Radioactive Waste Arisings from Various Fuel Cycle Options

This study examines all kinds of waste volumes from various fuel cycle options including DUPIC (Direct Use of Spent PWR Fuel In CANDU) fuel cycle and compares each other. The fuel cycle option considered the PWR (Pressurized Water Reactor) once-through cycle, the PHWR (Pressurized Heavy Water Reactor) once-through cycle and the thermal recycling option using an existing PWR with MOX (Mixed Oxide) fuel. This study focuses on the radioactive wastes including mill waste, low-level waste and high-level waste generated by all fuel cycle steps, which can be one of the effectiveness measures of waste management. All waste disposition volume is estimated in terms of m³/GWe-yr. We find in the estimation of radioactive waste volume that PWR-MOX option has the lowest mill tailings and spent fuel volumes among the options, but the option has high volume of ILW and HLW. Mill tailings and spent fuel volumes of the DUPIC fuel cycle are lower than those of other competitive options such as PWR-PHWR once-through cycle. PWR once-through cycle has the lowest LLW and ILW volume among the options, but has high mill tailings and spent fuel volume. The data obtained in this study would be helpful to further estimate environmental effect and/or waste disposition costs in various fuel cycle options.     

Development of LOcal-scale High-resolution atmospheric DIspersion Model using Large-Eddy Simulation. Part 5: detailed simulation of turbulent flows and plume dispersion in an actual urban area under real meteorological conditions

We have developed a LOcal-scale High-resolution atmospheric DIspersion Model using Large-Eddy Simulation (LOHDIM-LES) to assess the safety at nuclear facilities and to respond to emergency situations resulting from accidental or deliberate releases of radioactive materials (e.g., a terrorist attack in an urban area). In parts 1–4, LESs of turbulent flows and plume dispersion over a flat terrain, around an isolated building, within building arrays with different obstacle densities, and within an actual urban area were performed, which showed the basic performance comparable to wind tunnel experimental technique. In this study, we extend the LOHDIM-LES to turbulent flows and plume dispersion in an actual urban area under real meteorological conditions by coupling with a meso-scale meteorological simulation model. The LES results of wind speed, wind direction, and concentration values are generally reproduced well. It is concluded that our coupling approach between LES and meso-scale meteorological models is effective in detailed simulations of turbulent flows and plume dispersion in urban areas under real meteorological conditions.     

Forty-One Years with Zirconium

The MEGAPIE (Megawatt Pilot Experiment) Project has been primarily initiated in response to interest in the Accelerator Driven System (ADS). The ADS is a nuclear reactor concept with a fuel cycle which can burn minor actinide waste products from conventional reactors, while also taking advantage of the safety characteristics of a sub-critical reaction process. By this means, the fission chain reaction is maintained by additional neutrons generated by protons in a liquid-metal spallation target. The goal of the international MEGAPIE project is to design and construct such a target and install and test it over a period of one year in the SINQ spallation neuron source facility at the Paul Scherrer Institut (PSI). To aid the design process, Computational Fluid Dynamics (CFD) is being used to optimise the thermal- hydraulic behaviour of the target. Results obtained so far indicate that it should be possible to remove 700 kW of heat deposited by the proton beam in the target, under steady-state conditions (40kg/s total Lead Bismuth Eutectic flow rate), without peak window temperature rising above about 385°C, when using a bypass flow of 2.5 kg/s and a slanted- end guide tube. A tentative peak window temperature of 400°C is currently considered allowable, on the basis of material strength after irradiation, and CFD simulation is currently being validated by suitable LBE experiments in similar geometry.     

Uranium Crystallization Test with Dissolver Solution of Irradiated Fuel

The crystallization process has been developed as a part of the advanced aqueous process, NEXT (New Extraction System for TRU recovery) for fast reactor (FR) cycle. In this process, a large part of U is separated from dissolver solution by crystallization as UO₂(NO₃)₂.6H₂O. The U crystallization test was carried out with real dissolver solution of irradiated FR fuel to investigate the influence of cooling rate on the crystal size and the behavior of fission product (FP) compared with that of Pu(IV). In regard to the influence of the cooling rate, it was confirmed that the crystal size was smaller as the cooling rate is faster. Although it was expectable that the decontamination performance was improved by diminishing the specific surface of the crystals, it was suggested that a large crystal produced by crystallization was not always high purity. Concerning the behavior of FPs, Eu behaved similarly to Pu(IV). Cs accompanied with U into the crystals under the condition in this test.     

