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JAEA Reports

Post-processor coding for large-scale transient simulation computer codes

Yoshikawa, Shinji

JAEA-Technology 2019-024, 22 Pages, 2020/03

JAEA-Technology-2019-024.pdf:1.76MB
JAEA-Technology-2019-024-appendix(CD-ROM).zip:73.55MB

In various technical fields of nuclear energy, computer codes are often used for transient simulations of target phenomena. Some of the codes were developed many years ago and have been revised with newly acquired findings, rather than newly developed, because of many encompassed numerical models and complexity of algorithms. In many cases, available outputs for users are output text files and graphs showing temporal variations of parameters, despite diversified and huge number of output information items are posing difficulty to the users in grasping the whole picture of the reproduced phenomena. This report compiles a series of know-hows in building a post-processor software for large simulation codes which serves as an interactive tool for code users in understanding the reproduced consequence with visually understandable information items. These know-hows are acquired through post-processor developments for LWR severe accident simulation codes RELAP/SCDAPSIM and MELCOR.

Journal Articles

Study on eutectic melting behavior of control rod materials in core disruptive accidents of sodium-cooled fast reactors, 5; Validation of a multi-phase model for eutectic reaction between molten stainless steel and B$$_{4}$$C

Liu, X.*; Morita, Koji*; Yamano, Hidemasa

Proceedings of International Nuclear Fuel Cycle Conference / Light Water Reactor Fuel Performance Conference (Global/Top Fuel 2019) (USB Flash Drive), p.47 - 51, 2019/09

Investigation of the eutectic reaction in a core disruptive accident of sodium cooled reactor is of importance since reactor criticality will be affected by the change in reactivity after eutectic reaction. In this study, we performed 1st step of validation analysis using a fast reactor safety analysis code, SIMMER-III, with the developed model based on a new series of experiments, where a B$$_{4}$$C pellet was immersed into a molten stainless steel (SS) pool. The simulation results showed the general behavior of eutectic material formation measured in the experiments reasonably. The eutectic reaction consumes solid B$$_{4}$$C and liquid SS, and then the liquid eutectic composition is produced at the early stage of reaction due to the high temperature of molten SS. Movement of the eutectic material in the molten pool leads to the redistribution of boron element. Molten SS pool then freezes to solid SS and movement of eutectic material is stopped by surrounding solid SS. Boron concentration in the pool was measured after molten SS freezes into a solid. Simulation results indicate that boron tends to accumulate in the upper part of the molten pool. This is attributed to the buoyancy force acting on lighter boron in the molten SS pool. A parametric study was also conducted by changing the initial temperature of B$$_{4}$$C pellet and SS to investigate the temperature sensitivity on the eutectic reaction behavior.

Journal Articles

Review of $$gamma$$-ray exposure buildup factors

Matsuda, Norihiro; Onishi, Seiki*; Sakamoto, Yukio*; Nobuhara, Fumiyoshi*

Heisei 29-Nendo Kani Shahei Kaiseki Kodo Rebyu Wakingu Gurupu Katsudo Hokokusho (Internet), p.20 - 28, 2018/08

no abstracts in English

Journal Articles

Study on applicability of fast reactor plant dynamics analysis code to core thermal hydraulics under natural circulation decay heat removal conditions

Hamase, Erina; Doda, Norihiro; Nabeshima, Kunihiko; Ono, Ayako; Ohshima, Hiroyuki

Nippon Kikai Gakkai Rombunshu (Internet), 83(848), p.16-00431_1 - 16-00431_11, 2017/04

A plant dynamics analysis code Super-COPD is being developed in JAEA for the design and safety assessments of sodium-cooled fast reactors (SFRs). In this study, the friction loss coefficients in the whole core thermal-hydraulic model was modified to improve the prediction accuracy of the sodium temperature distribution in a fuel subassembly under the natural circulation conditions. The modified whole core model was applied to analyses of experiments that were performed by using JAEA's test facility PLANDTL as a part of the code validation study. The obtained numerical results of sodium temperature distributions in the core showed good agreement with the measured data. It implies that the modified whole core model can properly reproduce dominant thermal-hydraulic phenomena in the core region under natural circulation conditions, i.e., flow redistribution among fuel subassemblies as well as in a fuel subassembly and inter-subassembly heat transfer.

