Impact of PWR spent fuel variations on back-end features of advanced fuel cycles with tru-fueled VHTR

Alternative strategies are being considered as management option for current spent nuclear fuel transuranics (TRU) inventory. Creation of transmutation fuels containing TRU for use in thermal and fast reactors is one of the viable strategies. Utilization of these advanced fuels will result in transmutation and incineration of the TRU. The objective of this study is to analyze the impact of conventional PWR spent fuel variations on TRU-fueled very high temperature reactor (VHTR) systems. The current effort is focused on prismatic core configuration operated under a single batch once-through fuel cycle option. IAEA’s nuclear fuel cycle simulation system (VISTA) was used to determine potential PWR spent fuel compositions. Additional composition was determined from the analysis of United States legacy spent fuel that is given in the Yucca Mountain Safety Assessment Report. A detailed whole-core 3-D model of the prismatic VHTR was developed using SCALE5.1 code system. The fuel assembly block model was based on Japan’s HTTR fuel block configuration. To establish a reference reactor system, calculations for LEU-fueled VHTR were performed and the results were used as the basis for comparative studies of the TRU-fueled systems. The LEU fuel is uranium oxide at 15% ²³⁵U enrichment. The results showed that the single-batch core lifetimes ranged between 5 and 7 years for all TRU fuels (3 years in LEU), providing prolonged operation on a single batch fuel loading. Transmutation efficiencies ranged between 19% and 27% for TRU-based fuels (13% in LEU). Total TRU material contents for disposal ranged between 730 and 808 kg per metric ton of initial heavy metal loading, reducing TRU inventory mass by as much as 27%. Decay heat and source terms of the discharged fuel were also calculated as part of the spent fuel disposal consideration. The results indicated strong potential of TRU-based fuel in VHTR.     

Preliminary conceptual design of a geological disposal system for high-level wastes from the pyroprocessing of PWR spent fuels

The inventories of spent fuels are linearly dependent on the production of electricity generated by nuclear energy. Pyroprocessing of PWR spent fuels is one of promising technologies which can reduce the volume of spent fuels remarkably. The properties of high-level wastes from the pyroprocessing are totally different from those of spent fuels. A geological disposal system is proposed for the high-level wastes from pyroprocessing of spent fuels. The amount and characteristics of high-level wastes are analyzed based on the material balance of pyroprocessing. Around 665 kg of monazite ceramic wastes are expected from the pyroprocessing of 10 MtU of PWR spent fuels. Decay heat from monazite ceramic wastes is calculated using the ORIGEN-ARP program. Disposal modules consisting of storage cans, overpacks, and a deposition hole or a disposal tunnel are proposed. Four kinds of deposition methods are proposed. Thermal design is carried out with ABAQUS program and geological data obtained from the KAERI Underground Research Tunnel. Through the thermal analysis, the spacing between the disposal modules is determined for the peak temperature in buffer not to exceed 100 °C. Thermal analysis shows that the optimum spacing between the vertical deposition holes with 4 overpacks is 8 m when the disposal tunnel spacing is 40 m and optimum spacing of 2 m for horizontal disposal tunnel with 25 m tunnel spacing. Also, the spacing reduces to 6 m for vertical deposition when the double-layered buffer is used, which reduces the disposal area to one-sixty fifth (1/65th) compared with the direct disposal of spent fuels. Finally, the effect of cooling time on the disposal area is illustrated.     

Application of a cold spray technique to the fabrication of a copper canister for the geological disposal of CANDU spent fuels

A new method was proposed for the manufacture of a copper-cast iron canister for the spent fuel disposal based on the cold spray coating technique. The thickness of a copper shell could be fabricated to be as thin as 10 mm with the new method. Around 6 tons of copper could be saved with a 10 mm thick canister compared with a 50 mm thick canister. The electrochemical properties of the cold sprayed copper layer and forged copper were measured through a polarization test. The two copper layers showed very similar electrochemical properties. The lifetime of a 10 mm copper canister was estimated with a mathematical model based on the mass transport of sulfide ions through the buffer. The results showed that the canister lifetime was more than 140,000 years under the Korean granite groundwater condition. The thermal analysis with a current pre-conceptual design of a CANDU spent fuel canister showed that the maximum temperature between the canister and the saturated buffer was below the thermal criteria, 100 °C. Finally, the mechanical stability of the copper canister was confirmed with a computer program, ABAQUS, under the rock movement scenario.     