Interpretation of Acoustic Parameters Obtained by EMAR Measurement for Non-destructive Hydrogen Concentration Measurement in Zr Alloy

An obvious quantitative relation between hydrogen concentrations in zirconium alloy and acoustic anisotropy parameters obtained by the electromagnetic acoustic resonance (EMAR) method was reported. To elucidate the mechanism, the acoustic parameters were calculated based on the elastic theory and the equation of motion. The acoustic parameters obtained by the EMAR method were interpreted quantitatively using the anisotropic elastic constants of the specimen, and values calculated from texture data for non-hydrogen charged specimens showed good agreement with those obtained by the EMAR method. Calculated temperature dependence of the acoustic anisotropy for the non-hydrogen charged specimen also agreed well with that by the EMAR method.

The consistencies demonstrated that the absolute values of the acoustic parameters for non-hydrogen charged specimens can be calculated from both the texture data of (0002) pole figure and the elastic constants of the specimen. Hydrogen addition up to approximately 650 ppm was found not to change the original (0002) pole figure and, correspondingly, no hydrogen concentration dependence of the acoustic parameters was obtained from the calculation. These results implied that the zirconium hydride itself played an important role for the change in the acoustic parameters of the hydrogen charged specimens, and the importance of obtaining the information on the elastic constants of the zirconium hydride was pointed out.     

Development of a Code MOGRA for Predicting the Migration of Ground Additions and Its Application to Various Land Utilization Areas

A Code MOGRA (Migration Of GRound Additions) is a migration prediction code for toxic ground additions including radioactive materials in a terrestrial environment, which consists of computational codes that are applicable to various evaluation target systems, and can be used on personal computers. The computational code has the dynamic compartment analysis block at its core, the graphical user interface (GUI) for model formation, computation parameter settings, and results displays. The compartments are obtained by classifying various natural environments into groups that exhibit similar properties.

The functionality of MOGRA is being verified by applying it in the analyses of the migration rates of radioactive substances from the atmosphere to soils and plants and flow rates into the rivers. In this report, a hypothetical combination of land usage was supposed to check the function of MOGRA. The land usage was consisted from cultivated lands, forests, uncultivated lands, urban area, river, and lake. Each land usage has its own inside model which is basic module. Also supposed was homogeneous contamination of the surface land from atmospheric deposition of ¹³⁷Cs (1.0 Bq/m ²). The system analyzed the dynamic changes of ¹³⁷Cs concentrations in each compartment, fluxes from one compartment to another compartment.     

Oxygen Diffusion Analysis of Lead-Bismuth-Cooled Natural-Circulation Reactor

Lead-Bismuth Eutectic (LBE) coolant has been selected as one of the possible options for fast breeder reactors (FBR). To control a corrosion of structural materials in LBE is a key issue for LBE-cooled reactor feasibility, so that oxygen concentration is necessary to be controlled in the appropriate range.

A concentration diffusion analysis code (COCOA: Concentration COntrol Analysis code) was developed to evaluate diffusivity and controllability of oxygen in the LBE-cooled natural-circulation reactor design. The controllability was investigated through the sensitivity surveys on two ordinary models of oxygen control (ON-OFF and PID methods) and detection regions of oxygen concentration, which is referred in the control system.

It is concluded that the oxygen concentration should be detected at the center of the secondary vortex near the inner shell for both control methods to decrease the peak concentration below the upper constraint. The PID control method can decrease the peak concentration 400 s faster than the ON-OFF control method, when concentration is detected at the center of the vortex. It is also concluded that the diffusion and the equilibrium coefficients in LBE have negligible effects on the oxygen controllability through the sensitivity survey. It is prospected that oxygen concentration control is possible even in the natural-circulation reactor.     

压水堆乏燃料干式贮存的安全考量

通过对压水堆乏燃料干式贮存的准备、运输和贮存三个工艺区域的设备设施、操作过程和环境及其潜在的危险因素的研究, 重点针对作用于干式贮存设施的临界安全、放射性物质的包容、衰变热的移除和辐射防护的影响因素开展分析, 提出了针对性的防控对策;同时总结了秦山第三核电厂重水堆乏燃料干式贮存实施过程安全操作和管理的经验。研究成果可为解决压水堆乏燃料干式贮存设施的规划、设计、建设和运行过程有关安全问题提供思路。     