Journal Articles

Study of global wall saturation mechanisms in long-pulse ELMy H-mode discharges on JT-60U

Takenaga, Hidenobu; Nakano, Tomohide; Asakura, Nobuyuki; Kubo, Hirotaka; Konoshima, Shigeru; Shimizu, Katsuhiro; Tsuzuki, Kazuhiro; Masaki, Kei; Tanabe, Tetsuo*; Ide, Shunsuke; et al.

Nuclear Fusion, 46(3), p.S39 - S48, 2006/03

 Times Cited Count:17 Percentile:44.97(Physics, Fluids & Plasmas)

no abstracts in English

Journal Articles

Case study on unexpected tritium release happened in a ventilated room of fusion reactor

Iwai, Yasunori; Hayashi, Takumi; Kobayashi, Kazuhiro; Nishi, Masataka

Fusion Science and Technology, 48(1), p.460 - 463, 2005/07

 Times Cited Count:4 Percentile:66.99(Nuclear Science & Technology)

A code has been developed to investigate tritium behavior in a ventilated room at its accidental release. Purpose of present study is to; (1) investigate the effect of atmospheric exchange number on confinement at initial stage of tritium release; (2) investigate the effect of atmospheric exchange number on time necessary for release detection; (3) investigate the suitable location of exhaust ducts and monitors. Essential points of discussion are as follows: (1) Atmospheric exchange number is less influential in confinement. (2) Time until a monitor detects release depends on exchange number but it is within a few minutes in any case. Installation of a monitor in each duct placed uniformly in a room is effective for the prompt detection. (3) After closing the emergency isolation valve, a few hours are needed until the tritium concentration in a room reaches uniform. Released tritium forms plume and it migrates in a room by the eddy flow at its initial stage, so it is important not to discharge plume directly. Hence, it is effective to locate exhaust ducts with some distance from the wall.

Journal Articles

Effects of long time scale variation of plasma wall interactions on particle control in JT-60U

Takenaga, Hidenobu; Nakano, Tomohide; Asakura, Nobuyuki; Kubo, Hirotaka; Konoshima, Shigeru; Shimizu, Katsuhiro; Tsuzuki, Kazuhiro; Ide, Shunsuke; Fujita, Takaaki

Proceedings of 4th IAEA Technical Meeting on Steady-State Operation of Magnetic Fusion Devices and MHD of Advanced Scenarios (Internet), 8 Pages, 2005/02

no abstracts in English

JAEA Reports

A Research program for numerical experiments on coupled thermo-hydro-mechanical and chemical processes

Ito, Akira; Kawakami, Susumu; Yui, Mikazu

JNC-TN8400 2001-028, 38 Pages, 2002/01

JNC-TN8400-2001-028.pdf:2.35MB

In a repository for high-level radioactive waste, coupled thermo -hydro -mechanical and chemical (THMC) processes will ocurr, involving the interactive processes between radioactive decay heat from the vitrified waste, infiltration of groundwater, swelling pressure generation and chemical evolution of the buffer material and porewater chemistry. In this program, numerical experiment system for the coupled THMC processes will be developed in order to predict the long-term performance of the near-field (engineered barrier and host rock) for various geological environments. The simulation code development has been started in FY 2001 and three development steps are planned, because (1)development will be continued for some years, (2)feasibility of numerical experiment have to be confirmed by using existing tools. This report presents the following items of the simulation code development for the coupled THMC processes. (1)First step of the simulation code development (2)Mass transport passways in compacted bentonite (3)Parallelization of the simulation code

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

JNC-TN8400-2001-027.pdf:0.8MB

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.

