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1. Introduction
The Ministry of Trade, Industry and Energy of the Republic of Korea has announced in the Second Basic Plan for High-Level Radioactive Waste Management [1]. They planned to manage spent nuclear fuel through the stages of temporary and interim storages before final disposal. The specific operating systems for facilities at each management stage will be determined in the future, considering public acceptance and engineering safety. The metal overpack system is one of the operating systems for the dry storage facility. This system entails encapsulating spent nuclear fuel in canisters and placing these canisters within metal cylindrical overpacks designed for transport and storage. The overpacks are subsequently stored vertically in a designated storage area [2-3].
Meanwhile, a number of studies have been conducted in several countries on the radiation dose assessment of workers handling spent nuclear fuel loaded in cask. Holtec International, U.S., has developed operating procedures for loading and unloading spent nuclear fuel using HI-STAR 100 metal overpacks [4]. Furthermore, they have calculated the collective dose of workers using the Monte Carlo NParticle Transport (MCNP) code. Ko et al. conducted a radiation shielding evaluation of a metal overpack using the MCNP code [5]. An assessment of dose has been conducted for handling workers at spent nuclear fuel storage facilities, using NAC-UMS, VSC-24 and NUHOMS developed by companies like NAC, EnergySolutions, ORANO, based on their respective operating procedures [6-8]. Dominion conducted assessment of collective dose for workers handling TN-32 metal casks at the North Anna Nuclear Power Plant using the MCNP code to obtain licensing for a dry storage facility using TN-32 casks [9]. Pacific Northwest National Laboratory calculated the collective dose for workers handling CASTOR V/21 transport and storage casks through measurements [10]. P.F. Weck also assessed the dose of workers at different spent nuclear fuel management facilities and determined the collective dose by conducting measurement [11]. Kim et al. conducted a preliminary shielding evaluation of a concrete overpack using the MCNP code [12]. Kim et al. conducted a preliminary assessment of radiation impact from dry storage facilities for PWR spent fuel using the MCNP code [13].
Currently, spent nuclear fuel in Korea is planned to be discharged from nuclear power plants and passed through the temporary and interim storage stages on site, followed by the disposal stage. In this process, workers at the storage facility must perform the transport of spent nuclear fuel from the current stage to the management facility of the subsequent stage in the entire fuel cycle [14]. Radiation workers are subject to consideration of the as low as reasonably achievable principle in accordance with Article 53(3) and Article 91 of the Nuclear Safety Act, and an assessment of the dose should be conducted [15-16]. Therefore, the Korea Radioactive Waste Agency (KORAD) has conducted the dose assessment for workers involved in loading spent nuclear fuel from wet storage at nuclear power plants into transport casks and placing storage casks in dry storage facilities. However, as mentioned earlier, there is insufficient research on worker’s dose assessment conducted from the perspective of transporting spent nuclear fuel to a management facility of the subsequent stage within the entire life cycle management of spent nuclear fuel. This study assessed the collective dose received by workers transporting spent nuclear fuel by using metal overpacks designed for both transport and storage.
The objective of this study is to assess the dose to workers transporting spent nuclear fuel in metal overpacks. To this end, a cask was selected for assessment and the source term was set up. Additionally, an exposure scenario was set for workers transporting spent nuclear fuel, considering the use of metal overpacks. Based on the exposure scenario, locations of workers were determined, and the dose rates were derived for each location of worker using the MCNP code. The collective dose for each operation was calculated by multiplying the finally derived dose rates by the corresponding work duration and the number of workers involved.
2. Materials and Methods
This study assessed the dose to workers transporting spent nuclear fuel loaded in metal overpack. For this purpose, a cask was selected for assessment and the source term of the design basis fuel was selected for that cask. Then, the radiation exposure scenario for the transport workers for the selected cask was set up. To set up the exposure scenario, we set the scope and assumptions for the transport operation. According to the Final Safety Analysis Report (FSAR) of casks for spent nuclear fuel designed in foreign countries, we set up the work duration, the number of workers, distances of operation, and locations of workers for each operation. Finally, the dose rates at each location was calculated using a Monte Carlo computational code. Based on the radiation exposure scenario, a collective dose for the transport workers was derived.
