Showing papers in "Journal of Nuclear Science and Technology in 2000"
TL;DR: In this article, a continuous energy Monte Carlo burn-up calculation code MVP-BURN was applied to the burnup benchmark problems for a high conversion LWR lattice and a BWR with burnable poison rods.
Abstract: In order to confirm the reliability of a continuous-energy Monte Carlo burn-up calculation code MVP- BURN, it was applied to the burn-up benchmark problems for a high conversion LWR lattice and a BWR lattice with burnable poison rods. The results of MVP-BURN have shown good agreements with those of a deterministic code SRAC95 for burn-up changes of infinite neutron multiplication factor, conversion ratio, power distribution, and number densities of major fuel nuclides. Serious propagation of statistical errors along burn-up was not observed even in a highly heterogeneous lattice. MVP-BURN was applied to the analysis of a post irradiation experiment for a sample fuel irradiated up to 34.1GWd/t, together with SRAC95 and SWAT. It was confirmed that the effect of statistical errors of MVP-BURN on a burned fuel composition was sufficiently small, and it could give a reference solution for other codes. In the analysis, the results of the three codes with JENDL-3.2 agreed with measured values within an error of ...
124 citations
TL;DR: In this paper, the dissolution rates of amorphous silica at [NaOH]=10-1 mol-dm-3 were analyzed using the model, which assumed that the particle sizes decrease with the progress of dissolution.
Abstract: Cement is an essential materials to construct the subsurface radioactive waste disposal system. However, cementitious materials alter the groundwater pH to highly alkaline condition about 13. To comprehend the effect of such a hyperalkaline condition on the repository surroundings, this study focused on the dissolution rates of amorphous silica at [NaOH]=10-1 mol-dm-3. The used samples were three kinds of pure commercial silica and a natural silica scale which was obtained from inside wall of the hot-water pipe of a geothermal power plant. The observed dissolution rates were interpreted with using the model, which assumed that the particle sizes decrease with the progress of dissolution. Moreover, due to the particle size distribution anticipated in the natural silica scale, this analysis assumed it contained particles with various initial diameters. In the results, (1) all pure silica samples and at least 60wt% of the silica scale showed good agreement of the activation energy of the dissolution in the r...
109 citations
TL;DR: In this article, the main capabilities of an evolution code system, DARWIN, developed at CEA (France), are presented, which is devoted to radioactivity studies in various application fields such as nuclear fuel cycle, dismantling, thermonuclear fusion, accelerator driven system, medecine etc.
Abstract: The aim of this article is to present the main capabilities of an evolution code system, DARWIN, developed at CEA (France). It is devoted to radioactivity studies in various application fields such as nuclear fuel cycle, dismantling, thermonuclear fusion, accelerator driven system, medecine etc. All types of nuclides are dealt with: actinides, fission products, activation products, spallation products. Physical quantities calculated by the code are isotope concentration, isotope mass, activity, radiotoxicity, gamma spectra, beta spectra, alpha spectra, neutron production by spontaneous fission and (α,n) reaction, residual heating, for any cooling times until geological times. Both analytical and numerical schemes are developed in the PEPIN2 depletion module of DARWIN to solve the generalized coupled differential depletion equations. The depletion module PEPIN2 is automatically linked to international evaluations (JEF2, ENDF/B6, EAF97…) both for decay data and cross-sections, and to some transport codes su...
105 citations
TL;DR: In this article, molecular orbital calculations were performed on boric acid and borate monomers and dimers and the final goal was to elucidate boron isotope fractionation observed experimentally.
Abstract: With the final goal to elucidate boron isotope fractionation observed experimentally, molecular orbital calculations were performed on boric acid and borate monomers and dimers. The geometries of B...
60 citations
TL;DR: In this paper, an under-sodium ultrasonic visual inspection technique was developed in order to observe in-vessel structures in a Fast Breeder Reactor (FBR) whose reactor vessel is filled with opaque liquid sodium.
