Showing papers in "Journal of Nuclear Science and Technology in 2011"
TL;DR: The fourth version of the Japanese Evaluated Nuclear Data Library has been produced in cooperation with the Japanese Nuclear Data Committee as mentioned in this paper, and much emphasis is placed on the improvement of the original library.
Abstract: The fourth version of the Japanese Evaluated Nuclear Data Library has been produced in cooperation with the Japanese Nuclear Data Committee. In the new library, much emphasis is placed on the impro...
1,699 citations
TL;DR: In this article, the preliminary estimation of release amounts of 131I and 137Cs from the Fukushima Daiichi nuclear power plant into the atmosphere has been presented, with the aim to estimate the amount of radiation released by the plant.
Abstract: (2011). Preliminary Estimation of Release Amounts of 131I and 137Cs Accidentally Discharged from the Fukushima Daiichi Nuclear Power Plant into the Atmosphere. Journal of Nuclear Science and Technology: Vol. 48, No. 7, pp. 1129-1134.
896 citations
TL;DR: In this article, preliminary numerical experiments on Oceanic Dispersion of 131I and 137C Discharged into the Ocean because of the Fukushima Daiichi Nuclear Power Plant Disaster are presented.
Abstract: (2011). Preliminary Numerical Experiments on Oceanic Dispersion of 131I and 137Cs Discharged into the Ocean because of the Fukushima Daiichi Nuclear Power Plant Disaster. Journal of Nuclear Science and Technology: Vol. 48, No. 11, pp. 1349-1356.
211 citations
TL;DR: In this article, the equivalence of the first-order adjoint-weighted perturbation (AWP) method and firstorder differential operator sampling (DOS) method with fission source perturbations is proven.
Abstract: The adjoint-weighted perturbation (AWP) method, in which the required adjoint flux is estimated in the course of Monte Carlo (MC) forward calculations, has recently been proposed as an alternative to the conventional MC perturbation techniques, such as the correlated sampling and differential operator sampling (DOS) methods. The equivalence of the first-order AWP method and first-order DOS method with the fission source perturbation taken into account is proven. An algorithm for the AWP calculations is implemented in the Seoul National University MC code McCARD and applied to the sensitivity and uncertainty analyses of the Godiva and Bigten criticalities.
63 citations
TL;DR: In this paper, the progress of the design study and research and development (R&D) for the Japan Sodium-cooled Fast Reactor (JSFR) implemented in the “Fast Reactor Cycle Technology Development (FaCT)” project is described.
Abstract: This paper describes the progress of the design study and research and development (R&D) for the Japan Sodium-cooled Fast Reactor (JSFR) implemented in the “Fast Reactor Cycle Technology Development (FaCT)” project. A sodium-cooled fast reactor with an electric power of 1,500MWe is targeted for commercialization at around 2050, and a demonstration reactor assuming a power output from 500 to 750MWe is planned to start operation at around 2025. R&D on innovative technologies to achieve economic competitiveness and enhance reliability and safety is carried out for the commercialization. A compact reactor vessel without a vessel wall cooling system is pursued in consideration of the wall thickness enough to resist the severest seismic condition. A two-loop cooling system with shortened highchromium steel piping is a crucial feature, and studies on the hydraulics in the pipe elbow and the fabrication capability of the pipes are being carried out. A double-walled straight tube steam generator is investigated to...
62 citations
TL;DR: Much more accurate and reliable prediction for neutronic parameters for both thermal and fast reactors becomes possible by using the library JENDL-4.0.
Abstract: Benchmark testing for the newly developed Japanese evaluated nuclear data library JENDL-4.0 is carried out by using a huge amount of integral data. Benchmark calculations are performed with a continuous-energy Monte Carlo code and with the deterministic procedure, which has been developed for fast reactor analyses in Japan. Through the present benchmark testing using a wide range of benchmark data, significant improvement in the performance of JENDL-4.0 for fission reactor applications is clearly demonstrated in comparison with the former library JENDL-3.3. Much more accurate and reliable prediction for neutronic parameters for both thermal and fast reactors becomes possible by using the library JENDL-4.0.