2.1 Cask for Assessment and Radiation Source Term
To conduct the dose assessment for workers transporting spent nuclear fuel, the first step was to select a cask. Additionally, the flux per energy group for gamma rays and neutrons emitted from the spent nuclear fuel within the cask must be determined. To assess the dose for the workers, this study conducted the following tasks: 1) selection of the cask for the dose assessment of the workers and 2) setting the radiation source term.
2.1.1 Cask for assessment
Table 1 presents the specifications of the cask selected for the assessment in this study [17]. A metal overpack designed by KORAD for transport and storage was selected as the cask to be assessed. The cask has a capacity of 32 bundles of light water reactor spent nuclear fuel and is designed to offsite transport when installing impact limiters. The cask body has 210-mm thick wall made of steel and lead. Additionally, an 110-mm thick neutron shielding is affixed to the cask body. The lid of the metal overpack is constructed of steel and has a thickness of 150 mm. It is situated at the top of the overpack. The impact limiter, which can be attached to the top and bottom of the cask, has an outer diameter of 3,600 mm and a height of 1,090 mm. It is primarily composed of wood encased in steel for shock absorption. Fig. 1 shows the metal overpack selected for this study.
Table 1
Chemical composition of cask components
| Item | Material | Density (g·cm−3) | Nuclide | Weight fraction (wt%) |
|---|---|---|---|---|
|
|
||||
| Cask shell | SA-350 LF.3 | 7.82 | C | 0.002 |
| Si | 0.0028 | |||
| Cr | 0.003 | |||
| Mn | 0.009 | |||
| Fe | 0.9418 | |||
| Ni | 0.035 | |||
| Mo | 0.012 | |||
| P | 0.00035 | |||
| S | 0.0004 | |||
|
|
||||
| Lead | 11.35 | Pb | 1.00 | |
|
|
||||
| Cask lid, Resin cover | SA-516 Gr.70 | 7.75 | C | 0.0008 |
| Si | 0.006 | |||
| P | 0.0003 | |||
| S | 0.0003 | |||
| Cr | 0.1275 | |||
| Mn | 0.0075 | |||
| Fe | 0.8054 | |||
| Ni | 0.045 | |||
| Mo | 0.0075 | |||
|
|
||||
| Neutron shield | Resin | 1.682 | H | 0.06 |
| C | 0.277 | |||
| N | 0.02 | |||
| O | 0.428 | |||
| Al | 0.215 | |||
|
|
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| Impact limiter | Red wood | 0.486 | H | 0.070 |
| C | 0.374 | |||
| O | 0.556 | |||
2.1.2 Radiation source term
Workers handling metal overpacks containing spent nuclear fuel are susceptible to external exposure from gamma and neutron radiation. Gamma rays are emitted due to primary gamma rays from the radioactive decay of fission products and actinides as well as due to the activation of the fuel structures assembly. The activation of the fuel assembly structure is caused by neutrons emitted from spent nuclear fuel, which convert 59Co in the steel and Inconel components of the assembly into 60Co. The structure of the activated assembly release gamma rays with energies of 1.173 and 1.332 MeV. Neutrons emitted from spent nuclear fuel are primarily generated by (α, n) reactions and spontaneous fission within the nuclear fuel materials [18-19].
Tables 2–3 present the flux per energy group for gamma rays and neutrons set for this study. The flux per energy group for gamma rays and neutrons were set as the values used in the preliminary safety assessment of the selected cask. During the preliminary safety assessment, the flux per energy group were derived using ORIGEN-ARP computational code. The metal overpack selected for this study was designed to load 32 bundles of spent nuclear fuel assemblies, enabling the storage of various spent nuclear fuel types. For a conservative dose assessment of workers, the design basis fuel of the cask to be assessed was set as a source term [5]. KORAD has established the design basis fuel to be contained in the cask, specifying an initial enrichment of 4.0wt%, a burnup of 43,000 MWD/MTU, and a cooling period of 10 years. In this study, the energy group of gamma and neutron rays were divided into 18 and 27 groups, respectively, based on the characteristics of the fuel. The flux was set for each of these energy groups.