Abstract: We have developed an advanced under-sodium ultrasonic visual inspection technique in order to observe in-vessel structures in a Fast Breeder Reactor (FBR) whose reactor vessel is filled with opaque liquid sodium. The final goal of this work is achievement of resolution equivalent to that of an image obtained by optical fiber scope. The under-sodium ultrasonic visualizing system consists of a matrix-arrayed transducer and a signal- processing device. The matrix-arrayed transducer, in which 36x36 piezoelectric elements are arranged with 5 mm interval and sealed by a thin metal diaphragm, can realize a 3-dimensional image with high resolution. Regarding signal processing, the 3-dimensional image synthetic processing and the cross correlation processing for the purpose of improving S/N ratio of ultrasonic echoes are implemented on a high-speed parallel processor. Under-sodium imaging test was carried out, and it was confirmed that a 3-dimensional image of the blind target, which was prepared without informati...
58 citations
TL;DR: In this paper, the multiplicity and energy of the prompt neutrons emitted from the fission fragments for 239Pu(nth, f) were measured as functions of the fragment mass and total kinetic energy.
Abstract: The multiplicity and energy of the prompt neutrons emitted from the fission fragments for 239Pu(nth, f) were measured as functions of the fragment mass and total kinetic energy. The results were compared with those for 233U(nth, f) and with the predicted values by the multi-channel fission theory with the random neck rupture model. The measured and predicted values of the neutron multiplicity, ⟨V⟩(m*), show the saw-tooth trend and agree with each other. The total neutron multiplicity decreases linearly with increasing total kinetic energy resulting in —d⟨TKE⟩/d⟨Vtot⟩=16.5±0.4MeV/neutron. The slope of the neutron multiplicity vs. the total kinetic energy, —d⟨V⟩/d⟨TKE⟩, was plotted against the fragment mass. Its shape agrees with that for 233U(nth, f). The average neutron emission energy, ⟨η⟩(m*), follows a bell shape about the symmetric fission accompanying higher values for very asymmetric fissions and agrees with that for 233U (nth, f). The total excitation energy (TXE)(m*) was determined by two manners:...
53 citations
TL;DR: A thermal hydraulic design and performance analysis computer code for a once-through steam generator using helically coiled tubes, ONCESG, is developed and it is demonstrated that the ONC ESG code can be utilized for diverse purposes, such as, sensitivity analyses and optimum thermal design of aOnce-Through steam generator.
Abstract: Development of the conceptual design of a 300 MWt integral reactor, SMART (System-integrated Modular Advanced ReacTor), for utilization in nuclear cogeneration plants has been completed at the Korea Atomic Energy Research Institute (KAERI). The major primary components of the SMART such as modular helical steam generators, main circulation pumps and a self regulating pressurizer are integrated into a reactor vessel. It is a common practice to employ a once-through steam generator in integral reactor designs because of its advantages in compactness and simplicity of the flow path arrangements. In this study, a thermal hydraulic design and performance analysis computer code for a once-through steam generator using helically coiled tubes, ONCESG, is developed. To benchmark the developed physical models and computer code, once-through steam generators developed by other designers are simulated and ONCESG calculated results are compared with the design data. The overall characteristics of heat transfer area, p...
49 citations
TL;DR: A series of single-phase natural circulation experiments in a simulated marine reactor mounted on a rolling bed was performed and the average Nusselt number in the core was evaluated in order to investigate effects of the rolling motion on the heat transfer as mentioned in this paper.
Abstract: A series of single-phase natural circulation experiments in a simulated marine reactor mounted on a rolling bed was performed and the average Nusselt number in the core was evaluated in order to investigate effects of the rolling motion on the heat transfer in the core. Heat transfer with an upright attitude is well correlated with the Rayleigh number and is slightly lower than El-Genk's correlation. Heat transfer in the core is not affected by the inclination angle because the inclination of the present experiment is not large enough to cause any remarkable changes in the flow pattern of the core. Heat transfer in the core is enhanced by the rolling motion which is thought to cause internal flow in the core. Heat transfer during the rolling motion is correlated with the Richardson number for rolling motion, R iR , and is classified into three regimes: (1) region A (0.05
44 citations
TL;DR: A characteristics transport theory code, CHAPLET as discussed by the authors, has been developed for the purpose of making it practical to perform a whole LWR core calculation with the same level of calculational model and accuracy as that of an ordinary single assembly calculation.
