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JAEA Reports

Theoretical background and user's manual for the computer code on groundwater flow and radionuclide transport calculation in porous rock


JNC-TN8400 2001-027, 131 Pages, 2001/11


In order to document a basic manual about input data, output data, execution of computer code on groundwater flow and radionuclide transport calculation in heterogeneous porous rock, we investigated the theoretical background about geostastical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport which calculates water flow in three dimension, the path of moving radionuclide, and one dimensional radionuclide migration. In this report, based on above investigation we describe the geostastical background about simulating heterogeneous permeability field. And we describe construction of files, input and output data, a example of calculating of the programs which simulates heterogeneous permeability field, and calculates groundwater flow and radionuclide transport. Therefore, we can document a manual by investigating the theoretical background about geostastical computer codes and the user's manual for the computer code on groundwater flow and radionuclide transport calculation. And we can model heterogeneous porous rock and analyze groundwater flow and radionuclide transport by utilizing the information from this report.

JAEA Reports

Sodium combustion analysis for the secondary heat transport system of prototype fast breeder reactor MONJU

; Ohno, Shuji;

JNC-TN2400 2000-006, 56 Pages, 2000/12


Sodium combustion analyses were performed using ASSCOPS version 2.1 in order to obtain background data for evaluating the validity of the mitigation system against secondary sodium leak of MONJU. The calculated results are summarized as follows. (1)Peak atmospheric pressure $$sim$$ 4.3 kPa[gage] (2)Peak floor liner temperature $$sim$$ 870$$^{circ}$$C, Maximum thinning of liner $$sim$$2.6mm (3)Peak hydrogen concentration <2% (4)Peak floor liner temperature in the spilt sodium storage eell $$sim$$ 400$$^{circ}$$C , Peak floor concrete temperature in the spilt sodium storage cell $$sim$$ 140$$^{circ}$$C.

JAEA Reports

Development and validation of Multi-DimensionaI sodium combustion analysis code AQUA-SF

Takata, Takashi; Yamaguchi, Akira

JNC-TN9400 2000-065, 152 Pages, 2000/06


ln the liquid metal fast reactor (LMFR) using liquid sodium as a coolant, it is important to evaluate the effect of the sodium combustion on the structure, etc. Most of the previous analytical works are based on a zone model, in which the principal variables are treated as volume-average quantities. Therefore spatial distribution of gas and structure temperatures, chemical species concentration are neglected. Therefore, a multi-dimensional sodium combustion analysis code AQUA-SF (Advanced simulation using Quadratic Upstream differencing Algorithm - Sodium Fire version) has been developed for the purpose of analyzing the sodium combustion phenomenon considering the multi-dimensional effect. This code is based on a multi-dimensional thermal hydraulics code AQUA that employs SIMPLEST-ANL method. Sodium combustion models are coupled with AQUA; one is a liquid droplet model for spray combustion, and the other is a flame sheet model for pool combustion. A gas radiation model is added for radiation heat transfer. Some other models necessary for the sodium combustion analysis, such as a chemical species transfer, a compressibility, are also added. ln AQUA-SF code, bounded QUICK method in space scheme and bounded three-point implicit method in time scheme are implemented. Verification analyses of sodium combustion tests shown in the following have been carried out. (1)pool combustion test (RUN-D1) (2)spray combustion test (RUN-E1) (3)sodium leakage combustion test (Sodium Fire Test-II) (4)smaII-scale leakage combustion test (RUN,F7-1) ln each verification analysis, good agreements are obtained and the validity of AQUA-SF code is confirmed.