An investigation on recycling the recovered uranium from electro-refining process in a CANDU reactor

Feasibility studies for recycling the recovered uranium from electro-refining process of pyroprocessing into a Canada Deuterium Uranium (CANDU) reactor have been carried out with a source term analysis code ORIGEN-S, a reactor lattice analysis code WIMS-AECL, and a Monte Carlo analysis code MCNPX. The uranium metal can be recovered in a solid cathode during an electro-refining process and has a form of a dendrite phase with about 99.99% expecting recovery purity. Considering some impurities of transuranic (TRU) elements and fission products in the recovered uranium, sensitivity calculations were also performed for the compositions of impurities. For a typical spent PWR fuel of 3.0 wt.% of uranium enrichment, 30 GWD/tU burnup and 10 years cooling, the recovered uranium exhibited an extended burnup up to 14 GWD/tU. And among the several safety parameters, the void reactivity at the equilibrium state was estimated 15 mk. Additionally, a simple sphere model was constructed to analyze surface dose rates with the Monte Carlo calculations. It was found that the recovered uranium from the spent PWR fuel by electro-refining process has a significant radioactivity depending on the impurities such as fission products.     

Modeling and criticality evaluation of the voloxidation process

Voloxidation is a necessary process in the dry reprocessing of spent nuclear fuels. The criticality evaluation plays a considerable role in the design of voloxidation apparatus. As conservative results are always preferred in a criticality evaluation, an optimized model was built in consideration of both the geometry of voloxidation apparatus and the occurring forms of evaluation material. The criticality evaluation of fresh UO₂ fuel and PWR spent fuel were then performed by employing Monte Carlo techniques, respectively. It is demonstrated that there is no criticality risk concerning the voloxidation process dealing with fresh UO₂ or PWR spent fuel if water does not intrude into the cell. However, if water intrudes and mixes with the fuels, the subcritical mass limit is 40.1 ± 0.1 kg for fresh UO₂ and 19,155 ± 50 kg for spent fuel. The contributions of ¹H and ²³⁵U were analyzed quantitatively by the TSUNAMI code to clarify the competition between ¹H moderation effect and its dilution effect on the concentration of ²³⁵U.     

The role of pyro-processing in decreasing disposal cost in Korea

A comparative analysis regarding the disposal cost of HLW (High-Level Waste) from 20,000-ton PWR nuclear fuel, focusing on pyro-processing and direct disposal, was conducted in this study. A cost estimation of the major cost drivers in disposing of pyro-processed waste revealed that canisters would cost 67.32 MEUR and that the disposal holes and disposal tunnels would require about 11.2 MEUR for excavation. These estimates amount to 1/16 and 1/55 of the costs for direct disposal of PWR spent fuels, respectively. These significant disposal cost savings in pyro-processed radioactive waste result from a significant reduction in the amount of radioactive waste to be disposed of thanks to the recycling in a fast reactor.     

Characteristics of a Geological Disposal System for the Increasing Burn-up of Spent Nuclear Fuel in Korea

The characteristics of a geological disposal system that can accommodate increasingly higher burn-up levels of spent fuel were assessed based on the Korea reference disposal system concept. First, a status investigation that included a projection of spent fuel quantity versus burn-up was carried out to demonstrate the trend toward higher burn-up levels. Next, the main features of the Korea reference disposal system were introduced. Finally, the disposal tunnel length, excavation volume, and raw materials (e.g., a cast insert, copper, bentonite and backfill) necessary for a disposal system were comprehensively analyzed to define the characteristics and overall effects on geological disposal at increasingly higher burn-up levels. Our study determined that it is reasonable to use a canister containing 4 spent fuel assemblies with burn-up levels up to 50GWD/MTU, while a canister containing 3 spent fuel assemblies can accommodate burn-up levels beyond 50GWD/MTU. A remarkable increase of 33% in disposal tunnel length and that of 30% in excavation volume were observed as the burn-up increased from 50 to 60GWD/MTU. However, this was offset by a reduction of 17% in raw materials used in canister fabrication. Therefore, it seems that spent fuel at increasingly higher burn-up levels is not a serious concern for deep geological disposal in Korea.     