Abstract: A characteristics transport theory code, CHAPLET, has been developed for the purpose of making it practical to perform a whole LWR core calculation with the same level of calculational model and accuracy as that of an ordinary single assembly calculation. The characteristics routine employs the CACTUS algorithm for drawing ray tracing lines, which assists the two key features of the flux solution in the CHAPLET code. One is the direct neutron path linking (DNPL) technique which strictly connects angular fluxes at each assembly interface in the flux solution separated between assemblies. Another is to reduce the required memory storage by sharing the data related to ray tracing among assemblies with the same configuration. For faster computation, the coarse mesh rebalance (CMR) method and the Aitken method were incorporated in the code and the combined use of both methods showed the most promising acceleration performance among the trials. In addition, the parallelization of the flux solution was attempted...
41 citations
TL;DR: In this paper, a new neptunium redox battery is proposed for electric power storage, which can be used as an active material of the redox flow battery for non-aqueous solvents.
Abstract: The electrochemical properties of U, Np, Pu and Am were discussed from the viewpoint of cell active materials. From the thermodynamic properties and the kinetics of electrode reactions, it is found that neptunium in the aqueous system can be utilized as an active material of the redox flow battery for the electric power storage. A new neptunium redox battery is proposed in the present article: the galvanic cell is expressed by The neptunium battery is expected to have more excellent charge and discharge performance than the current vanadium battery, whereas the thermodynamic one of the former is comparable to the latter. For the development of a uranium redox flow battery, the application of the redox reactions in the non-aqueous solvents is essential.
36 citations
TL;DR: In this paper, the authors describe research and development (R&D) and design specification of the HTR-10 SEF, summarizes the fabrication technology and quality control mastered in R&D for HTR10 SFE.
Abstract: The fuel elements for Chinese 10 MW High Temperature Gas-cooled Reactor (HTR-10) are spherical all- ceramic fuel elements (SFE). TRISO (Tri-isotropic) coated fuel particles (CP) are uniformly dispersed in the graphite matrix of the fuel element. All radiological fission products are almost completely retained inside the SiC layer of the intact CP. The fabrication of SFE includes UO2 kernel preparation by sol-gel method, pyrolytic carbon (PyC) and SiC coating on the UO2 kernels by Chemical Vapor Deposition, manufacture of SFE by the quasi-isostatic pressing and the inspection of over 30 kinds of properties. This paper describes research and development (R & D) and design specification of the HTR-10 SEF, summarizes the fabrication technology and quality control mastered in R & D for HTR-10 SFE.
TL;DR: In this paper, the influence of precipitated hydride on the fracture behavior of Zircaloy cladding tubes was estimated by finite element method (FEM) analysis.
Abstract: In order to clarify the influence of precipitated hydride on the fracture behavior of Zircaloy cladding tubes, the stress-strain distribution of the cladding was estimated by finite element method (FEM) analysis. The mechanical properties of α-phase of zirconium and zirconium hydride required for the analysis were examined by means of an ultrasonic pulse-echo method and a tensile test. It was found from the analysis that the non-hydrided cladding has the highest equivalent plastic strain at the inner surface of the cladding, suggesting that the fracture initiated at the inner surface of the cladding. Since the hydride accumulated layer located in the outer surface of the hydrided cladding fails at a lower internal pressure, the crack appears to initiate at the outer surface of the cladding. The fracture behavior estimated from the stress states of the hydrided cladding was in good agreement with the experimental results obtained by pulse irradiation tests of the Nuclear Safety Research Reactor (NSRR) and ...
TL;DR: In this paper, the authors synthesized several 2,6-bis(5,6dialkyl)-1,2,4-triazine-3-yl)pyridine (R-BTP) ligands with different alkyl groups and prepared the novel silica-based extraction resins by impregnating the ligands into the SiO 2 -P support.
Abstract: In this work, we have synthesized several 2,6-bis(5,6-dialkyl-1,2,4-triazine-3-yl)pyridine (R-BTP) ligands with different alkyl groups (C=1∼3) and prepared the novel silica-based extraction resins by impregnating the ligands into the SiO 2 -P support. Furthermore, the adsorption performance of Am(III) and Ln(III) from nitrate acidic solution onto these novel silica-based nitrogen donor extraction resins was investigated experimentally.
TL;DR: A pyrometallurgical partitioning process for transuranic elements (TRUs) from high-level liquid waste (HLLW) has been studied to develop an effective and safety method of nuclear waste disposal as discussed by the authors.