Table 2
Gamma ray flux per fuel assembly of design basis fuel for metal overpack cask
| Energy groups (MeV) | Gamma flux (photons/sec∙FA) |
|---|---|
|
|
|
| 0.01–0.05 | 9.018×1014 |
| 0.05–0.10 | 2.528×1014 |
| 0.10–0.20 | 1.831×1014 |
| 0.20–0.30 | 5.430×1013 |
| 0.30–0.40 | 3.547×1013 |
| 0.40–0.60 | 1.484×1014 |
| 0.60–0.80 | 1.655×1015 |
| 0.80–1.00 | 7.764×1013 |
| 1.00–1.33 | 4.191×1013 |
| 1.33–1.66 | 5.895×1012 |
| 1.66–2.00 | 1.217×1011 |
| 2.00–2.50 | 4.793×1010 |
| 2.50–3.00 | 2.968×109 |
| 3.00–4.00 | 2.834×108 |
| 4.00–5.00 | 7.478×106 |
| 5.00–6.50 | 3.000×106 |
| 6.50–8.00 | 5.886×105 |
| 8.00–10.0 | 1.250×105 |
|
|
|
| Total | 3.356×1015 |
Table 3
Neutron ray flux per fuel assembly of design basis fuel for metal overpack cask
| Energy group (MeV) | Neutron flux (neutrons/sec∙FA) |
|---|---|
|
|
|
| 1.00×10−11 – 1.00×10−8 | 4.280×103 |
| 1.00×10−8 – 3.00×10−8 | 3.761×102 |
| 3.00×10−8 – 5.00×10−8 | 8.956 ×102 |
| 5.00×10−8 – 1.00×10−7 | 2.510×10−1 |
| 1.00×10−7 – 2.25×10−7 | 4.953 ×10−1 |
| 2.25×10−7 – 3.25×10−7 | 9.185 ×10−1 |
| 3.25×10−7 – 4.00×10−7 | 7.932 ×10−1 |
| 4.00×10−7 – 8.00×10−7 | 1.492 |
| 8.00×10−7 – 1.00×10−6 | 1.530 |
| 1.00×10−6 – 1.13×10−6 | 2.373×101 |
| 1.13×10−6 – 1.30×10−6 | 1.501×102 |
| 1.30×10−6 – 1.77×10−6 | 2.136×103 |
| 1.77×10−6 – 3.05×10−6 | 2.715×104 |
| 3.05×10−6 – 1.00×10−5 | 3.667×105 |
| 1.00×10−5 – 3.00×10−5 | 3.089×105 |
| 3.00×10−5 – 1.00×10−4 | 4.811×106 |
| 1.00×10−4 – 5.50×10−4 | 3.473×107 |
| 5.50×10−4 – 3.00×10−3 | 7.575×107 |
| 3.00×10−3 – 1.70×10−2 | 7.573×107 |
| 1.70×10−2 – 1.00×10−1 | 6.067×107 |
| 1.00×10−1 – 4.00×10−1 | 5.723×107 |
| 4.00×10−1 – 9.00×10−1 | 1.240×107 |
| 9.00×10−1 – 1.40 | 4.425×107 |
| 1.40 – 1.85 | 8.019×107 |
| 1.85 – 3.00 | 2.275×107 |
| 3.00 – 6.43 | 7.230×106 |
| 6.43 – 2.00×10 | 2.491×106 |
|
|
|
| Total | 4.789×108 |
Table 4 presents emission ratio of gamma and neutron rays depending on the height of active fuel region. Spent nuclear fuels have different burnups depending on the height of assembly, giving off different levels of gamma and neutron rays in the active fuel region by height of fuel. In this study, in order to calculate the dose to workers in practical conditions, emission ratios of gamma and neutron rays along the axial direction were considered. To set the emission ratios of gamma and neutron ray along the axial direction, the axial burnup profile of the burnup range of the design basis fuel was chosen. First, the axial burnup profile was chosen referring to the ‘Axial Burnup Profile Database Pressurized Water Reactors’ published by Yankee Atomic Electric [20]. The emission ratios of gamma ray along the axis of the active fuel region change linearly depending on burnup. The emission ratio of neutron ray along the axis of the active fuel was set to be proportional to the fourth power of the axial emission rate of gamma rays based on the methodology presented in NUREG/CR-6802 [18].
