polyethylene and 10 cm of iron should be adequate. This was confirmed with MCNP calculations, whose results are shown in Figures 23 and 24, and Table 8.

Figure 23. The distribution of the gamma dose rate (mSv/h) on the surface of the radiation shield end.

Figure 24. The distribution of the neutron dose rate (mSv/h) on the surface of the radiation shield end.

26

Table 8. Maximum dose rates on the surface of the radiation shield end assuming a shield consisting of 5 cm of borated polyethylene and 10 cm of cast iron.

Dose Component Dose Rate(mSv/h)

Primary gamma 0.378 Neutron 0.162 Secondary gamma 0.005 Total 0.545

3.5.7 Radiation shielding of the driver's cabin

The shielding of the driver's cabin of the canister transfer and installation vehicle was studied assuming that the radiation shield end consists of 5 cm of borated polyethylene and 10 cm of iron. The driver's cabin was assumed to have dimensions of 1 m, 1 m and 1.5 m, and the distance between the radiation shield end and the back of the driver's cabin was assumed to be some 2.5 m. The floor of the driver's cabin was assumed to be on the same level as the centre of the horizontally mounted disposal canister, and the cabin was assumed to be positioned 90 cm off the symmetry axis of the canister. According to the preliminary MCNP calculations the maximum gamma and neutron dose rates at 2.5 m from the radiation shield end are of the order of 150 and 40 µSv/h, respectively. If the dose rate inside the cabin were required not to exceed 2,5 µSv/h, it can be calculated that about 6 cm of lead (or an equivalent amount of other shielding material) is required to achieve the required gamma attenuation factor. Moreover, borated polyethylene or other shielding material against neutrons is needed. In this work the driver's cabin was assumed to have a 9 cm thick outer neutron of shield consisting of borated (5 weight-%) polyethylene. The rest of the transport vehicle was not included in the model. The overall effect of the simplification on the calculation results is difficult to assess as the transport vehicle structures partly provide additional shielding, but they also cause scattering of radiation. The transfer vehicle was assumed to be located in a rock tunnel whose width and height were 3.5, and 5.25 m, respectively. Figures 25 and 26 show the calculated gamma and neutron dose rates in the lowest part of the driver's cabin. The maximum dose rates in the cabin are found there, because the lower part of the cabin is closer to the radiation source. The results show that it is relatively straight-forward to design a radiation shield against gamma radiation, while shielding against scattering neutron radiation is more complicated. According to the calculation results the dose rate in the driver's cabin can in principle be lowered to the targeted dose rate level, but the corresponding radiation shield will be heavy. For example the radiation shield of the MCNP model used in this work consisted of 4800 kg of lead and 2300 kg of borated polyethylene. The weight could possibly be reduced significantly by optimization of the shape of the radiation shield. It should also be noted that the radiation shield is opaque and mirror or camera and display systems would be needed for driving of the vehicle. Thus, a remote controlled canister transfer and installation vehicle could be a better solution from the radiation shielding point of view. Alternatively, the radiation shielding efficiency of the radiation shield end could be increased.

27

Figure 25. Calculated gamma dose rate distribution on the level of the lowest part of the driver's cabin. White colour indicates that the dose rate is either above 100 Sv/h or below 0.1 µSv/h.

28

Figure 26. Calculated neutron dose rate distribution on the level of the lowest part of the driver's cabin. White colour indicates that the dose rate is either above 100 Sv/h.

29

In Figure 27 is shown the maximum total dose rate in the rock tunnel near the canister transfer and installation vehicle on the level of the lowest part of the driver's cabin. At longer distances the dose rate is caused mainly by neutron radiation. The total dose rate in the tunnel falls below 25 µSv/h when the distance from the canister is more than five meters. In reality the dose rates would be smaller as due to the total decay heat limit the canister can not be loaded with 12 spent fuel bundles with the maximum radiation source as assumed in the MCNP modelling.

Figure 27. Maximum total dose rate in the rock tunnel near the canister transfer and installation vehicle.