JAEA Reports

Development of an on-site plant analyzer (1); Development of a GUI for building plant models for analyzes and retrieval of real-time plant data

JNC-TN4400 2000-002, 33 Pages, 2000/06


An on-site plant analyzer can provide analysis support in evaluating plant dynamic characteristics when unplanned events occur in a nuclear power station. The plant analyzer contains a plant-dynamics analysis code, which efficiently and quickly analyzes the plant dynamic characteristics. Elements being developed for the on-site plant analyzer include utilities to build plant models for performing analyses and to retrieve plant operation data. The addition of these elements to the analysis code supports the plant-dynamics analysis works in MONJU, in particular, to assist the analyses of start up tests. The system contains the FBR plant-dynamics analysis code "Super-COPD", which is based on the "COPD" code that was used in the safety licensing of MONJU. One feature of the system is that all operations, e.g., assembling plant models for analysis, are prepared using a GUI (Graphical user Interface). In addition to this feature, the system is able to retrieve directly on- and off-line plant data from MIDAS, the Monju Integrated Data Acquisition System. These plant data are used to supply time-dependent boundary conditions for the plant analysis models. For this report, two case studies were performed. First, the analysis result of a turbine trip test at 40% power operation using the full plant model is described. For the second, performance of the IHX model was evaluated using retrieved plant data for boundary conditions. With the development of this system, improvement in the efficiency of analyses of MONJU start-up tests is expected.

JAEA Reports

Expansion of material balance analysis function on nuclear fuel cycle

Ohtaki, Akira; ; ; *; *;

JNC-TN9410 2000-006, 74 Pages, 2000/04


To evaluate materials balance in nuclear fuel cycle quickly and quantitatively the fuel cycle requirement code "FAMILY" was improved. And an accumulated TRU&LLFP quantity analysis code was developed. The contents are as follows: (1)A calculation ability of minor actinide production and expenditure was added to the "FAMILY" code. (2)An output program for the "FAMILY" calculation results was developed. (3)A simple version of "FAMILY" code was developed. (4)An analysis code for accumulated TRU&LLFP quantity in nuclear fuel cycle was developed.

JAEA Reports

Desgin study of pyrochemical process operation by using virtual engineering models

; Tozawa, Katsuhiro; ; ; *

JNC-TN9400 2000-053, 99 Pages, 2000/04


This report describes accomplishment of simulations of Pyrochemical Process Operation by using virtual engineering models. The pyrochemical process using molten salt electrorefining would introduce new technologies for new fuels of particle oxide, particle nitride and metallic fuels. This system is a batch treatment system of reprocessing and re-fabrication, which transports products of solid form from a process to next process. As a result, this system needs automated transport system for process operations by robotics. ln this study, a simulation code system has been prepared, which provides virtual engineering environment to evaluate the pyrochemical process operation of a batch treatment system using handling robots. And the simulation study has been conducted to evaluate the required system functions, which are the function of handling robots, the interactions between robot and process equipment, and the time schedule of process, in the automated transport system by robotics. As a result of simulation of the process operation, which we have designed, the automated transport system by robotics of the pyrochemical process is realistic. And the issues for the system development have been pointed out.

JAEA Reports

Development of system analysis code for pyrochemical process using molten salt electrorifining

Tozawa, Katsuhiro; ;

JNC-TN9400 2000-052, 110 Pages, 2000/04


This report describes accomplishment of development of a cathode processor calculation code to simulate the mass and heat transfer phenomena with the distillation process and development of an analytical model for cooling behavior of the pyrochemical process cell on personal computers. The pyrochemical process using molten salt electrorefining would introduce new technologies for new fuels of particle oxide, particle nitride and metallic fuels. The cathode processor calculation code with distillation process was developed. A code validation calculation has been conducted on the basis of the benchmark problem for natural convection in a square cavity. Results by using the present code agreed well for the velocity-temperature fields, the maximum velocity and its location with the benchmark solution published in a paper. The functions have been added to advance the reality in simulation and to increase the efficiency in utilization. The test run has been conducted using the code with the above modification for an axisymmetric enclosed vessel simulating a cathode processor, and the capability of the distillation process simulation with the code has been confirmed. An analytical model for cooling behavior of the pyrochemical process cell was developed. The analytical model was selected by comparing benchmark analysis with detailed analysis on engineering workstation. Flow and temperature distributions were confirmed by the result of steady state analysis. In the result of transient cooling analysis, an initial transient peak of temperature occurred at balanced heat condition in the steady-state analysis. Final gas temperature distribution was dependent on gas circulation flow in transient condition. Then there were different final gas temperature distributions on the basis of the result of steady-state analysis. This phenomenon has a potential for it's own metastable condition. Therefore it was necessary to design gas cooling flow pattern without ...