Table 4
Emission ratios of gamma and neutron rays depending on the height of active fuel region
| Height of fuel (%) | Emission ratio | |
|---|---|---|
|
|
||
| Gamma ray | Neutron ray | |
|
|
||
| 2.78 | 0.76376 | 0.34027 |
| 8.33 | 0.98938 | 0.95819 |
| 13.89 | 1.09971 | 1.46256 |
| 19.44 | 1.15675 | 1.79043 |
| 25.00 | 1.18349 | 1.96182 |
| 30.56 | 1.17940 | 1.93484 |
| 36.11 | 1.18836 | 1.99431 |
| 41.69 | 1.18257 | 1.95572 |
| 47.22 | 1.15726 | 1.79359 |
| 57.80 | 1.16437 | 1.83808 |
| 58.33 | 1.14425 | 1.71429 |
| 63.89 | 1.13813 | 1.67791 |
| 69.44 | 1.13094 | 1.63591 |
| 75.00 | 1.10136 | 1.47135 |
| 80.56 | 1.07613 | 1.34109 |
| 86.11 | 1.03945 | 1.16739 |
| 91.67 | 0.90845 | 0.68109 |
| 97.22 | 0.65577 | 0.18493 |
Table 5 presents the gamma flux from the decay of 60Co generated from the activation of fuel assembly structures used in this study. Gamma flux from activation of fuel assembly was set by multiplying the radioactivity generated by 1 g of 60Co, the contents of 59Co in the fuel assembly structures, and flux scaling factors of each region. The radioactivity generated by 1 g of 60Co was 43.25 Ci·g−1, which was derived from the ORIGEN-ARP code from the preliminary safety assessment. The contents of 59Co in fuel assembly structures were taken from the ORNL-6051. Initial 59Co contents in zircaloy, stainless steel, Inconel 718 and Inconel X750 in structure materials was set to 10 ppm, 800 ppm, 4,700 ppm, and 6,500 ppm respectively [21]. Flux scaling factors were taken from PNL-6906 [22].
Table 5
Gamma flux from the decay of 60Co generated from the activation of fuel assembly structures
| Structure contents | Specific radioactivity of 60Co (Ci·g−1) | Flux scaling factor | Gamma flux (photons/sec∙FA) |
|---|---|---|---|
|
|
|||
| Top end fitting | 43.25 | 0.1 | 1.177×1012 |
| Plenum region | 0.1 | 3.212×1011 | |
| Active fuel region | 0.2 | 8.895×1011 | |
| Bottom end cap | 0.2 | 6.364×1011 | |
| Bottom end fitting | 1.0 | 3.722×1012 | |
|
|
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| Total | - | 6.746×1012 | |
2.2 Radiation Exposure Scenario for Workers Transporting Spent Nuclear Fuel
When handling spent nuclear fuel loaded in cask, workers may be exposed to external radiation. The magnitude of this exposure can vary depending on the various operational factors such as work duration, the number of workers, distance, and specific locations. Operational factors are different by operations. To set up the radiation exposure scenario for workers transporting spent nuclear fuel, this study defined the scope of the spent nuclear fuel transport operation. Then, the operations that can be carried out by the transport workers were determined, and the operational factors for these operations were set.
2.2.1 Scope of spent nuclear fuel transportation operations
Fig. 2 presents the scope of the spent nuclear fuel transport operation set in this study. In this study, transporting spent nuclear fuel was defined as a series of operations involved in transporting spent nuclear fuel from dry storage to a subsequent management facility. Therefore, the scope of spent nuclear fuel transportation operations was defined to encompass operations ranging from spent nuclear fuel retrieving to the preparation of the package for offsite transport. At this time, the worker’s operations of transporting the spent nuclear fuel to the subsequent-stage management facility located outside the facility after transport preparation was excluded from the assessment scope due to the uncertainty of the route and method of transporting spent nuclear fuel. In this study, workers transporting spent nuclear fuel were categorized according to purpose as follows: 1) retrieval spent nuclear fuel cask workers and 2) workers preparing for offsite transport. Retrieval spent nuclear fuel cask workers perform the operation of transferring cask loading spent nuclear fuel at dry storage area to a preparation area for offsite transport. Operations preparing for offsite transport refer to a series of preparatory activities conducted to transport spent nuclear fuel out of the site.