研究堆铝包壳元件水池贮存的腐蚀问题

评述了研究堆乏燃料管理的重要性；总结了影响研究堆元件铝包壳腐蚀的因素；从美国和其它一些国家的经验与教训，指出关键是池水的水质；介绍了减轻铝包壳腐蚀的措施。     

Aging management and post-storage transport of dual purpose cask for spent fuel

Aging management of spent fuel storage facility may follow lessons learned from literature for nuclear power plant and a review for spent fuel dry cask storage system by US NRC, DOE, by German BAM, that by Japan NISA, etc. Namely, the essence of systematic approach to aging management includes Understanding aging, Plan (Development and optimisation of activities for aging management), Do (Managing aging mechanisms), Check (Monitoring, inspection and assessment), and Act (Maintenance). The PDCA cycle will optimise the systematic approach to the aging management. An aging management programme (AMP) for the storage system over the period of extended storage will address uncertainties in the safety relevant functions of the system that may otherwise be impaired by aging mechanisms. The AMP identifies system, structure and components (SSCs) that need specific actions to mitigate aging and ensures that no aging effects result in a loss of their intended function of the SSCs, during an intended licensed period. AMPs generally include Prevention, Mitigation, Monitoring, Inspection, and Maintenance programmes. Aging management plans should ensure compliance with transportation requirements after extended storage. Potential issue would be a significant change of the transport regulations in the future. If the regulations changed significantly, a gap analysis should be performed to identify any impact to the cask safety. Compensating arrangements, if necessary, should be proposed at that time. Assuming that the regulations will not change significantly after long term storage, we will be able to renew the license both for transport and storage of the cask during the storage period. For example, in Japan, a holistic approach was established for the license of a 50 year storage and transport. In this approach, we can evaluate integrity of spent fuel, basket, etc. with respect to chemical, thermal, mechanical, and radiation factors. With this approach we will not have to open the cask lid for visual inspection of the spent fuel, basket, etc. prior to the post-storage transport.     

Radiotoxicity and Residual Energy Release of Actinides and Fission Products of Spent VVÉR Fuel in Long-Term Storage

The time dependence of the residue energy release and radiotoxicity of spent VVÉR-1000 nuclear fuel with long-term storage or uniform accumulation in long-term storage is investigated. The calculations of the energy release take account of the contribution of , , and radiation, and the calculations of the radiotoxicity take account of the maximum admissable activity of nuclides in water and air. The data presented can be used for developing a strategy for long-term storage of spent nuclear fuel from power reactors. 4 figures, 4 tables, 4 references.     

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.     

Calculating the transient temperature field in a reactor tube and the thermoelastic stresses in the fuel element cladding

Approximate analytic methods are given for calculating the transient temperature field in the fuel elements and the coolant temperatures at any point along the reactor tube, as well as the transient thermoelastic stresses in the cladding of a cylindrical fuel element. The coolant temperature at the input to the tube is constant, and the coolant undergoes no changes in state of aggregation. The approximate methods are illustrated by examples.Results are given, for comparison, of accurate calculations of the same examples made with a rapid calculating machine.List of symbols time - r; z coordinates (radius, distance along tube) - r₁; r₂ internal and external radii of fuel element cladding respectively - H total active length of fuel element - a₁; ₁;c _{1 1} coefficients of temperature conductivity, heat conductivity, specific heat capacity and specific gravity of fissionable material respectively - a₂; ₂; C_p2; ₂ cladding parameters - a; ; c_p; coolant parameters - mean cladding radius - f:f₂ cross-sectional area of tube for coolant and cladding respectively - w coolant velocity - coefficient of heat release to coolant - t (r, ); (); () fuel temperature, mean temperature over cross section of cladding, and coolant temperature at pointz. along tube respectively - qv() specific volume of coolant at pointz - values averaged overz - quantities at the initial instant of time - ₃ delay time - _n time required for coolant to go from z=0 to the point in question     

Feasibility of the electrochemical way in molten fluorides for separating thorium and lanthanides and extracting lanthanides from the solvent

An alternative way of reprocessing nuclear fuel by hydrometallurgy could be using treatment with molten salts, particularly fluoride melts. Moreover, one of the six concepts chosen for GEN IV nuclear reactors (Technology Roadmap - http://gif.inel.gov/roadmap/) is the molten salt reactor (MSR). The originality of the concept is the use of molten salts as liquid fuel and coolant. During the running of the reactor, fission products, particularly lanthanides, accumulate in the melt and have to be eliminated to optimise reactor operation. This study concerns the feasibility of the separation actinides-lanthanides-solvent by selectively electrodepositing the elements to be separated on an inert (Mo, Ta) or a reactive (Ni) cathodic substrate in molten fluoride media. The main results of this work lead to the conclusions that:

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The solvents to be used for efficient separation must be fluoride media containing lithium as cation.