JAEA Reports


Yamanaka, Shinsuke*; Abe, Kazuyuki

JNC-TY9400 2000-004, 78 Pages, 2000/03


no abstracts in English

JAEA Reports

Analysis of weld residual stresses by FINAS (1)


JNC-TN9400 2000-047, 114 Pages, 2000/03


Prediction of weld residual stresses by a general finite element code is beneficial to the improvement of the accuracy of integrity assessment and residual life assessment of FBR plants. This reports develops an evaluation method of weld residual stresses using FINAS. Firstly, we suggested a basic procedure derived from parametric analyses with a simple weld joint model. The procedure can be summarized as follows: (1)For heat conduction analysis, prepare different models corresponding to the number of layers to be modeled. Hand over the analytical results to the following model. (2)Use multi-linear stress-strain curves for modeling the stress-strain response of base metal and weld metal. Use the isotropic hardening rule. (3)When metals are melt, use a user-subroutine to keep stresses from arising. (4)Put the thermal expansion coefficient as zero when heat is being input. Then, using the above procedure and TIG welding, we predicted the weld residual stresses of plate and tube. The results agreed well with the other reports, showing the suggested procedure was reasonable.

JAEA Reports

Modification of the evaluation model for Pu redistribution phenomena

; *;

JNC-TN9400 2000-045, 64 Pages, 2000/03


During the irradiation, the Pu redistribution phenomena would occur in the FBR MOX fuel pellets. The phenomena would considerably affect on the thermal properties of the fuels, therefore, it is need to establish the evaluation method for Pu redistribution phenomena. ln JNC, the efforts for development of the evaluation model for the phenomena had been continued and the simple evaluation model was constructed in 1992. In this work, the modification of the simple model developed in JNC has been done and the following results were obtained. (1)Based on the recent data of the MOX fuel irradiation tests, the evaluation model for Pu redistribution phenomena constructed in l992 is modified. And the model is included into the fuel performance analysis code "CEDAR". (2)To calibrate the modified CEDAR code, it is confirmed that the uncertainty in the Pu concentration evaluation for the center of the fuel pellet at EOL is about $$pm$$3wt.%. (3)Based on the results of the evaluations using the modified CEDAR code, it is found that, in the early stage of the irradiation, the Pu redistribution is controlled by the vapor transportation mechanism via pores, and after that, the Pu redistribution is kept in progress due to the thermal diffusion mechanism with the change of the Pu concentration due to the degradation of U and Pu by fissions. And it is also found that the O/M ratio dependence of the U-Pu inter diffusion coefficients would affect on the Pu redistribution mechanisms, in especial, in the early stage of the irradiation.

JAEA Reports

Preparation of next generation set of group cross sections; A Task report to the Japan Nuclear Cycle Development Institute)