2.2.2 Assumptions for spent nuclear fuel transport
In this study, assumptions were made to set an exposure scenario for spent nuclear fuel transport workers. The components of a radiation exposure scenario, such as operational procedures, work duration, the number of workers, and distance of operation, can manifest in various ways depending on the structure of the storage facility, transfer and transport of cask means, and the auxiliary equipment used in handling casks. Moreover, the FSAR of the foreign commercial transport and storage system did not provide specific details regarding the required distance of operations between the worker and cask. Therefore, in this study, assumptions were made for 1) storage facilities, 2) transfer and transport means, 3) auxiliary equipment for cask handling, and 4) distance of operations.
First, the structure of the storage facility was assumed to be divided into two areas: 1) storage area and 2) receiving and handling area. The storage area is the area where the spent nuclear fuel is dry stored, and the receiving and handling area is the area where operations are conducted to prepare the spent nuclear fuel for transport to or from the site. The transfer means is the means transferring spent nuclear fuel between the storage area and receiving and handling area, and it is assumed that a vertical cask transporter is used. A vertical cask transporter is equipment used to hoist and transfer casks or overpacks in a vertical direction, and it is equipment with relatively simple operational procedures. It was also assumed that horizontal trailers are used to prepare for the final transport of spent nuclear fuel from the receiving and handling area. Auxiliary equipment refers to facilities utilized for purposes such as facilitating operations or protecting workers during the handling of spent nuclear fuel. For the auxiliary equipment, it was assumed that the equipment used for the commercially available HI-STAR 100 metal overpack would be utilized. Therefore, it was assumed that a metal overpack bottom shield ring and a pocket trunnion plug would be used as auxiliary equipment for the metal overpack. It was also assumed that impact limiters and personnel barriers are used during transport to safely transport the overpack and spent nuclear fuel. The metal overpack bottom shield ring is utilized to reduce the dose rate around the overpack. The pocket trunnion plug is a neutron shielding plug in the trunnion void at the top of the cask [4]. An impact limiter is a device installed at the top and bottom of the cask to mitigate impact in the event of a cask drop during transport. A personnel barrier is the equipment that acts as a physical barrier on the side of a metal overpack during transport to prevent people from accessing the vicinity of the cask [23]. The distances of operation between the worker and cask was assumed to be 1 m and surface of cask (10 cm). The commercial transport and storage systems HI-STAR 100 and HI-STORM 100, referenced in this study for setting the radiation exposure scenario, did not specify the exact distance of operations. Therefore, this study assumed a distance of 10 cm for operations that require worker contact with the cask and a distance of 1 m for operations that do not require direct contact with the cask, such as visual inspection and transferring casks.
2.2.3 Radiation exposure scenario and location of workers transporting spent nuclear fuel
In this study, the exposure scenario for workers transporting spent nuclear fuel was set by referring to the FSAR of commercially available casks in foreign countries in a similar manner to the previously selected cask for assessment. In the U.S., licensing of spent nuclear fuel transport and storage casks requires the submission of an FSAR to the U.S. Nuclear Regulatory Commission, which includes procedures for loading and unloading operations [24-25]. The FSAR of HI-STAR 100, commercially available metal overpack system from Holtec International, was used in this study for set the radiation exposure scenario for worker transporting spent nuclear fuel. Furthermore, the FSAR for the HI-STORM 100 concrete overpack from Holtec International was referenced also. Although HI-STORM 100 is a different type of cask from the metal overpack assessed in this study, metal overpack similar to HI-STAR 100 are used for receiving or shipping spent nuclear fuel [26]. Therefore, this study referred to the operating procedures described in the FSARs of HI-STAR 100 and HI-STORM 100, and set the operational procedures and operational factors considering the scope and assumptions of the transport operations set earlier.
Table 5 presents the exposure scenario set for the transport workers in this study. The transport of spent nuclear fuel in metal cask was divided into a total of 11 tasks. Among these, retrieval and preparation for offsite transport operations were further subdivided into five and six tasks, respectively. We set the total work duration of the retrieval operation to 84 min. The operations of removing bottom shield ring and the pocket trunnion were set to 12 min respectively, while the transfer to the receiving and handling area and the connecting and disconnecting of the vertical cask transporter with the cask were set to 40, 10, and 10 min, respectively. For the transfer of metal overpacks to the receiving and handling area, a distance of operation was set 100 cm from the cask. For all other operations, a distance of operation was set 10 cm from the cask. The total number of workers needed for the retrieval operation was set to 6, with 2 for removing the bottom of the overpack and 1 for the other operations.