Abstract: A pyrometallurgical partitioning process for transuranic elements (TRUs) from high-level liquid waste (HLLW) has been studied to develop an effective and safety method of nuclear waste disposal. A salt and metal as solvents, Li as reductant and chlorine gas can be recycled in the process. The process is expected to generate less secondary radioactive waste because of no irradiation damage of the solvents, and to require a relatively compact installation because of larger critical mass in comparison with a conventional aqueous method. In this study, the material balance of the solutes and the volume of the solvents were estimated in the pyrometallurgical partitioning process that was constructed on the basis of our previous study. The results show that the volume of the salt as solvent is about 200 l and that of Cd, Bi and Pb as metal solvent is 54, 71 and 140 l, respectively, in the case that HLLW from PUREX reprocessing of 1t LWR spent fuel is treated in the process. The results also show that the amount...
TL;DR: In this paper, a new approach for estimating the fission source intensity ratio in an array is proposed by obtaining the eigenvector of a coupling coefficient matrix, which also gives the uncertainty of the ratio as well as the ratio, which is available for evaluating the accuracy of the obtained ratio.
Abstract: Anomalous fission source convergence in a Monte Carlo criticality calculation for a weakly coupled array of two fissile material units are demonstrated. Introducing coupling coefficients among array units, it is quantitatively explained that this anomaly is caused by an insufficient restoring force to the true distribution and its large statistical uncertainty, especially, in a symmetric system. A new approach for estimating the fission source intensity ratio in an array is proposed by obtaining the eigenvector of a coupling coefficient matrix. This method also gives the uncertainty of the ratio as well as the ratio, which is available for evaluating the accuracy of the obtained ratio. The correlation between a calculated k eff and the fission source intensity ratio is formulated. It is illustrated theoretically and empirically that there is no significant correlation in a symmetric two-unit array system. In general, care should be taken that a calculated k eg may be biased by an incorrect fission source ...
TL;DR: The tensile strength of ferritic-martensitic HCr-0.5Mo-2W, Nb, V stainless steel (PNC-FMS) was evaluated for the effects of thermal aging, sodium exposure, and neutron irradiation.
Abstract: The tensile strength of ferritic-martensitic HCr-0.5Mo-2W, Nb, V stainless steel (PNC-FMS), which had been developed for core component applications in LMFBR by Japan Nuclear Cycle Development Institute, was evaluated for the effects of thermal aging, sodium exposure, and neutron irradiation. The tensile strength of thermal aged specimens (~1,023K, ~12,000h) decreased at aging conditions above the initial tempering parameter, and the aging effect was considerably enhanced for the wrapper tubes tempered at lower temperatures. The tensile strength of sodium exposed specimens (~973K, ~10,000h) decreased more than aged specimens due to decarburization, and the effect of decarburization was greater in thin wall cladding tubes. Evaluation of the contribution of both thermal aging and decarburization effects on the tensile strength of cladding tubes irradiated in JOYO (~1,013K, ~6,030h, ~29dpa) suggested that the radiation showed smaller effect on tensile properties than thermal aging and decarburization. By usi...
TL;DR: In this paper, a least square method was applied to selected absolute and relative measurements on the fission cross sections of 233U, 235U, 238U, 239Pu, 240Pu and 241Pu was carried out for the evaluated nuclear data library JENDL-3.3.
Abstract: A simultaneous evaluation of the fission cross sections of 233U, 235U, 238U, 239Pu, 240Pu and 241Pu was carried out for the evaluated nuclear data library JENDL-3.3. A least-squares method was applied to selected absolute and relative measurements on the fission cross sections. Covariance matrices of the experimental data were constructed from the uncertainty information reported in the references. The fission cross sections obtained were compared with the JENDL-3.2 and ENDF/B-VI evaluations. It was found from the comparison that the present results are not so different from those in JENDL-3.2, except for the fission cross sections of 233U and the cross sections above 15MeV, and give smaller X2 value than the JENDL-3.2 cross sections. The averaged fission cross sections of 233U, 238U, and 239Pu relative to that of 235U were calculated for a neutron spectrum produced by 9Be (d,xn) reaction. It was confirmed that the calculated cross-section ratios are in good agreement with the experimental data.