30

3.6 Disposal canister in the disposal hole

After the disposal canister filled with spent nuclear fuel has been placed in the disposal hole, the canister transfer vehicle is driven away, and there is no radiation shield on top of the canister until the bentonite buffer blocks are placed on top of the canister to close the disposal hole. The case was analyzed using a VVER canister model with a homogenized fuel region and assuming that the top of the canister is 2 meters below the tunnel floor. The VVER canister was assumed because the dose rate on top of the VVER canister is somewhat higher than that of the BWR canister (the dose rate on top of the EPR type canister is of the same order than that of the VVER canister). The height of the tunnel was assumed to be 4 meters. The bentonite blocks surrounding the disposal canister were assumed to consist of SiO4 and to have a density of 2.1 kg/m3.

3.6.1 Dose rate in the disposal tunnel

Figure 28 shows the total dose rate as a function of the radial and horizontal distance from the centre of the disposal canister. At longer distances the maximum dose rates are found 2.4 meters above the tunnel floor (Z=600 in Figure 28). In Figure 29 are shown the dose rates on this level as a function of the distance from the centre of the disposal hole. The results show that the total dose rate falls below 25 µSv/h at distances longer than 4 meters. At short distances primary gamma radiation is the dominant dose contributor. The dose rate caused by neutron radiation equals the gamma radiation dose at a distance of 4.5 meters, but there the dose rates are already very low. Considering that the placement of the bentonite block on top of the final disposal canister can be completed relatively quickly, the placement can in principle be carried out using a manned vehicle designed so that the distance between disposal hole and the driver seat is at least 4 meters.

Figure 28. Total dose rate in the disposal tunnel when a canister is placed in the disposal hole, but bentonite blocks have not yet been placed on top the canister.

31

Figure 29. Dose rates in the disposal tunnel 2.4 meters above the tunnel floor as functions of the distance from the centre of the disposal hole.

32

4 CONCLUSIONS

This report summarizes the results of the supplementary radiation shielding calculations for the final disposal facility. The gamma and neutron shielding calculations were made using the MCNP Monte Carlo code. The radiation source terms were based on ORIGEN-S calculations, where it was assumed that the burnup of the spent fuel was 60 MWd/kgU and that it had been stored for 20 years after reactor use. Additionally, the radiation shielding of the weld inspection room was assessed. First analysis was related to the dry VVER transfer cask option. The gamma and neutron dose rates were calculated in the decontamination centre above the handling cell in case a dry transfer cask containing 84 VVER fuel bundles is connected to the handling cell. The total dose rate in the decontamination centre was found to be smaller than 0.2 µSv/h, and thus, there is no need to increase the thickness of the handling cell ceiling in order to enable the dry VVER spent fuel transfer option. The radiation shielding adequacy of the front and back end shielding structures of the canister transfer corridor were studied using a simplified model of the VVER type disposal canister. The analyses confirmed that radiation shielding designs of the ends of the canister transfer corridor are sound. The radiation shielding of the X-ray linear accelerator was studied with two simple MCNP models. Firstly, the required thickness of the primary radiation shield was studied using a MCNP model, where the radiation beam entered a concrete block. Secondly, the radiation beam was set to hit a small copper target in order to study the scattering of the gamma radiation. According to the calculation results the primary shield made on concrete need to be at least 2.9 meters thick, and it should have a sufficient radius in order to account for the forward-peaking scattering radiation. The walls of the weld inspection room with a thickness of 80 cm were found sufficient to attenuate the rest of the scattering radiation. It was noted that the photons generated by the linear accelerator have a sufficient energy for production of photoneutrons, but the production rate was not evaluated in this work. Several MCNP models were made in order to study the radiation shielding of the canister storage of the final disposal canisters in the repository. The dose rate above the canister storage and in the maintenance gap between the canister storage and the rock tunnel was studied with a model corresponding to an infinitely long canister storage room. The planned ceiling and wall of the canister storage were found to provide sufficient radiation shielding. Interestingly neutron radiation was found to be the dominant dose contributor in this geometry. It should be noted that the analysis is conservative because the canister-wise limit for the overall decay heat results in considerably smaller radiation source than what was assumed and in the case of BWR and VVER fuel the most radioactive fuel bundles can, and should, be placed in the centre of the disposal canister. The canister hoist is located on the same level as the canister storage, and an entrance maze is planned to separate them. A generic MCNP model was used to study the