60 citations
TL;DR: The Moving Particle Semi-implicit (MPS) method is one of the particle methods in which continuum mechanics is analyzed using the concept of particles, and has been applied to multiphase flow analysis in nuclear engineering.
Abstract: The Moving Particle Semi-implicit (MPS) method is one of the particle methods in which continuum mechanics is analyzed using the concept of particles. Since meshes are not used, large deformation of free surfaces and material interfaces can be simulated without the problems of mesh distortion. Thus, the MPS method has been applied to multiphase flow analysis in nuclear engineering. The advantages of the particle methods are also useful for applications in other engineering fields: ship engineering, civil engineering, microflow, biomechanics, visualization, etc. In this review, calculation examples are described and classified. Commercial codes have been released and applied in industries. The particle methods are also used in TV programs, movies, and computer games. Combinations of numerical techniques for multiphysics problems, fast calculations, and high-quality visualizations are expected to lead to real-time particle simulations for various new applications in the near future.
56 citations
TL;DR: The present JSFR design can exclude severe power burst events and Phenomenological consideration of general characteristics and preliminary evaluations for the long-term material relocation and cooling phases gave the perspective that in-vessel retention would be attained with appropriate design measures.
Abstract: In the Japan Sodium Cooled Fast Reactor (JSFR) design, elimination of severe power burst events in the Core Disruptive Accident (CDA) is intended as an effective measure to ensure retention of the core materials within the reactor vessel. The design strategy is to control the potential of excessive void reactivity insertion in the initiating phase by selecting appropriate design parameters such as maximum void reactivity on one hand, and to exclude core-wide molten-fuel-pool formation, which has been the main issue of CDA, by introducing an inner duct on the other hand. The effectiveness of these measures is evaluated based on existing experimental data and computer simulation with validated analytical tools. It is judged that the present JSFR design can exclude severe power burst events. Phenomenological consideration of general characteristics and preliminary evaluations for the long-term material relocation and cooling phases gave the perspective that in-vessel retention would be attained with appropri...
51 citations
TL;DR: In this article, a 6-nylon fiber was used as a starting material for the removal of cesium-137 ion from water contaminated with radioisotopes, and precipitates of metal ferrocyanide (M-FC) were formed among the polymer chains grafted onto the fiber.
Abstract: Radioisotopes are currently being released by Fukushima Daiichi nuclear power plant, which was damaged by the magnitude 9.0 earthquake and the tsunami that followed. Cesium-137, cesium-134, and iodine-131 produced by the fission of nuclear fuels are expected to seriously contaminate the sea and ground. Cesium and iodine dissolve in water and seawater in ionic forms of Csþ and I or IO3 , respectively. Cesium-137, with a longer half-life (30 years) than iodine131 (8 days), should be removed and confined. Adsorption is a feasible method of removing radioactive species from water contaminated with radioisotopes. Zeolites and insoluble metal ferrocyanides are known as adsorbents capable of specifically capturing cesium-137 ions. Moreover, the insoluble metal ferrocyanides are loaded onto solid supports in bead or granule form such as porous anionexchange resins, silica gels, and zeolites. However, fibrous adsorbents are effective for easy operation at sites for processing water contaminated with radioisotopes. Here, we propose a novel fiber that can immobilize metal ferrocyanides for the removal of cesium ions using a commercially available 6-nylon fiber as a starting material. Precipitates of metal ferrocyanide (M-FC) were formed among the polymer chains grafted onto the 6-nylon fiber. Therefore, the M-FC-impregnated fibers with a small diameter enhance the mass transfer of cesium ions from the bulk of liquid to the M-FC particles, leading to the rapid removal of cesium ions. In addition, the M-FC-impregnated fibers can be fabricated into various fiber modules for easy operation at the sites. In this study, the cobalt ion was used as a metal ion that precipitates with potassium ferrocyanide. II. Experimental
45 citations
TL;DR: In this paper, a series of experiments were carried out to investigate phenomena from bubble nucleation to lift-off for a subcooled boiling flow in a vertical annulus channel, and a high-speed digital video camera was used to capture the bubble dynamics.