TL;DR: The MARS Monte Carlo code as discussed by the authors was developed for simulation of hadronic and electromagnetic cascades in shielding, accelerator and detector components in the energy range from a fraction of an electron volt up to 100 TeV.
Abstract: Recent developments of the MARS Monte Carlo code system for simulation of hadronic and electromagnetic cascades in shielding, accelerator and detector components in the energy range from a fraction of an electron volt up to 100 TeV are described. The physical model of hadron and lepton interactions with nuclei and atoms has undergone substantial improvements. These include a new nuclear cross section library, a model for soft prior production, a cascade-exciton model, a dual parton model, deuteron-nucleus and neutrino-nucleus interaction models, a detailed description of negative hadron and muon absorption, and a unified treatment of muon and charged hadron electro-magnetic interactions with matter. New algorithms have been implemented into the code and benchmarked against experimental data. A new Graphical-User Interface has been developed. The code capabilities to simulate cascades and generate a variety of results in complex systems have been enhanced. The MARS system includes links to the MCNP code for neutron and photon transport below 20 MeV, to the ANSYS code for thermal and stress analyses and to the STRUCT code for multi-turn particle tracking in large synchrotrons and collider rings. Results of recent benchmarking of the MARS code are presented. Examples of non-trivial code applications are given for the Fermilab Booster and Main Injector, for a 1.5 MW target station and a muon storage ring.
TL;DR: In this article, the authors have conducted diffusion experiments at 923 K using two couples: U-13 at%Pu-22 at%Zr/Fe and U-22at%Pus- 22 at% Zr /Fe, and examined the influence of the Pu content in the fuel alloy on the phases formed in the reaction zones.
Abstract: In metallic U-Pu-Zr fuel, metallurgical reactions occur between the fuel slug and the cladding, and a phase of which the melting point is relatively low is formed in the reaction zone. If a liquid phase is formed, it can degrade cladding integrity. The potential for liquid phase formation near the cladding, therefore, should be excluded during normal reactor operation. In order to clarify the mechanism of liquefaction, the authors have conducted diffusion experiments at 923 K using two couples: U-13 at%Pu-22 at%Zr/Fe and U-22at%Pu- 22at%Zr/Fe, and examined the influence of the Pu content in the fuel alloy on the phases formed in the reaction zones. The liquid phase has been observed in the U-22 at%Pu-22 at%Zr/Fe couple. An assessment of the diffusion paths for these couples has indicated that the Pu content in the (U, Pu)6Fe-type phase in the reaction zone is a crucial factor in determining the conditions that lead to liquefaction. The Pu content in the (U, Pu)6Fe-type phase increases with that in the ini...
TL;DR: In this paper, a 1/4T compact tension specimen was used for measurement of crack growth rates (CGRs) of sensitized type 304 stainless steel in high temperature and high purity water Crack length was monitored by a reversing direct current potential drop method.
Abstract: The stress corrosion cracking (SCC) of structural materials used in boiling water reactors has been studied at relatively low hydrogen peroxide (H 2 O 2 ) concentrations, around 10ppb, which was assumed to be representative of the corrosion environment formed in hydrogen water chemistry (HWC) The 1/4T compact tension specimen was used for measurement of crack growth rates (CGRs) of sensitized type 304 stainless steel in high temperature and high purity water Crack length was monitored by a reversing direct current potential drop method Since H 2 O 2 is easily decomposed thermally, a polytetrafluoroethylene-lined autoclave was used to minimize its decomposition on the autoclave surface The CGR in the H 2 O 2 environment differed from that in the O 2 environment even though the electrochemical corrosion potential (ECP) for both conditions was the same The data implied that the ECP could not be used as a common environmental deterministic parameter for SCC behavior at higher potentials for different oxidant conditions The corrosion current density was found to play an important role as an environmental index for SCC, which was given as just the current density at the ECP at a specific oxidant concentration The CGRs were found to be written as CGR = (38±06)×10 -3 i cor +(15±16) × 10 -8 mm/s using the calculated corrosion current density i cor below 10 -4 Acm -2
TL;DR: In this article, a substance for solidifying waste containing 129I is sought that effectively sorbs iodine to inhibit its release from repository into the environment, and three candidate media (commercial alumina cemen...
Abstract: A substance for solidifying waste containing 129I is sought that effectively sorbs iodine to inhibit its release from repository into the environment. Three candidate media—commercial alumina cemen...