Journal Articles

Computer code analysis on fuel rod behavior

Suzuki, Motoe

Saishin Kaku Nenryo Kogaku; Kodoka No Genjo To Tembo, p.131 - 140, 2001/06

no abstracts in English

Journal Articles

Three-dimensional numerical simulations of dust mobilization and air ingress characteristics in a fusion reactor during a LOVA event

Takase, Kazuyuki

Fusion Engineering and Design, 54(3-4), p.605 - 615, 2001/04

 Times Cited Count:12 Percentile:32.65

no abstracts in English

JAEA Reports

Fundamental study on interfacial area transport model, 1 (Contract research)

Mishima, Kaichiro*; Nakamura, Hideo

JAERI-Review 2001-012, 51 Pages, 2001/03

JAERI-Review-2001-012.pdf:2.26MB

no abstracts in English

JAEA Reports

Validation of sodium fire analysis code ASSCOPS

Ohno, Shuji; Matsuki, Takuo*

JNC-TN9400 2000-106, 132 Pages, 2000/12

JNC-TN9400-2000-106.pdf:2.8MB

Sodium fire analyses were performed on 7 kinds of sodium leak tests using a computer code ASSCOPS which has been developed to evaluate thermal consequences of sodium leak accident in an FBR plant. By the comparison between the calculated and the test results of gas pressure, gas temperature, sodium catch pan temperature, wall temperature, and of oxygen concentration, it was confirmed that the ASSCOPS code and the parameters used in the analysis give valid or conservative results on thermal consequences of sodium leak and fire.

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

JNC-TN9400-2000-065.pdf:6.26MB
JNC-TN9400-2000-065(errata).pdf:0.12MB

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

Reliability evaluation of simulation models for nearfield groundwater flow and radionuclide transport computation

*; *; *; *

JNC-TJ8400 2000-005, 71 Pages, 2000/05

JNC-TJ8400-2000-005.pdf:4.0MB

In this research, simulations with some parameters which characterize ground water flow and the reliability evaluation for the expansion of the calculation method of groundwater flow were carried out by using the radionuclide transport computations in nearfield heterogeneous porous media. Concretely contents are follows: (1)With the series of calculation method for three-dimensional saturated/unsaturated groundwater flow and one-dimensional radionuclide transport, the computational analyses with the parameters used in JNC report in 2000 was carried out and the influence of the different input flux was evaluated. (2)The examination of the application for the different ways of inverse laplace transformation which is used in one-dimensional radionuclide transport analysis code "MATRICS" was carried out. (3)The examination of the application of multi-element "MATRICS" (m-MATRICS) for radionuclide transport computations in nearfield heterogeneous porous media was carried out. (4)The series of calculation methods from three-dimensional saturated/unsaturated ground water flow simulation code to one-dimensional radionuclide transport simulation code was integrated.

JAEA Reports

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

Tozawa, Katsuhiro; ;

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

JNC-TN9400-2000-052.pdf:4.39MB

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

Reliability evaluation for radionuclide transport analysis code MATRICS

*; Ijiri, Yuji*; *; *

JNC-TN8400 2000-021, 66 Pages, 2000/04

JNC-TN8400-2000-021.pdf:1.74MB

A reliability evaluation for radionuclide transport analysis code, MATRICS, used in radionuclide transport analysis in the natural barrier system PA in H12 report has been carried out. Sensitivity analysis to radionuclide transport parameter in MATRICS and analytical solution has been performed, and the results of each analysis have been compared. Additionally sensitivity analysis using Talbot Method, Crump method and Hosono method has been carried out, and the results of each inverse Laplace transform method has been compared. The conclusions obtained from the results of the evaluation are summarized as follows, (1)In case of the infinite matrix diffusion distance, an error among the results of each calculation is maximum about 0.4% in the range of Pe number from 1.0 to 100. And, an error among the results of each calculation is maximum about 5.5% in the range of transmissivity from 1.0$$times$$10$$^{-10}$$ to 1.0$$times$$10$$^{-5}$$(m$$^{2}$$/s). (2)In case of the finite matrix diffusion distance (0.03$$sim$$1.0(m)), an error among the results of each calculation is maximum about 0.7% in the range of Pe number from 1.0 to 100. And, an error among the results of each calculation is maximum about 2.4% in the range of transmissivity from 1.0$$times$$10$$^{-10}$$ to 1.0$$times$$10$$^{-5}$$(m$$^{2}$$/s). 3)By comparing Talbot method with other inverse Laplace transform method, Talbot method is confirmed to give similar results with other inverse Laplace transform method in the range of Pe number from 5.0$$times$$10$$^{-1}$$ to 2.0$$times$$10$$^{3}$$, and that of transmissivity below 1.0$$times$$10$$^{-7}$$(m$$^{2}$$/s). Therefore, it is concluded that the reliability of MATRICS are confirmed by conducting sensitivity analysis in the range of Pe number and transmissivity coefficient used in H12 report.