We set a total of 79 min for the work duration of the preparation for offsite transport operation. Loading time for the cask onto the transport vehicle was set to 20 min, while installing the tie-down and impact limiters were set to 6 and 16 min, respectively. The visual inspection of the cask and the measurement of the dose rates and contamination level were set as 10 and 17 min, respectively. The distance of operation was set to 100 cm from the cask for loading the transport cask onto the transport vehicle and the visual inspection of the cask. For all other operations, a distance of operation was set to 10 cm from the cask. The total number of workers required for preparation for offsite transport was set to 12, with 3 and 1 for dose rate and contamination level measurement and visual inspection, respectively. For all other operations, we set the number of workers to 2.
Fig. 3 presents the locations of workers transporting spent nuclear fuel using metal overpack in this study. The operational procedures described in the FSAR of HISTORM 100 were mainly used to set the locations of workers. As a result of setting the locations of workers, the number of location for the worker trasporting spent nuclear fuel using a metal overpack was 5. The locations of the workers were broadly categorized as: 1) without the impact limiter, and 2) with an impact limiter installed. Within each situation, 3 and 2 locations were defined, respectively. Without the impact limiters, locations A and C were set at 10 cm from the cask. Location B was set at 100 cm from the cask. With the impact limiter installed, locations E and F were set at 10 cm from the cask surface.
2.3 Occupational Radiation Dose to Workers Transporting Spent Nuclear Fuel
In this study, the metal overpack was simulated using the MCNP code as shown in Fig. 1, and the dose rates were calculated for the locations of workers. F4 tally was used to calculate the dose rates for each location, and the flux was calculated for each location [27]. Then, the location-specific dose rates were derived by considering the flux-dose conversion factor of ICRP-116 to the derived location-specific flux [28]. The total gamma dose rates were determined by adding the primary gamma dose rate and that resulting from the activation of spent nuclear fuel structure.
The International Commission on Radiological Protection (ICRP) has proposed collective dose assessment in ICRP Publication 103 as a means to optimize occupational radiation exposure [29]. Furthermore, the U.S. Nuclear Regulatory Commission, through NUREG-1536, mandates the calculation of collective dose for workers as a part of radiation protection measures during the operation of dry storage facilities, following operational procedures. Equation (1) presents the formula used in this study to derive the collective dose for each operation. In this study, collective doses were calculated for each operation to assess the dose received by workers transporting spent nuclear fuel. To derive the collective dose for each operation, the exposure scenarios and dose rates at each location for spent nuclear fuel transport workers based on a previously set metal overpack were utilized.
-
• Di = Collective dose for operation A (man-mSv)
-
• DRi,rate = Radiation dose rate at location i (mSv·hr−1)
-
• T = Work duration for operation A (hr)
-
• P = The number of worker for operation A (man)
3. Results and Discussions
The purpose of this study is to assess the dose to workers transporting spent nuclear fuel using metal overpack. For this purpose, a Monte Carlo computational code was used to simulate the selected overpack and set the source term, and the dose rates at each location were calculated. Finally, based on the previously defined the exposure scenario for transport workers, we carried out the assessment of collective and individual doses, considering operational factors such as work duration and the number of workers.
3.1 Radiation Dose Rate at Locations of Operation
Table 6 presents the results of the dose rate by location of operation. In this study, the MCNP code was used to derive the dose rates for gamma rays and neutrons for the previously derived locations. The results of dose rate from MCNP calculations contain uncertainty as a relative error. The uncertainty below 10% is generally reliable [27]. In this study, the uncertainty for the calculated dose rates was less than 10% at all locations. An assessment of the dose rates per location revealed that location C, positioned 10 cm below the bottom surface of the overpack, exhibited the maximum dose rate of 1.257 mSv·hr−1. Alternatively, the lowest dose rate of 0.002 mSv·hr−1 was derived at location E, 10 cm below the surface of the impact limiter. Location A and C presented significantly higher dose rates compared to other locations, with values of 0.400 and 1.257 mSv·hr−1, respectively. This is likely due to the fact that each location was positioned 10 cm from the overpack surface and lacked a neutron shielding comprising resin material, unlike other locations. However, location E, situated 10 cm from the cask surface, presented a lower dose rate because the distance increased from the source caused by the impact limiter.