JNC-TJ9400 2000-005, 182 Pages, 2000/03


The SLAROM code, performing fast reactor cell calculation based on a deterministic methodology, has been revised by adding the universal module PEACO of generating Ultra-fine group neutron spectra. The revised SLAROM, then, was utilized for evaluating reaction rate distributions in ZPPR-13A simulated by a 2-dim RZ homogeneous model, although actually ZPPR-13A composed of radial heterogereous cells. The reaction rate distributions of ZPPR-13A were also calculated by the code MVP, that is a continuous energy Monte Carlo calculation code based on a probabilistic methodology. By coparing both results, it was concluded that the module PEACO has excellent capability for evaluating highly accurate effective cross sections. Also it was proved that the use of a new fine group cross section library set (next generation set), reflecting behavior of cross sections of structural materials, such as Fe and O, in the fast neutron energy region, is indispensable for attaining a better agreement within 1% between both calculation methods. Also, for production of a next generation set of group cross sections, the code NJOY97.V107 was added to the group cross section production system and both front and end processing parts were prepared. This system was utilized to produce the new 70 group JFS-3 library using the evaluated nuclear data library JENDL-3.2. Furthermore, to confirm the capability of this new group cross section production system, the above new JFS-3 library was applied to core performance analysis of ZPPR-9 core with a 2-dim RZ homogeneous model and analysis of heterogeneous cells of ZPPR-9 core by using the deterministic method. Also the analysis using the code MVP was performed. Bycoaparison of both results the following conclusion has been derived; the deterministic method, with the PEACO module for resonance cross sections, contributes to improve accuracy of predicting reaction rate distributions and Na void reactivity in fast reactor cores. And it ...

JAEA Reports

Development of the evaluation methodology for earthquake resistance of the engineered barrier system (III)

Mori, Koji*; Neyama, Atsushi*; Nakagawa, Koichi*

JNC-TJ8400 2000-064, 175 Pages, 2000/03


In this study, the following tasks have been performed in order to evaluate the stability of earthquake resistance for the engineered barrier system(EBS) of High Level Waste (HLW) geological isolation system. (1)validation studies for the liquefaction model. The function of single-phase analysis without interaction between soil and pore water in three-dimensional effective stress analysis code, which had been developed in this study, have been verified using by actual vibration test data. This fiscal year, some validation studies for the function of liquefaction analysis was conducted usig by actual measured data through the laboratory liquefaction test. (2)Supplemental Studies for JNC Second Progress Report. Through the JNC second progress report, it was considered that the stability of earthquake resistance of the engineered barrier system would be maintained under the major seismic event. At the same time we have recognized that several model parameters for joint-crack element, which takes into account for the response behavior of material discontinuous surface such as between overpack and buffer material, will become important in the response behavior of the whole EBS. This year, we have studied about several topics, which arise from technical discussion on JNC second progress report and we have discussed about total seismic stability of EBS. (3)Supplemental Studies for joint study with NRIDP. At this fiscal year, the joint study with National Research Institute for Disaster Prevention (NRIDP) will be final stage. UP to this day, incremental validation studies had been continued using by mesuared data obtained from vibration test. In this final stage, validation analysis has been conducted again using by current version new analysis code and maintained the validation data which will be contribute to the joint study mentioned above.

JAEA Reports


Ikeda, Takao*; Yoshida, Hideji*

JNC-TJ7400 2000-006, 159 Pages, 2000/02


no abstracts in English

JAEA Reports

Sodium combustion computer code ASSCOPS Version 2.1; User's manual

Ohno, Shuji; Matsuki, Takuo*; ;

JNC-TN9520 2000-001, 196 Pages, 2000/01


ASSCOPS (Analysis of Simultaneous Sodium Combustion in Pool and Spray) has been developed for analyses of thermal consequences of sodium leak and fire accidents in LMFBRs. This report presents a description of the computational models, input and output data as the user's manual of ASSCOPS version 2.1. ASSCOPS is an integrated computational code based on the sodium pool fire code SOFIRE II developed by the Atomics International Division of Rockwell International, and on the sodium spray fire code SPRAY developed by the Hanford Engineering Development Laboratory in the U.S. The users of ASSCOPS need to specify the sodium leak conditions (leak flow rate and temperature, etc.), the cell geometries (cell volume, surface area and thickness of structures, etc.), and the atmospheric initial conditions such as gas temperature, pressure, and composition. ASSCOPS calculates the time histories of atmospheric temperature, pressure and of structural temperature.