Abstract: A series of experiments were carried out to investigate phenomena from bubble nucleation to lift-off for a subcooled boiling flow in a vertical annulus channel. A high-speed digital video camera was used to capture the bubble dynamics. The bubble lift-off diameter and bubble nucleation frequency were evaluated in terms of heat flux, mass flux, and degree of subcooling. The fundamental features of the lift-off diameter and nucleation frequency (i.e., the variations across nucleation sites and the dependence on the flow and heat flux conditions) were addressed based on the present observation. A database for the bubble lift-off diameter was built by gathering and summarizing the data of Prodanovic et al., Situ et al., and the present experiments. We evaluated the predictive capabilities of Unal's model, Situ et al.'s model, and Prodanovic et al.'s correlation against the database. We obtained the best prediction results by modifying the wall superheat correlation in Unal's model. In addition, we suggested a...
44 citations
TL;DR: In this article, the mitigation effect brought by the water film, liquid droplet impingement (LDI) on a wet rigid wall has been simulated using the moving particle semi-implicit (MPS) method.
Abstract: In order to investigate the mitigation effect brought by the water film, liquid droplet impingement (LDI) on a wet rigid wall has been simulated using the moving particle semi-implicit (MPS) method...
TL;DR: In this article, an analysis of core Melt Accident in Fukushima Daiichi-Unit 1 nuclear Reactor is presented, where the core-melter was removed from the core of the Reactor.
Abstract: (2011). Analysis of Core Melt Accident in Fukushima Daiichi-Unit 1 Nuclear Reactor. Journal of Nuclear Science and Technology: Vol. 48, No. 8, pp. 1135-1139.
TL;DR: In this paper, the authors describe the seismic design of Japan Sodium-Cooled Fast Reactor (JSFR), which includes the seismic condition, the seismic isolation system, and the seismic evaluation of the primary components.
Abstract: This paper describes the seismic design of Japan Sodium-Cooled Fast Reactor (JSFR), which includes the seismic condition, the seismic isolation system, and the seismic evaluation of the primary components. Since the design seismic loading is set out severely than ever since The Niigata-ken Chuetsu-oki Earthquake in 2007, an advanced seismic isolation system is aimed to reduce the seismic force loaded on the primary components of JSFR to be less than that of the previous seismic isolation system. The advanced seismic isolation system is developed by optimizing the performance based on the previous seismic isolation system considering the natural frequency of the primary components. The laminated rubber bearings thicker than the previous ones and oil dampers are adopted for the advanced seismic isolation system of SFR. The seismic evaluation of nuclear reactor components applying the advanced seismic isolation system is performed and its feasibility is confirmed.
TL;DR: In this article, the authors observed boiling behavior of water on horizontal and vertical surfaces at pressures from 0.35 to 5 MPa, and the growth curves of the primary bubbles were well described by the t1/2 variation over the whole range of pressures.
Abstract: Experiments were carried out to observe boiling behaviors of water on horizontal and vertical surfaces at pressures from 0.35 to 5 MPa. The growth curves of the primary bubbles are well described by the t1/2 variation over the whole range of pressures. The growth rates of primary bubbles are proportional to the square root of the Jakob number, and agree with the correlation by Labuntsov with the arbitrary constant β = 3. The conventional correlations of bubble growth rates, which are directly proportional to the Jakob number, predict slower growth rates at higher pressures. The coalescence behaviors of the primary bubbles were also measured on the vertical surface at 3.66 MPa. The coalesced bubbles, which were formed by the coalescence of two primary bubbles, grow at rates similar to the rates of the primary bubbles. The nucleation site densities measured on the vertical surface at pressures up to 5 MPa increase in proportion to about the 1.5th power of the pressure under equivalent heat flux conditions. ...