33

shielding effectiveness of the maze. According to the calculation results the entrance maze lowers the total dose rate near the canister hoist below 2.5 µSv/h. Also in this case neutron radiation was found to be the dominant dose rate contributor near the canister hoist. During the transport of a disposal canister from the encapsulation plant into the canister storage using the canister lift the disposal canister passes the level above the canister storage. The effectiveness of an asymmetric radiation shielding wall was studied using a simplified MCNP model. According to the calculation results the total dose rate at some 25 meters from the canister lift is of the order of 25 µSv/h, which is an acceptable dose rate level for a temporary situation. There is however, a need to ensure that the area near the canister lift can not be entered during the transfer of the canister in the lift. The feasibility of the transport vehicle's radiation shield was investigated rigorously. The objective of the shield design is to ensure that the dose rate on the surface of the radiation shield is lower than 2 mSv/h. Moreover, the dose rate in the driver's cabin of the vehicle should be low. The nonlinear behaviour of the neutron source as a function of burnup and the azimuthal variability of the dose rate on surface of the disposal canister were realized to affect the surface dose rates significantly, and the required additional margins were determined in order to account for them. The calculated axial and azimuthal maximum total dose rate on the surface of the radiation shield was 1.5 mSv/h. Thus, the radial design of the radiation shield meets the requirements. Neutrons were found to account for over two thirds of the total dose rate. The secondary gamma radiation caused by primary neutrons was found insignificant for the overall dose rates. The top and bottom of the disposal canister require somewhat less radiation shielding. Both rear radiation shield (bottom) and radiation shield end (top) were found to require some 5 cm of borated (5 weight-%) polyethylene and 10 cm of stainless steel in order to reduce the top and bottom surface dose rates below 2 mSv/h. In case the canister transfer and installation vehicle is not controlled remotely, the driver's cabin was evaluated to require a massive radiation shield in order to reduce the dose rate in the cabin below 2.5 µSv/h. The preliminary radiation shield concept that provides sufficient shielding consists of both borated polyethylene and lead would weight 7000 kg. Additionally, there is a need to equip the driver's cabin with mirror or camera and display systems. The reference dose rate level chosen for the calculations may be excessively low, and even a dose rate higher by an order magnitude maybe adequate. The dimensioning of the shielding for the driver's cabin shall be optimized considering the real expected exposure time and the effectiveness of the shielding. This kind of dimensioning is beyond the scope of the current report and left for the future design phases. In addition, the alternative of remotely controlling the canister transfer and installation vehicle should be further investigated. The placement of the disposal canister in the disposal hole includes a phase when the disposal canister is in the disposal hole, but there is no bentonite block on top of it. The case was analyzed for the VVER type disposal canister. According to the calculations results the total dose rate falls below 25 µSv/h at a distance of some 4 meters from the

34

centre of the disposal hole. The radiation situation is not especially challenging considering the task of designing a vehicle for the placement of the bentonite blocks. The calculations carried out in this study confirmed that radiation shielding has been adequately accounted for in the basic design of the investigated structures and systems of the final disposal facility. One noteworthy finding of the study is that in some specific cases neutron radiation can be the dominant dose contributor, which should be taken into account in monitoring of occupational exposure at the final disposal facility. It's also important to note that the neutron source has a strong axial variation and the neutron source term calculated assuming a uniform axial burnup profile can be as much as 30 % underestimated.

35

REFERENCES Anttila, M. 1996, Gamma and neutron dose rates on the outer surface of the nuclear waste disposal canisters, Posiva-96-10. Anttila, M. 1998, Radiation protection calculations for an encapsulation plant, Working Report 98-81. Anttila, M. 2005, Radioactive Characteristics of the Spent Fuel of the Finnish Nuclear Power Plants, Posiva Working Report 2005-71. Bell, Z. W. 1995, Evaluation of photoneutron production at high energy linacs, Oak Ridge Y-12 Plant Report Y/DW-1367. IAEA, 2000, Handbook on photonuclear data for applications, IAEA-TECDOC-1178. ICRP, 1996, Annals of the ICRP, Conversion Coefficients for use in Radiological Protection against External Radiation, ICRP Publication 74, Volume 26 No. 3/4, 1996. Kirkkomäki, T & Raiko, H. 2005. Canister Transfer in Access Tunnel: Lay-out, system and operation description. Posiva Working Report 2005-54. (in Finnish) Müller, Ch. et al. 2006, Reliability of non-destructive testing (NDT) of the copper canister seal weld. SKB report R-06-08. Ranta-Aho, A. 2008, Review of the Radiation Protection Calculations for the Encapsulation Plant, Posiva Working Report 2008-63. Ronnetag, U. et al, 2006. Reliability in sealing of canister for spent nuclear fuel. SKB report R-06-26. Salo, J-P. 1997, Arvio säteilytasoista käytetyn polttoaineen loppusijoituskapselin kuljetussuojan ulkopuolella, Muistio KH-M-65-97, Posiva Oy. (in Finnish) Sandelin, S. 2010, X-ray inspection setups for the disposal canister lid weld. Posiva Working Report 2009-98. Varian Medical Systems, 2007. Varian Linatron High-Energy X-ray Applications. http://www.os.varian.com/media/security_and_inspection/resources/technical_informati on/pdf/LinatronAppManual7.pdf Würz, H. A simple nondestructive measurement system for spent-fuel management. Nuclear Technology, Vol.95, pp. 193-206.

