Proposal on how to proceed with Verification and Validation of radiation shielding analyses

Okumura, Keisuke; Sakamoto, Yukio*; Tsukiyama, Toshihisa*

Shahei Kaiseki No V&V Gaidorain Sakutei Ni Mukete, p.4 - 8, 2023/03

https://www.aesj.net/uploads/com_vvshieldingcalc20230428_sup.pdf

no abstracts in English

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 BC

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 BC 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 BC 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 BC pellet and SS to investigate the temperature sensitivity on the eutectic reaction behavior.

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

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

https://doi.org/10.1299/transjsme.16-00431

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.

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

https://doi.org/10.13182/FST05-A965

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.

Research on physico-chemical behaviors of actinides

Ban, Yasutoshi; Mineo, Hideaki; Asakura, Toshihide; Hotoku, Shinobu; Matsumura, Masakazu; Kim, S.-Y.; Morita, Yasuji

JAERI-Conf 2004-011, p.101 - 102, 2004/07

Experimental studies and numerical analysis on physical and chemical behavior of actinide elements in the solutions of aqueous reprocessing process have been done for compiling technical data that are necessary for evaluating processes such as reprocessing facilities. The objective of the present study is obtaining technical data that are conductive to the construction of reprocessing process that cope with high burn-up nuclear fuel, the evaluation of nuclear fuel cycle, and the drawing up policy on reprocessing.

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.

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

https://doi.org/10.1016/S0920-3796(00)00573-1

no abstracts in English

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.

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.

Reliability evaluation for radionuclide transport analysis code MATRICS

*; Ijiri, Yuji*; *; *

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

JNC-TN8400-2000-021.pdf:4.38MB

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 1.010 to 1.010(m/s). (2)In case of the finite matrix diffusion distance (0.031.0(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 1.010 to 1.010(m/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 5.010 to 2.010, and that of transmissivity below 1.010(m/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.