JAEA Reports

Modification of the evaluation model for Pu redistribution phenomena

; *;

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

JNC-TN9400-2000-045.pdf:2.47MB

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

A feasibility study of the particle interaction method for the flow regimes with the chemical reaction; (Report under the contract between JNC and Toshiba Corporation)

Shirakawa, Noriyuki*; *; *; *

JNC-TJ9440 2000-008, 47 Pages, 2000/03

JNC-TJ9440-2000-008.pdf:1.96MB

The numerical thermohydraulic analysis of a LMFR component should involve its whole boundaly in order to evaluate the effect of chemical reaction within it. Therefore, it becomes difficult mainly due to computing time to adopt microscopic approach for the chemical reaction directly. Thus, the thermohydraulic code is required to model the chemically reactive fluid dynamics with constitutive correlations. The reaction rate denpends on the binary contact areas between components such as continuous liquids, droplets, solid particles, and bubbles. The contact areas change sharply according to the interface state between components. Since no experiments to study the jet flow with sodium-water chemical reaction have been done, the goal of this study is to obtain the knowledge of flow regimes and contact areas by analyzing the fluid dynamics of multi-pahse and reactive components mechanistically with the particle interaction method. For the first stage of the study, the applicability of this method to the nalysis of a liquid jet into the other liquid pool was investigated. Based on the literatures, we investigated the jet flow mechanisms and analyzed the experiment of a water jet into a gasoline pool. We also analyzed SWAT3/Run19 test, the jet flow in a rod bundle, to study the applicability of the method to a complicated boundary without a chemical reaction model. The calculated fluid dynamics was in good agreement with the experiment. Furthermore, we studied and formulated the paths of phase change and chemical reaction, and conceptually designed the adopting the heat-transfer-limited phase change model and the synthesizd reaction model with a water-hydrogen conversion ratio.

JAEA Reports

Current status and future plan for thermaI striping investigations at JNC

; kasahara, Naoto; ; ; Kamide, Hideki

JNC-TN9400 2000-010, 168 Pages, 2000/02

JNC-TN9400-2000-010.pdf:8.78MB

Thermal striping is significant issue of the structural integrity, where the hot and cold fluids give high cycle fatigue to the structure through the thermal stress resulted from the time change of temperatur distibution in the structure. In the sodium cooled fast reactor, temperature change in fluid easily transfers to the structure because of the high thermal conductivity of the sodium. It means that we have to take care of thermal striping, The thermal striping is complex phenomena between the fluid and structure engineering fields. The investigations of thermal striping are not enough to evaluate the integrity directly. That is the fluctuation intensity at the structure surface is assumed to be temperature difference between source fluids (upstream to the mixing region) as the maximum value in the design. 0therwise, the design conditions are defined by using a mockup experiment and large margin of temperature fluctuation intensity. Furthermore, such evaluation manners have not yet been considered as a design rule. Transfer mechanism of temperature fluctuation from fluid to structure has been investigated by the authors on the view points of the fluid and structure. Attenuation of temperature fluctuation was recognized as a dominant factor of thermal fatigue. We have devdoped a numerical analysis system which can evaluate thermal fatigue and crack growth with consideration of the attenuation of temperature fluctuation in fluid, heat transfer, and structure. This system was applied to a real reactor and the applicability was confirmed. Further verification is planned to generalize the system. For the higher cost performance of the fast reactor, an evaluation rule is needed, which can estimate thermal loading with attenuation and can be applied to the design. An idea of the rule is proposed here. Two methods should be prepared; one is a precise evaluation method where mechanism of attenuation is modeled, and the other is simple evaluation method where ...

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