TL;DR: The human errors in soft controls are classified into six types, and the reasons that affect the occurrence of the human errors were also analyzed.
Abstract: The operation environment of main control rooms (MCRs) in modern nuclear power plants (NPPs) has considerably changed over the years Advanced MCRs, which have been designed by adapting digital and computer technologies, have simplerinterfaces using large display panels, computerized displays, soft controls, computerized procedure systems, and so on The actions for the NPP operations are performed using soft controls in advanced MCRs Soft controls have different features from conventional controls Operators need to navigate the screens to find indicators and controls and manipulate controls using a mouse, touch screens, and so on Due to these different interfaces, different human errors should be considered in the human reliability analysis (HRA) for advanced MCRs In this work, human errors that could occur during operation executions using soft controls were analyzed This work classified the human errors in soft controls into six types, and the reasons that affect the occurrence of the human errors
TL;DR: The inorganic-based hydrophobic Pt catalyst named H1P has been developed especially for efficient oxidation over a wide concentration range of hydrogen isotopes at room temperature, even in the presence of saturated water vapor as discussed by the authors.
Abstract: A room-temperature reactor packed with hydrophobic catalysts for the oxidation of hydrogen isotopes released in a nuclear facility will contribute to nuclear safety. The inorganic-based hydrophobic Pt catalyst named H1P has been developed especially for efficient oxidation over a wide concentration range of hydrogen isotopes at room temperature, even in the presence of saturated water vapor. The overall reaction rate constant for hydrogen oxidation with the H1P catalyst in a flow-through system using a tritium tracer was determined as a function of space velocity, hydrogen concentration in carriers, temperature of the catalyst, and water vapor concentration in carriers. The overall reaction rate constant for the H1P catalyst in the range near room temperature was considerably larger than that for the traditionally applied Pt/Al2O3 catalyst. Moreover, the decrease in reaction rate for H1P in the presence of saturated water vapor was slight compared with the reaction rate in the absence of water vapor due t...
TL;DR: In this paper, the heat transfer of supercritical fluids in vertical and horizontal circular tubes has been investigated numerically to understand the thermal-hydraulic behavior of super-critical fluids, and simulations are carried out using different turbulence models and the numerical results are compared with the experimental data to evaluate the accuracy and applicability of those turbulence models.
Abstract: In this study, the heat transfer of supercritical fluids in vertical and horizontal circular tubes has been investigated numerically to understand the thermal-hydraulic behavior of supercritical fluids. The simulations are carried out using different turbulence models and the numerical results are compared with the experimental data to evaluate the accuracy and applicability of those turbulence models. Six turbulence models are used in this study, the LB low-Re k-ϵ model, the LS low-Re k-ϵ model, the RNG k-ϵ model, the realizable k-ϵ model, the standard k-ϵ model, and the Reynolds stress model. The comparison shows that the Reynolds stress model gives better agreement with the experimental data than other turbulence models studied in this work.
TL;DR: In this article, a least square method was used to deduce photoneutron cross sections for Au in the entire energy range of the (γ,n) channel based on a direct neutron counting with quasi-monochromatic γ-rays produced in inverse Compton scattering of laser photons with relativistic electrons.
Abstract: Photoneutron cross section measurements were made for Au in the entire energy range of the (γ,n) channel based on a direct neutron counting with quasi-monochromatic γ-rays produced in inverse Compton scattering of laser photons with relativistic electrons. The data were analyzed by a least-squares method to deduce photoneutron cross sections. The analysis significantly reduced experimental uncertainties compared with those resulting from the photon difference method. The result is compared with the previous data by direct neutron counting with γ-rays produced in positron annihilation in flight and by photoactivation with bremsstrahlung. The present data are in good agreement with the previous data near the neutron threshold, while there remain some discrepancies between the present and the previous data above 10 MeV.
TL;DR: The European Lead Fast Reactor (ELSY) as discussed by the authors is a 600MWe pool-type reactor cooled by pure lead, which demonstrates the possibility of designing a competitive and safe fast critical reactor using simple engineered technical features, whilst fully complying with the Generation IV goals.