Comparative cost analysis of direct disposal versus pyro-processing with DUPIC in Korea

There are 20 nuclear reactors operating in Korea and four more are under construction. The spent nuclear fuel and radioactive wastes will be accumulated and an effective management must be introduced. In Korea, pyro-processing and Direct Use of spent PWR fuel In CANDU (DUPIC) are drawing attentions of many researchers and policy makers. However, cost comparison studies of each or both options against the direct disposal have not been adequately conducted. Therefore, the purpose of this study is to compare the fuel cycle strategy in cost terms. Based on mass balance of the Heavy Metal, cost of each process in the fuel cycle was considered. It was found that the pyro-processing-only cannot win against the direct disposal, but it can be compensated by adding DUPIC. Pyro-processing with DUPIC was cheaper by 925 M$ than the direct disposal. Further time-considered analysis will supplement this work with reasonable basis.     

Effect of helium on the microstructure of spent fuel in a repository: An operational approach

In a repository, the release of radionuclides from spent fuel rods will strongly depend on the pellet microstructure existing when water comes into contact with the spent fuel surface, i.e. after 10,000 years of disposal. During this period, a large quantity of He atoms is produced by α-disintegrations of actinides in the spent fuel. A conservative model is proposed here to evaluate the consequences of He on the spent fuel microstructure. According to the solubility and diffusion properties of He under repository conditions, two scenarios are considered: He atoms can be trapped in fission gas bubbles or form new bubbles. In spite of the conservative assumptions of the model, the calculated values of bubble or pore pressure are much lower than critical values derived from rupture criteria. No evolution of the microstructure of the spent UO₂ fuel is thus expected before the breaching of the canister.     

Finite element analyses and verifying tests for a shock-absorbing effect of a pad in a spent fuel storage cask

A spent fuel storage cask is required to prove the safety of its canister under a hypothetical accidental drop condition which means that the canister is assumed to be free dropped on to a pad of the storage cask during the loading of the canister into a storage cask. In this paper, finite element analyses and verifying tests for a shock-absorbing effect of a pad in a spent fuel dry storage cask were carried out to improve the structural integrity of the canister under a hypothetical accidental drop condition. The pad of the storage cask was originally designed as cylindrical steel structure filled with concrete. The pad was modified by using the structure composed of steel and polyurethane-foam instead of the quarter of the upper concrete as an impact limiter. The effects of the shape and the thickness of the steel structure and the density of the polyurethane-foam which was used in between steel structures were studied. As the optimized pad of a spent fuel dry storage cask, the quarter of the upper concrete was replaced with 12 mm thick circular steel structure and polyurethane-foam whose density was 85 kg/m³. The drop tests of a 1/3 scale model for the canister on to the original pad and the optimized pad were conducted. The effect of the pad structure was evaluated from the drop tests. The optimized pad has a greater shock-absorbing effect than the original pad. In order to verify the analysis results, strains and accelerations in the time domain by the analytical methods were compared with those by a test. The numerical method of simulating the free drop test for a dry storage cask was verified and the numerical results were found to be reliable.     

Cost efficiency analysis of disposal alternatives in Korea

This research presents the results of calculating the disposal cost efficiency for the four disposal alternatives for the CANDU spent fuel that are under development in Korea currently. The KRS-1 alternative, developed first, was set as the standard, and the efficiency of the KRS-1 alternative was assumed to be 100%.The cost calculation result shows that the A-KRS-22, which was developed most recently among the CANDU spent fuel disposal alternatives, manifested −61.7%, −45.7%, −47.0%, −78.9% and −61.7% when compared to the KRS-1 alternative concerning disposal tunnel excavation, disposal hole excavation, bentonite, disposal canister and backfilling.Moreover, the cost calculation method for the dominant cost driver that uses the unit disposal module concept for the calculation of cost efficiency was used. As for the reason that the standard for efficiency measurement was taken per each bundle, it is because the amount of bundle capacity concerning the spent fuel differs by disposal canister.     

乏燃料密封容器开盖及其内容物回取技术研究

对于采用干湿法贮存的乏燃料而言,其后处理时面临的最大问题是如何安全高效地将乏燃料等内容物从封焊的密封容器中取出。针对这一问题,基于乏燃料密封容器及其内容物的结构特点,开展了乏燃料密封容器开盖及内容物回取技术研究,综合考虑切割热室使用环境、内容物回取后的收集和转移以及产生废物的收集处理等因素,制定了合理可行的开盖及回取工艺,研制了用于开盖和筒体分段切割的解体装置以及回取和吊装工具,并通过试验验证了工艺的可行性以及研制的工装具的可用性。      

Analysis of the processes in spent fuel pools of Ignalina NPP in case of loss of heat removal