JAEA Reports

The Development of MESHNOTE Code for Radionuclide Migration in the Near Field

; Makino, Hitoshi; Peter*

JNC-TN8400 99-095, 69 Pages, 1999/12


MESHNOTE code was developed to evaluate the engineered barrier system in collaboration with QuantiSci. This code is used to simulate glass dissolution, diffusive transport of nuclides in the buffer material and release to surrounding host rock. MESHNOTE is a one-dimensional finite difference, code, which uses cylindrical co-ordinates for the solution of a radially symmetric diffusion problem. MESHNOTE has the followig characteristics: (1) MESHNOTE can solve for diffusive transport of nuclides through an annulus shaped buffer region while accounting for multiple decay chains, linear and non-linear sorption onto the buffer materials and elemental solubility limits; (2) MESHNOTE can solve for ingrowth of plural daughter nuclides from a singular parent nuclide (branching), and the ingrowth of a singular daughter nuclide from plural parent nuclides (rejoining); (3) MESHNOTE can treat the leaching of nuclide from the vitrified waste and the release of nuclide from buffer to surrounding rock, which are boundary conditions for migration in the buffer, basing on the phenomena; (4) MESHNOTE can treat principal parameters (e.g. solubility and distribution coefficient) relevant to nuclide migration as time and space-dependence parameters; (5) The time stepping scheme in MESHNOTE is controlled by tolerance defined by the user. The time stepping will increase automatically while checking the accuracy of the numerical solution. The conceptual model, the mathematical model and the numerical implementation of the MESHNOTE code are described in this report and the characteristic functions of MESHNOTE are verified by comparing with analytical solutions or simulations produced with other calculation codes.

JAEA Reports

Analysis of the Rossendorf SEG experiments using the JNC route for reactor calculation

Dietze, K.

JNC-TN9400 99-089, 20 Pages, 1999/11


The integral experiments performed at the Rossendorf fast-thermal coupled reactor RRR/SEG have been reanalyzed using the JNC route for reactor calculation JENDL3.2/SLAROM / CITATION / JOINT / PERKY. The Rossendorf experiments comprise sample reactivity measurements with pure fission products and structural material in five configurations with different neutron and adjoint spectra. The shapes of the adjoint spectra have been designed to get high sensitivities to neutron capture or the scattering effect. The calculated neutron and adjoint spectra are in good agreement with former results obtained with the European route JEF2.2/ECCO/ERANOS. The C/E-values of the central reactivity worths of samples under investigation are given. Deviations in the results of both routes are due to the different libraries, codes, and self-shielding treatments used in the calculations. Results outside of the error are discussed.

JAEA Reports

Thermal calculation of bituminized product, 1; Thermal evaluation of bituminized product using heat transporting calculation


JNC-TN8410 99-044, 189 Pages, 1999/10


This report includes several results that were made by calculation with several methods to clarify the cause of the fire and explosion incident. In the early times, we didn't have exact information of chemicaI property, reaction rate and any physical constants that we needed. But because the only data that indicate the cooling process of bituminized product was reported, we made heat-transporting calculation with taking this data. Based on the theory of the thermal hazard evaluation that was called Semenov theory or Frank-Kamenetskii theory, the amount of heat generation was estimated using the heat transporting calculation. Common theories were introduced in first section. In the second section, several results of heat transporting calculation were indicated. Calculations were made as follows. First, the model of bituminized product that was filled in the drum was created with the data of cooling process. Second, when the heat was generated in the drum, time-dependent temperature distribution was calculated. And last, judging from the balance of heat generation and heat radiation the critical heat rate was estimated.