Abstract: The European Lead Fast Reactor has been developed in the frame of the European lead system (ELSY) project funded by the Sixth Framework Programme of EURATOM. The project, coordinated by Ansaldo Nucleare, involved a wide consortium of European organizations. The ELSY reference design is a 600MWe pool-type reactor cooled by pure lead. The project demonstrates the possibility of designing a competitive and safe fast critical reactor using simple engineered technical features, whilst fully complying with the Generation IV goals. The paper focuses on the main aspects of the proposed design for the European lead fast reactor highlighting the innovation of this reactor concept and overall objectives. Special attention has been dedicated to safety starting from the first step of the design development taking into account other important aspects, such as the investment protection, the compactness of the primary system as well as sustainability. The main safety features of the proposed innovative decay heat removal...
TL;DR: Extraction of Am(III) was performed at the interface of organic-aqueous two-layer flow in a micro-channel having an asymmetric cross-section as discussed by the authors, where a solution of 3 mol/dm3 nitric acid containing 243Am(III), and octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphineoxide diluted with n-dodecane were introduced into the microchannel as the aqueous phase and organic phase, respectively.
Abstract: Extraction of Am(III) was performed at the interface of organic-aqueous two-layer flow in a micro-channel having an asymmetric cross section. A solution of 3 mol/dm3 nitric acid containing 243Am(III) and octyl(phenyl)-N,N-diisobutylcarbamoylmethylphosphineoxide diluted with n-dodecane were introduced into the microchannel as the aqueous phase and organic phase, respectively. The two phases formed a stable two-layer flow with an interface parallel to the sidewall of the microchannel, and they were separated from each other at the divergence point of the microchannel. The extraction reaction of Am(III) proceeded at the interface of the two phases, and reached the equilibrium state while the two phases passed through the microchannel.
TL;DR: In this paper, the thermoluminescence (TL) responses of commercially available Ge- and Al-doped optical fibers subjected to low-dose electron irradiation were reported.
Abstract: In this paper, we report quantitative features of the thermoluminescence (TL) responses of commercially available Ge- and Al-doped optical fibers subjected to low-dose electron irradiation. The TL response of Ge-doped optical fibers increases monotonically over the investigated dose range 0.02– 0.24 Gy. The Ge-doped optical fibers have a greater response than the Al-doped optical fibers by a factor of at least 15. Results for the doped fibers are also compared with results for TLD-100.
TL;DR: In this paper, the current status of flow-induced vibration evaluation methodology development for primary cooling pipes in the Japan sodium-cooled fast reactor (JSFR), with particular emphasis on recent research and development activities that investigate unsteady elbow pipe flow.
Abstract: This paper describes the current status of flow-induced vibration evaluation methodology development for primary cooling pipes in the Japan sodium-cooled fast reactor (JSFR), with particular emphasis on recent research and development activities that investigate unsteady elbow pipe flow. Experimental efforts have been made using 1/3-scale and 1/10-scale single-elbow test sections for the hot-leg pipe. The 1/10- scale experiment simulating the hot-leg pipe indicated no effect of pipe scale in comparison with the 1/3- scale experiment under inlet-rectified-flow conditions. The next experiment using the 1/3-scale test section was performed to investigate the effect of swirl flow at the inlet. Although the flow separation region was deflected at the downstream from the elbow, the experiment clarified a less significant effect of swirl flow on pressure fluctuation onto the pipe wall. An additional experiment was intended to study the effect of elbow curvature. The experiments with water revealed no clear flow ...
TL;DR: In this article, the amounts of 79Se and 135Cs in a spent nuclear fuel solution were measured by inductively coupled plasma quadrupole mass spectrometry (ICP-QMS) and showed good agreement with those obtained by the ORIGEN2 code using the data library of JENDL-3.3.