TL;DR: In this article, the authors proposed a methodology of estimation for dose rates after shutdown in a tokamak fusion experimental reactor, which is a very complex geometry, and it is important to have reliable estimation of the dose rate levels after reactor shutdown for realising hands-on maintenance around the torus.
Abstract: In the shielding design of the ITER machine, which is a tokamak fusion experimental reactor and has a very complex geometry, it is important to have a reliable estimation of the dose rate levels after reactor shutdown for realising hands-on maintenance around the torus. The ITER project position is that dose rates inside the cryostat be kept low enough to allow for human access shortly after shutdown for limited periods to provide rescue and/or maintenance activities.The methodology of estimation for dose rates after shutdown in such a complex geometry machine is discussed. The Monte Carlo method is preferable to conduct neutron transport calculations in the ITER geometry. The Conversion Factor method, which was used for dose rate estimation in the 1998 ITER shielding design, is described with an example of dose rate estimation around penetrations in the vacuum vessel. A Full Monte Carlo method is proposed showing the possibility of eliminating uncertainty accompanied with conversion factors from neutron ...
TL;DR: In this article, a new integrated approach is presented, based on an extension of methods used at proton accelerators and on the unique capability of the FLUKA Monte Carlo code to handle the whole joint electromagnetic and hadronic cascade.
Abstract: Radioactive nuclides are produced at high-energy electron accelerators by different kinds of particle interactions with accelerator components and shielding structures. Radioactivity can also be induced in air, cooling fluids, soil and groundwater. The physical reactions involved include spallations due to the hadronic component of electromagnetic showers, photonuclear reactions by intermediate energy photons and low-energy neutron capture. Although the amount of induced radioactivity is less important than that of proton accelerators by about two orders of magnitude, reliable methods to predict induced radioactivity distributions are essential in order to assess the environmental impact of a facility and to plan its decommissioning. Conventional techniques used so far are reviewed, and a new integrated approach is presented, based on an extension of methods used at proton accelerators and on the unique capability of the FLUKA Monte Carlo code to handle the whole joint electromagnetic and hadronic cascade...
TL;DR: In this paper, thermal migration of hydrogen under temperature gradient has been estimated according to traditional diffusion theory for δ-ZrHx+U fuel, and it has been shown that hydrogen profiles at steady are stable.
Abstract: To evaluate irradiation behavior of δ-ZrHx+U fuel, thermal migration of hydrogen under temperature gradient has been estimated according to traditional diffusion theory Hydrogen profiles at steady
TL;DR: In this article, LiClO4 was inserted into zinc from a mixed solution of ethylene carbonate and methyl ethyl carbonate, and the lithium isotope effect accompanying the insertion was investigated.
Abstract: Lithium was electrochemically inserted into zinc from the mixed solution of ethylene carbonate and methyl ethyl carbonate containing 1M LiClO4, and the lithium isotope effect accompanying the insertion was investigated. The x value of LixZn after the electrolysis ranged from 0.013 to 0.139. The single-stage separation factor, S, ranged from 1.005 to 1.023 with the lighter isotope, 6Li, being preferentially fractionated into the zinc phase. The present results on lithium isotope effect qualitatively agreed with calculated results based on the molecular orbital theory.
TL;DR: In this paper, batch and column experiments were carried out to investigate the ion exchange characteristics of Pd(II) and Rh(III), including the effects of the ionic group of ion exchangers, solution temperature, and the concentration of nitric acid by various ion exchanger such as IRN 78 and Dowex 1x8.
Abstract: Radioactive high-level liquid waste contains significant quantities of platinum group metals (PGM) such as palladium (Pd(II)), rhodium (Rh(III)) and ruthenium (Ru(III)). In this study, batch and column experiments were carried out to investigate the ion exchange characteristics of Pd(II) and Rh(III), including the effects of the ionic group of ion exchangers, solution temperature, and the concentration of nitric acid by various ion exchangers such as IRN 78 and Dowex 1x8. The elution characteristics of Pd(II) and Rh(III) by various eluents were also examined. The anion exchangers such as Dowex 1x8 with the ionic functional group of quaternary methyl ammonium had a higher adsorption capacity than anion exchangers such as IRN 78 with the amine group for the adsorption of Pd(II) from nitric acid solution. The cation exchanger, Dowex 50W, with the sulfonic group shows significantly strong Rh adsorption at a diluted nitric acid solution. The optimal nitric acid concentration by anion exchange resins was shown ...