Table 6
Exposure scenario of worker transporting spent nuclear fuel by using metal overpack cask
| No. | Process | Duration (min) | Distance (cm) | No. of Workers (man) |
|---|---|---|---|---|
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| Retrieval process | ||||
|
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| 1 | Remove bottom shield ring of cask | 12 | 10 | 2 |
| 2 | Remove pocket trunnion plug | 12 | 10 | 1 |
| 3 | Connect cask to VCTa | 10 | 10 | 1 |
| 4 | Transfer cask to receiving and handling area | 40 | 100 | 1 |
| 5 | Disconnect cask to VCTa | 10 | 10 | 1 |
|
|
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| Preparation process for facility off-site transportation | ||||
|
|
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| 1 | Loading cask on transportation vehicle | 20 | 100 | 2 |
| 2 | Install tie-down with cask | 6 | 10 | 2 |
| 3 | Perform a visual inspection of cask | 10 | 100 | 1 |
| 4 | Install impact limiter | 16 | 10 | 2 |
| 5 | Install personnel barrier | 10 | 10 | 2 |
| 6 | Cask dose rate and contamination survey | 17 | 10 | 3 |
aVertical Cask Transporter
3.2 Radiation Dose for Workers Transporting Spent Nuclear Fuel Using Metal Overpack
Table 7 presents the results of the dose assessment for workers transporting spent nuclear fuel using metal overpack. In this study, collective doses for each task were derived by considering the work duration and the number of workers based on the previously obtained the dose rate for each location of operation. For the retrieval workers, the total collective dose to workers was derived to be 0.783 man-mSv, with the highest collective dose of 0.503 manmSv resulting from the removal of the bottom shield ring of the cask. In contrast, the operation of transferring the cask to the receiving and handling area derived the lowest collective dose of 0.066 man-mSv. For the preparation for offsite transport work, the total collective dose to workers was derived 0.356 man-mSv, with the highest collective dose of 0.138 man-mSv resulting from the operation of measuring the contamination level and the dose rates of cask. In contrast, the operation of installing impact limiters resulted in the lowest collective dose of 0.0009 man-mSv. Consequently, the total collective dose of workers transporting spent nuclear fuel using metal overpack was derived to be 1.138 man-mSv.
Table 7
Calculated radiation dose rates by each location of operation
| Location of operation | Gamma exposure rates (mSv·hr−1) | Neutron exposure rates (mSv·hr−1) | Total exposure rates (mSv·hr−1) |
|---|---|---|---|
|
|
|||
| A | 0.309 | 0.092 | 0.400 |
| B | 0.046 | 0.054 | 0.100 |
| C | 0.559 | 0.697 | 1.257 |
| D | 0.067 | 0.095 | 0.162 |
| E | 0.002 | 0.0001 | 0.002 |
Table 8
Calculated operational exposures of worker for transportation using metal overpack
| No. | Process | Location of operation | Estimated total dose for task (man-mSv) |
|---|---|---|---|
|
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| Retrieval process | |||
|
|
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| 1 | Remove bottom shield ring of cask | C | 0.503 |
| 2 | Remove pocket trunnion plug | A | 0.080 |
| 3 | Connect cask to VCT | A | 0.067 |
| 4 | Transfer cask to receiving and handling area | B | 0.066 |
| 5 | Disconnect cask to VCT | A | 0.067 |
|
|
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| Preparation process for facility off-site transportation | |||
|
|
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| 1 | Loading cask on transportation vehicle | B | 0.066 |
| 2 | Install tie-down with cask | A | 0.080 |
| 3 | Perform a visual inspection of cask | B | 0.017 |
| 4 | Install impact limiter | E | 0.0009 |
| 5 | Install personnel barrier | D | 0.054 |
| 6 | Cask dose rate and contamination survey | D | 0.138 |
Fig. 4 presents a comparison of the dose assessment results for retrieval operations and preparation for offsite transport workers. The transport worker’s dose assessments have shown that retrieval workers are exposed to higher dose than preparation for offsite transport worker. This difference was attributed to the fact that retrieval workers are exposed to radiation from spent nuclear fuel for ~1.06 times longer than those preparing for offsite transport. Additionally, retrieval workers spent 55.7% of their work duration at locations A and C, where higher dose rates were anticipated. Alternatively, the preparation for offsite transport workers spent only 7.6% of their total work time at locations with relatively high radiation exposure potential. Conversely, they allocated 25.1% of their time to work in location E, where radiation exposure was expected to be low. Consequently, the retrieval workers, who spent more time at the relatively high-radiation exposure locations compared to the preparation for offsite transport workers, were expected to receive higher doses.