Ignalina NPP is the only nuclear power plant in Lithuania consisting of two units, commissioned in 1983 and 1987. Unit 1 of Ignalina NPP was shutdown for decommissioning at the end of 2004 and Unit 2 is to be operated until the end of 2009. Both units are equipped with channel-type graphite-moderated boiling water reactors RBMK-1500. According to the design, the spent fuel should be returned for reprocessing to Russia. However actually any fuel assembly has not been taken out from territory of the Ignalina NPP and all assemblies of spent fuel are stored in the spent fuel pools and dry on-site storage facility. Thus, the safety justification of facilities for intermediate spent fuel assemblies’ storage in Ignalina NPP is very important. This paper presents the results of loss of heat removal accidents (the most probable beyond design basis accident) in spent fuel pools of Ignalina NPP. The analysis was performed by employing best-estimate system thermal hydraulic code RELAP5 and codes for severe accidents ATHLET-CD and ASTEC. The best-estimate analysis, performed using RELAP5, allowed to investigate in the details the water evaporation, uncovering and fuel assemblies heat-up processes, when heat removal from the structures of buildings and pools are evaluated. The processes of spent fuel assemblies’ degradation due to loss of long-term heat removal were analyzed using ATHLET-CD and ASTEC codes. The results of calculations showed that the increase in water temperature in the pools from 50 °C up to 100 °C takes approximately 80-110 h, the evaporation of water volume down to uncovering of fuel assemblies takes approximately 220-260 additional hours. Later, after 200-300 h, the temperature of fuel claddings exceeds 800-1000 °C and the failures of fuel claddings occur due to cladding ballooning. The total amount of hydrogen generated up to time of complete water evaporation from spent fuel pools is about 7500-16,000 kg. These results of performed analysis were used for development of accident management guidelines for spent fuel pools of RBMK-1500.     

Review of the development of the transportation,aging, and disposal (TAD) waste disposal system for the proposed Yucca Mountain geologic repository

The U.S. Department of Energy (DOE) began studying Yucca Mountain in 1978 to determine whether it would be suitable for the nation’s first long-tem geologic repository for over 70,000 metric tons of spent (or used) nuclear fuel and high-level radioactive waste. The purpose of the continuing Yucca Mountain study, or project, is to comply with the Nuclear Waste Policy Act of 1982 as amended in 1987 and develop a national disposal site for spent nuclear fuel and high-level radioactive waste disposal. In 2005, DOE shifted the design of the proposed repository from a concept of unloading spent nuclear fuel from transportation canisters and loading into disposal canisters (which required a great deal of handling radioactive material at the repository site) to a “clean” facility, unveiling the transportation, aging, and disposal (TAD) canister system. The TAD waste system consists of a canister loaded with commercial spent nuclear fuel.This review paper provides a comprehensive review on the status of TAD, technical and licensing requirements, the work that has been done so far, and the challenges and issues that must be addressed before TAD can be successfully implemented. Though the future of the Yucca Mountain project is bleak at this point, the progress that has come in the field of TAD will be one of its lasting legacies.     

Current status of spent fuels and the development of computer programs for the PWR spent fuel management in Korea

Nuclear power has supplied the national electric power demand for three decades in the Republic of Korea, which has resulted in the accumulation of a large amount of spent fuels. The government has a policy on the temporary storage of these at nuclear power plants at present. In order to establish a proper policy for spent fuel management in the near future, the characteristics and amount of spent fuels should be figured out properly. In this paper, the current status of spent fuels in the Republic of Korea is outlined focusing on the major characteristics of spent fuels such as initial enrichment and discharge burnup. According to the current trend, the average burnup of PWR spent fuels will reach 55 GWd/MtU by the middle of 2010s. Three different kinds of computer programs were developed to supply crucial data regarding spent fuels. The first one was developed to project the amount of spent fuels in the future based on three different projection models. The projection was verified with real spent fuel data. The second Database program was prepared for the analysis of statistics regarding PWR spent fuels. Each PWR spent fuel assembly was specified with 18 items of data such as fuel type, initial enrichment, and discharge burnup. The usefulness of the Database program was illustrated through an analysis of the geological disposal density and cooling time of PWR spent fuels. Disposal area could be reduced by 50% through a proper analysis of the cooling time of PWR spent fuels. Finally, A-SOURCE program was developed to easily calculate source-terms such as decay heat and radionuclide concentration after the pyro-processing of PWR spent fuel assemblies. Linked to the Database program, the A-SOURCE program selected PWR spent fuel assemblies and could calculate the source-terms for any combination of them. An illustration of the usage of the program was demonstrated.     