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APPENDIX I

ASSESSMENT OF THE NEUTRON INDUCED ACTIVATION AT THE ENCAPSULATION PLANT AND THE REPOSITORY OF SPENT NUCLEAR FUEL Tapani Eurajoki / 7.12.2012 To assess the effect of neutron-induced activation in the components, systems and structures of the encapsulation plant and the final disposal facility, reaction rates at the surface of a BWR canister have been calculated for those activation reactions, which typically contribute most to the induced activity at nuclear power plants. According to the guide YVL C.1 special attention is to be paid on nickel, cobalt, silver and antimony contents in the materials, which is reflected in the formation of the activation products Co-58, Co-60, Ag-110m and Sb-124. In addition to these Eu-152 is a typical activation product in concrete, because of its large activation cross section. The highest neutron flux densities which structures and components may be exposed to are found outside an unshielded EPR element (averaged over the length of the element 3.3E4 n/cm²/s) or outside a BWR canister (3.9E4 n/cm²/s). Even though the flux densities are low e.g. compared to those at a nuclear power plant, they may result in activation if the irradiation time is long and the activation cross section large. Thus, the components and structures of materials may have to set maximum concentrations of some elements. The activation has been assessed with an MCNP5 calculation. The calculation assumes the given materials to be present on the BWR canister surface. The canister model is similar to the one reported by Ranta-aho (2008), and to calculate the reaction rate the cross section data provided with MCNP 5 are applied. The following equation gives the specific induced activity (in terms of Bq/g) of each nuclide:

cutoffE

Att dEEEM

Nee

m

A)()(1 21

,

where A total activity of a sample m mass of the component (or unit mass) disintegration constant t1 irradiation time t2 disintegration time

mass fraction of the target element abundancy of the target isotope in natural element NA Avogadro number M Atom weight of the target isotope Ecutoff cutoff energy

37

activation cross section neutron flux density. The energy integral of the product of the neutron flux density and reaction cross section is the primary quantity calculated in the MCNP run. The results for pure elements are presented in Table I-1. The table shows the saturation activities of pure (i.e. consisting 100 %) elements. Since the activation is relatively low, the design targets can also be set low without excessive requirements for the materials. If design targets are set to 10 Bq/g for the saturation activity and 1 Bq/g after a decay time of 1 year, the maximum concentrations for the studied elements are those presented in Table I-1. Hence the following conclusion can be drawn regarding the material selection for components, systems and structures exposed to neutron radiation at the encapsulation plant and the final disposal facility. The neutron-induced activity remains low as long as the Co content of the materials staying in close contact to the spent fuel elements and canisters does not exceed 0.1 %, and the Eu content of concrete structures does not exceed 10 ppm. For cobalt and europium, these limits are set somewhat lower than given in the table, because finding materials achieving these limits should be no problem. For example at nuclear power plants an even lower limit for Co content in steels is typically applied. In addition to cobalt and europium, silver and antimony should be avoided as significant concentrations in structural materials, but as trace concentrations they have no significance. Table I-1. Neutron induced saturation activities and limit values for the target elements resulting in saturation activity below 10 Bq/g and activity below 1 Bq/g after a cooling time of 1 year. Target element

Activation product

Saturation activity of pure target element, Bq/g

Limit for the target element if activity below 1 Bq/g after 1 year is required

Limit for the target element if saturation activity below 10 Bq/g is required

Ni Co-58 1.34 1 1 Co Co-60 236 0.0048 0.042 Eu Eu-152 783 0.0014 0.013 Ag Ag-110m 19.7 0.14 0.5 Sb Sb-124 73.9 0.9 0.1