KEK1
TL;DR: In this paper, the joint project prepared by the Joint Project Team of JAERI and KEK to construct accelerators and research facilities necessary both for the Neutron Science Project (NSP) and the Japan Hadron Facility Project (JHF) at the site of JERI Tokai Establishment is described.
Abstract: Japan Atomic Energy Research Institute (JAERI) and the High Energy Accelerator Organization (KEK) are promoting the joint project integrating both the Neutron Science Project (NSP) of JAERI and the Japan Hadron Facility Project (JHF) of KEK for comprehensive studies on basic science and technology using high-intensity proton accelerator.This paper describes the joint project prepared by the Joint Project Team of JAERI and KEK to construct accelerators and research facilities necessary both for the NSP and the JHF at the site of JAERI Tokai Establishment.
TL;DR: In this article, the authors have studied on the calculation method to be used in the estimation of spent fuel compositions and adopted the real irradiation calculation in which axial burnup and moderator distribution are considered in the burnup calculation.
Abstract: The compositions and quantities of minor actinide (MA) and fission product (FP) in spent fuels will be diversified with the use of high discharged burnup fuels and MOX fuels in LWRs which will be a main part of power reactors in future. In order to investigate above diversities, we have studied on the calculation method to be used in the estimation of spent fuel compositions and adopted the real irradiation calculation in which axial burnup and moderator distribution are considered in the burnup calculation. On the basis of the calculations, compositions and burnup quantities of various LWR spent fuels (reactor type: PWR and BWR, discharged burnup: 33, 45 and 60 GWd/tHM, fuel type: U02 and MOX) are apparently estimated among various forms of fuels. As an example, it is shown that there are considerable discrepancy in MA burnup between PWR and BWR spent fuels.
TL;DR: In this article, the authors describe flow regimes, interfacial area modeling and physics models such as the momentum exchange and phase change models, and two separate effect analyses are presented as verification of MCRA's models.
Abstract: IMPACT is a complex software system under development at the Nuclear Power Engineering Corporation, that includes the severe accident analysis code (SAMPSON). SAMPSON is an integration of twelve modules and will be capable of simulating hypothesized severe accidents in a nuclear power plant in the final phase of the IMPACT project. As one of these modules, the Molten Core Relocation Analysis (MCRA) module simulates the relocation behavior of a molten core during a severe accident. MCRA adopts a multi-phase, multi- component, multi-velocity field model to simulate severe accident phenomena mechanistically. Herein, we describe flow regimes, interfacial area modeling and physics models such as the momentum exchange and phase change models. Two separate effect analyses are presented as verification of MCRA's models following the model descriptions. First, the fluid dynamics of the multi-velocity field model was verified in the calculation of nitrogen gas bubbling through water in Leung's experiment. Second, d...
TL;DR: In this paper, structural and kinetic studies of U(VI) complex with benzamidoxime (Hba) as ligand in CD3COCD3 have been studied by means of 1H and 13 C NMR.
Abstract: The structural and kinetic studies of U(VI) complex with benzamidoxime (Hba) as ligand in CD3COCD3 have been studied by means of 1H and 13 C NMR. The Hba molecule was found to coordinate to UO2+ 2 in the form of anionic benzamidoximate (ba), and the number of ba coordinated to UO2+ 2 was determined to be 3 by analyzing the chemical shift of 13C NMR signal for Hba in the presence of UO2+ 2. The exchange rate constants (kex) of ba in [UO2(ba)3]- were determined by the NMR line-broadening method. The kinetic parameters were obtained as follows: kex (25°C)=1.3× 103 s-1, ΔH≠ =35.8±3.5kJ·mol-1, and ΔS≠=-65±13.7 J·K-1·mol-1. The UV-visible absorption spectra of solutions containing UO2+ 2 and Hba were also measured. The molar extinction coefficient of the complex was found to be extremely large compared with those of UO2(L)2+ 5 (L=unidentate oxygen donor ligands) complexes. This is due to the strong electron withdrawing of UO2+ 2 from Hba and suggests that an interaction between UO2+ 2 and Hba is very strong. Su...