JAEA Reports

Groundwater Evolution Modeling for the Second Progress Performance Assessment (PA) Report

Yui, Mikazu; Sasamoto, Hiroshi; Randolph C Arthu*

JNC-TN8400 99-030, 201 Pages, 1999/07


According to the Japanese program for research and development of high level radioactive waste (HLW) disposal defined by Atomic Energy Commission (AEC), the second progress report (i.e., H-12 report) for performance assessment (PA) of HLW disposal is to be published by the Japan Nuclear Cycle Development Institute (JNC) and submitted to the Japanese government before the year 2,000 (AEC, 1997). This report presents the establishment of generic groundwater chemical compositions for the PA supporting the H-12 report. The following five hypothetical groundwaters are categorized for PA based on the results of the first progress report (i.e., H-3 report) and binaly statistical analyses of the screened groundwater dataset: (1)FRHP(Fresh-Reducing-High-pH) groundwater (2)FRLP(Fresh-Reducing-Low-pH) groundwater (3)SRHP(Saline-Reducing-High-pH) groundwater (4)SRLP(Saline-Reducing-Low-pH) groundwater (5)MRNP(Mixing-Reducing-Neutral-pH) groundwater. In order to define representative groundwater compositions for the PA for the H-12 report, JNC has established the representativeness of the above five hypothetical groundwaters by considering the results of multivariate statistical analyses, data reliability, evidence for geochemical controls on groundwater chemistry and exclusion criteria for potential repository sites in Japan. As a result, the following hypothetical reference groundwaters are selected for the performance assessment analysis in H-12 report, respectively: (1)Reference Case groundwater: FRHP groundwater, and (2)Alternative Geological Environment Case groundwater: SRHP groundwater. In addition, JNC has consulted with overseas experts on the concepts used in groundwater evolution modeling. This modeling effort has focussed on simulating equilibrium water-rock interactions to predict groundwater compositions resulting from reactions between initial water compositions and rock mineral assemblages. These discussions have centered on recommendations for developing ...

JAEA Reports

PEGASUS: A Preequilibrium and multi-step evaporation code for neutron cross section calculation

Nakagawa, Tsuneo; *; Sugi, Teruo*; *

JAERI-Data/Code 99-031, 78 Pages, 1999/06


no abstracts in English

JAEA Reports

Hydrogen and tritium behaviour in Monju; Validation of an analysis code for tritium transport in fast reactor system, TTT, and estimation for Monju full power operation in future


JNC-TN4400 99-002, 192 Pages, 1999/03


The tritium transport analysis code, TTT, has been validated using data from the low power test of Monju, and then its behaviour at along term full power operation of Monju in future has been estimated, when the estimated transport and distribution of tritium in the reactor system has been also compared with the result in Joyo and Phenix, which had been already experienced long term operations. The TTT code had been develpped using the tiritium and hydrogen transport model proposed by R. Kumar, ANL, and had been applied to the evaluation in Monju design work. After then, futhermore, the code has been improved using the data from long term operation of Joyo with MK-II core, and in this work the code has been validated for the first time for Monju data. The results from this work are as follows; (1)Comparison of the best fitted tritium source rates from cores in Joyo, Phenix and Monju makes an estimation of the major source from control rods, (2)The calculated tritium concentration in each medium for cooling and its change is a reasonable agreement to the measured, C/E=1.1, (3)The cover gas transport model cosidering isotopic exchange of H and H$$^{3}$$ can reproduce reasonably the measured concentration distirbution of tritium in sodium and cover gas, (4)The tritium concentration in secondary sodium of Monju was about l/50 times as much as the primary one, which shows the acceraration effect on cold tarapping of tritium due to coprecipitation with permeated hydrogen through Evaporater (EV) heat conduction tube walls. The tritium cold trapping efficiency was estimated to be 1 for coprecipitation with hydrogen and 0.3 for isotopic exchange, respectively, (5)Tritium transport and distribution for along term full power operation of Monju in future was estimated, which could involve a excess factor to 4 at the maximum. The tritium concentration in sodium and Steam Generator (SG) water will be substantially saturated after somthing like 10 years full power operation, ...

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