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Inert substrates are suitable for actinide/lanthanide separation; nickel substrate is more suitable for the extraction of lanthanides from the solvent, owing to the depolarisation occurring in the cathodic process through alloy formation.

     

3D Neutronic Analysis in MHD Calculations at ARIES-ST Fusion Reactors Systems

In this study, we developed new models for liquid wall (FW) state at ARIES-ST fusion reactor systems. ARIES-ST is a 1,000 MWe fusion reactor system based on a low aspect ratio ST plasma. In this article, we analyzed the characteristic properties of magnetohydrodynamics (MHD) and heat transfer conditions by using Monte-Carlo simulation methods (ARIES Team et al. in Fusion Eng Des 49–50:689–695, 2000; Tillack et al. in Fusion Eng Des 65:215–261, 2003) . In fusion applications, liquid metals are traditionally considered to be the best working fluids. The working liquid must be a lithium-containing medium in order to provide adequate tritium that the plasma is self-sustained and that the fusion is a renewable energy source. As for Flibe free surface flows, the MHD effects caused by interaction with the mean flow is negligible, while a fairly uniform flow of thick can be maintained throughout the reactor based on 3-D MHD calculations. In this study, neutronic parameters, that is to say, energy multiplication factor radiation, heat flux and fissile fuel breeding were researched for fusion reactor with various thorium and uranium molten salts. Sufficient tritium amount is needed for the reactor to work itself. In the tritium breeding ratio (TBR) >1.05 ARIES-ST fusion model TBR is >1.1 so that tritium self-sufficiency is maintained for DT fusion systems (Starke et al. in Fusion Energ Des 84:1794–1798, 2009; Najmabadi et al. in Fusion Energ Des 80:3–23, 2006).     

Operational experience of on-site transport of spent nuclear fuel in Korea

Abstract

With the rapid development of the nuclear power programme in Korea, the amount of accumulated spent nuclear fuel has inevitably increased year by year. The spent nuclear fuel is being stored in on-site storage pools at the nuclear power plants. As the current storage capacity for spent nuclear fuel is insufficient, at-reactor storage is being expanded at each site with regard to optimisation of technical and economic factors. On-site transport between neighbouring reactors has been necessary to secure sufficient storage capacity for pressurised water reactor spent nuclear fuel assemblies. A complete on-site transport system has been developed, and so far more than 800 spent nuclear fuel assemblies have been transported using two kinds of transport cask.     

Radiation Conditions during the Preparation of Irradiated BOR-60 Fuel Assemblies for Reprocessing

The basic stages in the preparation of irradiated BOR-60 reactor fuel for reprocessing are examined. It is determined that during the separation of the fuel part of the fuel elements the coefficient of transfer of ¹³⁷Cs from the fuel into aerosol is 5·10^–6 and for fragmentation the value is 3·10^–5. It is found that the real catching efficiency for aerosol particles caught by a V-05 filter ranges from 42 to 99%. The specific entry of radioactive aerosols into the ventillation center after the first stage of air purification was 0.3 MBq for -emitters and 7.7 MBq for and emitters per 1 kg of reprocessed fuel. The total collective dose formed at the stages of preparation of a large batch of irradiated fuel (four spent fuel asemblies with average burnup 11.4% and a 10.5 to 23.7 yr holding period) for reprocessing was 11.5·10^–3 persons·Sv.     

The adsorption and desorption of water from a carbonaceous deposit layer on the surface of stainless steel representing spent AGR nuclear fuel cladding

Spent fuel discharged from advanced gas-cooled reactor power stations carries a deposit of carbon firmly attached to the cladding surface. The fuel route involves contact with water, for cooling and transport. Long-term storage potentially includes dry storage, however, the carry-over of water entrained within the carbon deposit needs to be considered regarding the storage environment. Drying of the fuel is possible, but little is known concerning the drying characteristics of such deposits. This work reports preparation of a laboratory simulant of a carbon deposit on a fuel pin surface and measurement of its adsorption and desorption properties regarding liquid and vapour phase water. This work found that water vapour equilibration is rapid and reversible. Liquid water uptake is appreciable (up to 5.7 times the mass of carbon) and most (up to 88%) is removed on standing for 12 h. Heating removes little more. The implications for spent fuel management are discussed.