Abstract: 79Se and 135Cs are long-lived fission products and are found in high-level radioactive waste (HLW). The estimation of their inventories in HLW is essential for the safety assessment of geological disposal, owing to their mobility in the strata. In this study, the amounts of 79Se and 135Cs in a spent nuclear fuel solution were measured. About 5 g of irradiated UO2 fuel discharged from a commercial Japanese pressurized water reactor (PWR) with a burn-up of 44.9 GWd/t was sampled and dissolved with 50mL of 4M nitric acid in a hot cell for 2 h. After Se and Cs were chemically separated, the amounts of 79Se and 135Cs in the spent nuclear fuel solution were measured by inductively coupled plasma quadrupole mass spectrometry (ICP-QMS). The amounts of 79Se and 135Cs were 5:2 ± 1:5 and 447 ± 40 g/MTU, respectively. The results presented in this study, which are the first postirradiation experimental data in Japan, showed good agreement with those obtained by the ORIGEN2 code using the data library of JENDL-3.3.
TL;DR: In this paper, an empirical model was developed to describe the transient variation in the bed inclination angle during the self-leveling process, which was confirmed through detailed analysis of the effects of experimental parameters such as particle size, particle density, and gas flow rate.
Abstract: Studies on the self-leveling behavior in debris beds are crucial in the assessment of core-disruptive accidents (CDAs) that could occur in sodium-cooled fast reactors (SFRs). To clarify this behavior, a series of experiments have been performed in which nitrogen gas has been percolated uniformly through a particle bed. In these experiments, solid particles and water contained in a rectangular tank simulate respectively fuel debris and coolant. Based on the data obtained, an empirical model was developed to describe the transient variation in the bed inclination angle during the self-leveling process. Good agreement has been obtained between calculated and experimental values. Verification of the model has been confirmed through detailed analysis of the effects of experimental parameters such as particle size, particle density, and gas flow rate. Its applicability to extended conditions was further discussed by performing modeling simulations and comparing results against experimental data obtained from a ...
TL;DR: In this article, the physical properties and irradiation behavior of Np-and Am-bearing MOX were evaluated for the development of advanced fast reactor fuels, and the effect of a few percent Np and Am addition into MOX on those properties was found to be negligibly small.
Abstract: The physical properties and irradiation behavior of Np- and Am-bearing MOX were evaluated for the development of advanced fast reactor fuels. The physical properties, lattice parameter, melting temperature, thermal conductivity, and oxygen potential of the fuels were described as functions of Np content, Am content, and oxygen-to-metal (O/M) ratio, and the effect of a few percent Np and Am addition into MOX on those properties was found to be negligibly small. The irradiation tests of fuel pellets having O/M ratios of 1.98 or 1.96 were carried out at a high linear heat rate of about 430 W/cm for 24 h. The redistributions of actinide element and oxygen were analyzed by using the physical properties for Np- and Am-bearing MOX, and the maximum temperatures of the pellets were estimated. The maximum temperatures of the pellets of O/M = 1:96 and 1.98 were estimated to be 2,701 and 2,510 K, respectively. The analysis result showed that the higher maximum temperature in the low-O/M pellet was caused by the low t...
TL;DR: In this paper, the effective diffusion coefficients of selenium through bentonite/sand mixtures were obtained by the through-diffusion method under reducing conditions, where Se4 2 2 was determined to be SeO3 2 2.
Abstract: Effective diffusion coefficients of selenium through bentonite/sand mixtures were obtained by the through-diffusion method under reducing conditions where selenium was dissolved as HSe− and Se4 2–. Experiments were carried out under variable bentonite content and porewater salinity. Bentonite contents in bentonite/sand mixtures varied from 20 to 70 wt%. The salinities of experimental solutions varied from 0.05 to 0.5 mol dm−3, simulating brackish groundwater and seawater. Effective diffusion coefficients were also obtained under anaerobic conditions where the dominant selenium species was determined to be SeO3 2–. The effective diffusion coefficients of selenium species decreased with increasing bentonite content and with decreasing salinity. This tendency is primarily due to anion exclusion that was caused by the negative surface charge of montmorillonite. A diffusion model for anionic species was developed based on the electric double layer theory and the pore diffusion model, to evaluate the diffusion ...