Fig. 4
Results of radiation dose assessment for worker transporting spent nuclear fuel using metal overpack.
An assessment of individual doses for workers transporting spent nuclear fuel showed that a single worker could be exposed to a dose of 0.695 mSv when handling a single metal overpack. This dose was ~3.5% of the average annual dose limit of 20 mSv for radiation workers and was sufficient for transporting 28 transport casks selected in this study [30-31]. If radiation protection measures for workers at high-radiation exposure work locations A and C during transport tasks are strengthened, it is expected that the dose to the relevant transport workers can be further reduced.
This study did not assess the dose to vehicle drivers after loading spent nuclear fuel onto transport vehicles. This was due to the uncertainty caused by the fact that the transport route and means of spent nuclear fuel in Korea are not specifically determined currently. Note that the assumptions made in this study may differ from the actual operating conditions of a spent nuclear fuel dry storage facility. Therefore, the results of this study alone are insufficient to definitively determine the dose of spent nuclear fuel transport workers. In the future, when the management methods, transport routes, and operating conditions of spent nuclear fuel in Korea are more specifically established, it will be possible to more realistically assess the dose to transport workers based on the method of assessment conducted in this study.
4. Conclusion
In this study, we assessed the dose to workers transporting spent nuclear fuel using metal overpack. For this purpose, we selected a cask and set the flux per energy group for gamma rays and neutrons, considering the design basis fuel characteristics of the cask. Then, the exposure scenario for the transport workers for the selected cask was set up. Finally, the MCNP code was used to derive the dose rates for each work location for workers transporting spent nuclear fuel and derived the collective and individual doses to workers, considering operational factors such as work duration and the number of workers.
Transport spent nuclear fuel workers using metal overpack were found to conduct a total of 11 operations to transport spent nuclear fuel in a single metal overpack to a different management facility. Additionally, 5 work locations for the transport workers were identified. As a results of the dose rate assessment at each work location, the highest dose rates of 0.400 and 1.257 mSv·hr−1 were calculated at locations A and C, located 10 cm from the upper and lower surfaces of the overpack, respectively. On the contrary, location E, located 10 cm from the surface of the impact limiter, had the lowest dose rate of 0.002 mSv·hr−1. The other work locations, B and D, showed dose rates of 0.100 and 0.162 mSv·hr−1, respectively.
The results of the dose assessment for workers transporting spent nuclear fuel using metal overpack showed that the total collective dose for workers during the transport of a single metal overpack was 1.138 man-mSv. During the transport operation, the retrieval worker and the preparation for offsite transport worker exposed to 0.783 man-mSv and 0.356 man-mSv, respectively. Among the operations conducted during the retrieval operation, the removal of the bottom shield ring from the metal overpack resulted in the highest estimated the dose to workers, with an average of 0.503 man-mSv. Among the preparation for offsite transport operations, the dose rate and contamination level measurement operations for metal overpack caused the highest radiation exposure of 0.138 man-mSv.
In the future, spent nuclear fuel will be managed by stages in Korea and the transport of spent nuclear fuel between these management stages is anticipated. Among the spent nuclear fuel management facilities, dry storage facility has different operational procedures depending on the method of storage. Therefore, the transport procedures and worker’s dose to subsequent management facilities will vary. In this study, the dose assessment was conducted for workers transporting spent nuclear fuel using metal overpack, one of the methods for dry storage of spent nuclear fuel. The results of this study are expected to be utilized in the future to strengthen radiation protection for workers during transport operations between spent nuclear fuel management stages using metal overpacks.