Development of the detecting method of helium gas leak from canister

In the concrete cask storage system, spent fuel is installed and weld-sealed in a cylindrical container called a canister. The canister is filled with helium gas and its containment shall be maintained and inspected during storage. The helium gas enhances heat removal from spent fuel. When the helium gas leaks, the effect of helium gas convection is weakened in the canister. Thereof, the temperature on the canister surface changes.In present tests, it was found that temperatures of the center of the top and the bottom on the canister surface change remarkably during the helium gas leak. Therefore, we defined the temperature difference as ΔT_BT. And one can detect helium gas leak using the change of ΔT_BT. ΔT_BT increases monotonously toward a constant value during helium gas leak, even if the inlet air temperature drops. The helium gas leak can be detected at the early stage of the leak by observing both ΔT_BT and inlet air temperature.     

High-density support matrices: Key to the deep borehole disposal of spent nuclear fuel

Deep (4-5 km) boreholes are emerging as a safe, secure, environmentally sound and potentially cost-effective option for disposal of high-level radioactive wastes, including plutonium. One reason this option has not been widely accepted for spent fuel is because stacking the containers in a borehole could create load stresses threatening their integrity with potential for releasing highly mobile radionuclides like ¹²⁹I before the borehole is filled and sealed. This problem can be overcome by using novel high-density support matrices deployed as fine metal shot along with the containers. Temperature distributions in and around the disposal are modelled to show how decay heat from the fuel can melt the shot within weeks of disposal to give a dense liquid in which the containers are almost weightless. Finally, within a few decades, this liquid will cool and solidify, entombing the waste containers in a base metal sarcophagus sealed into the host rock.     

The economic effects of the deferred disposal of spent fuel in Korea

The present study analyzes the economic effects concerning deferred disposal of spent fuel through long-term storage. According to the cost analysis, a scenario that a 90-year deferral of an HLW (High-Level Waste) repository construction in favor of a long-term storage of spent fuel would be economically preferable to another scenario based on the year 2040 chosen as the starting point for construction on a repository. That is, the former scenario would cost about 1/2 of the latter. This finding is an estimated result from an economic perspective only, assuming the disposal of 20,000-ton PWR spent fuel and 16,000-ton CANDU spent fuel. Still, it seems necessary to elicit proper term of storage for radioactive waste in order to comply with the so-called Polluter-Pays principle that the current generation cannot pass on its radioactive waste to the next generation.     

Development of monitoring technique for the confirmation of spent fuel integrity during storage

A method is developed to monitor integrity of spent fuels stored in a canister that is sealed by weld. To achieve the monitoring, Kr-85 gas was newly adopted as a kind of probe. In the case of a fuel rod defect, Kr-85 gas of the fuel rod is leaked in the canister. By detection of gamma ray (514 keV) emitted from Kr-85 outside of the canister, defected rods can be detected without unsealing the canister. The monitoring technique was developed using small-scaled mock-up experiments and simulated calculation. The result showed that Kr-85 gas leakage of about 10¹¹ Bq is detectable under the noise gamma rays by using the detection system with collimator, which is about 10% of Kr-85 inventory in a fuel rod. Therefore, this monitoring technique is considered as an inspection method prior to transportation of spent fuel from an interim storage facility to a reprocessing plant.     

MCNP simulation of neutron energy spectra for a TN-32 dry shielded container

This paper presents a numerical analysis of neutron energy spectra for a TN-32 spent fuel dry storage cask using Monte Carlo simulation. The analysis results were compared with experimental measurements to determine the suitability of using such codes for neutron flux calculations in soft-spectrum neutron environments. Complete spent fuel compositions were generated using Scale 4.4a. Variations in source definition and geometry determined that geometric and source simplifications in the computational model have negligible effect on final neutron energy distribution. Variations between experimental and computed spectra at energies above 1 MeV and below 100 keV demonstrated the shortfalls of various detection instruments used to collect the experimental neutron energy spectra data principally because these instruments were calibrated based on high neutron energy spectra. The MCNP calculations were generally in agree with the experimental data, but predicted that the detectors would over-respond to the neutron spectra around a spent fuel dry shielded container. Computed neutron energy spectra were always conservative when compared to experimental spectra.