TL;DR: In this paper, the basic properties of U-Pu-Zr alloys containing minor actinides (MAs) Np, Am, and Cm and rare earths (REs) Y, Nd, Ce, and Gd are characterized by performing ex-reactor experiments.
Abstract: Fast reactor metal fuels containing minor actinides (MAs) Np, Am, and Cm and rare earths (REs) Y, Nd, Ce, and Gd are being developed by the Central Research Institute of Electric Power Industry (CRIEPI) in collaboration with the Institute for Transuranium Elements (ITU) in the METAPHIX project. The basic properties of U-Pu-Zr alloys containing MA (and RE) were characterized by performing ex-reactor experiments. On the basis of the results, test fuel pins including U-Pu-Zr-MA(-RE) alloy ingots in parts of the fuel stack were fabricated and irradiated up to a maximum burnup of ∼10 at% in the Phenix fast reactor (France). Nondestructive postirradiation tests confirmed that no significant damage to the fuel pins occurred. At present, detailed destructive postirradiation examinations are being carried out at ITU.
TL;DR: In this paper, a new release model with pressure effect was proposed and the pressure effect on source term evaluation and effectiveness of accident management measures were investigated, and the validation of the proposed release model tog...
Abstract: In the VEGA program on radionuclide release from irradiated fuel under severe accident conditions, 10 tests in total were performed at JAEA from 1999 to 2004 under inert and steam atmospheres including the highest pressure or temperature conditions. These tests showed the increase in release rate above 2,800K or at the fuel liquefaction and the decrease in release rate under elevated pressure, which was a first observation in the world. The data on low-volatility radionuclide release, release from MOX fuel, effect of fuel oxidation, and eutectic reaction with cladding on release were obtained from the tests. The mechanism of pressure effect on release was examined and a new release model with pressure effect was proposed. In addition, the pressure effect on source term evaluation and effectiveness of accident management measures were investigated. This article summarizes the major outcomes described above that have already been published and newly describes the validation of the proposed release model tog...
TL;DR: In this article, the effect of nuclear data and its uncertainties on the core design of the KALIMER-600 Sodium Fast Reactor was analyzed. But the main components of the reactor were not included in the analysis.
Abstract: In this paper, we present the effect of nuclear data and its uncertainties on the core design of the KALIMER-600 Sodium Fast Reactor. The void reactivity coefficient, k eff, and βeff are calculated with uncertainties due to the nuclear data of the main components of the reactor: 238U, 239;240Pu, 23Na, 56Fe, and 90Zr. Two methods are used: the “Total Monte Carlo” method involving many identical calculations with different sets of randomized nuclear data, and the perturbation method using MCNP. In a second step, the depletion is calculated together with radiotoxicity components after irradiation. By the Total Monte Carlo method, uncertainties due to nuclear data are propagated to the fuel composition and radiotoxicity curves.
TL;DR: In this article, a soft-type neutron-shielding resin has been developed by improving an existing hard-type neutronshielding material using the epoxy-based resin as an additional shielding material.
Abstract: A soft-type neutron-shielding resin has been developed by improving an existing hard-type neutronshielding material using the epoxy-based resin as an additional shielding material. A flexible heat-resistant neutron-shielding material has been developed, which consists of a new polymer-based resin with boron. The neutron shielding performance of the developed flexible heat-resistant resin with the 252Cf neutron source is almost the same as that of polyethylene. The outgases of H, H2, NH4, H2O, CO, O2, C4H10, and CO2 from the developed resin have been measured at high temperature (up to 250°C) by thermal desorption spectroscopy methods. The soft-type resin and the newly developed heat-resistant resin will be applied to prevent the effects of neutron streaming and to control the movement of a vibrated pipe as the seal material around the plumbing in the future fast reactor and innovative fission reactor.