The OECD/NEA Working Group on the Analysis and Management of Accidents (WGAMA); Advances in codes and analyses to support safety demonstration of nuclear technology innovations

Nakamura, Hideo; Bentaib, A.*; Herranz, L. E.*; Ruyer, P.*; Mascari, F.*; Jacquemain, D.*; Adorni, M.*

Proceedings of International Conference on Topical Issues in Nuclear Installation Safety; Strengthening Safety of Evolutionary and Innovative Reactor Designs (TIC 2022) (Internet), 10 Pages, 2022/10

The Study of oxidative stress status in the organs exposed to low dose/low dose-rate radiation (Contract research); FY2019 Nuclear Energy Science & Technology and Human Resource Development Project

Collaborative Laboratories for Advanced Decommissioning Science; Tohoku University*

JAEA-Review 2020-048, 49 Pages, 2021/01

JAEA-Review-2020-048.pdf:4.38MB

The Collaborative Laboratories for Advanced Decommissioning Science (CLADS), Japan Atomic Energy Agency (JAEA), had been conducting the Nuclear Energy Science & Technology and Human Resource Development Project (hereafter referred to "the Project") in FY2019. The Project aims to contribute to solving problems in the nuclear energy field represented by the decommissioning of the Fukushima Daiichi Nuclear Power Station, Tokyo Electric Power Company Holdings, Inc. (TEPCO). For this purpose, intelligence was collected from all over the world, and basic research and human resource development were promoted by closely integrating/collaborating knowledge and experiences in various fields beyond the barrier of conventional organizations and research fields. The sponsor of the Project was moved from the Ministry of Education, Culture, Sports, Science and Technology to JAEA since the newly adopted proposals in FY2018. On this occasion, JAEA constructed a new research system where JAEA-academia collaboration is reinforced and medium-to-long term research/development and human resource development contributing to the decommissioning are stably and consecutively implemented. Among the adopted proposals in FY2019, this report summarizes the research results of the "The study of oxidative stress status in the organs exposed to low dose/low dose-rate radiation". This study investigates the biological effects of low dose/low dose-rate radiation exposure, which is of great social interest, on the oxidative stress status of individual organs and will contribute to the collection of scientific data in a dose range to be required. An interdisciplinary collaborative study discussed the correlation between radiation dose and the biological effect by analyzing the samples of wild Japanese macaques exposed to radiation due to the accident of Fukushima nuclear power station and of animal experiments.

Thermal design study of lead-bismuth cooled accelerator driven system, 1; Study on thermal hydraulic behavior under normal operation condition

Akimoto, Hajime; Sugawara, Takanori

JAEA-Data/Code 2016-008, 87 Pages, 2016/09

JAEA-Data-Code-2016-008.pdf:15.62MB

Thermal hydraulic behavior in a lead-bismuth cooled accelerator driven system (ADS) is analyzed under normal operation condition. Input data for the ADS version of J-TRAC code have been constructed to integrate the conceptual design. The core part of the ADS is modeled in detail to evaluate the core radial power profile effect on the core cooling. As the result of the analyses, the followings are found; (1) Both maximum clad temperature and fuel temperature are below the design limits. (2) The radial power profile has little effect on the coolant flow distribution among fuel assemblies. (3) The radial power profile has little effect on the heat transfer coefficients along fuel rods. (4) The thermal hydraulic behaviors along four steam generators are identical. The thermal hydraulic behaviors along two pumps are also identical. A fast running input data is developed by the simplification of the detailed input data based on the findings mentioned above.

Development of thermal-hydraulic design code for transmutation system with lead-bismuth cooled accelerator driven reactor

Akimoto, Hajime

JAEA-Data/Code 2014-031, 75 Pages, 2015/03

JAEA-Data-Code-2014-031.pdf:37.23MB

A thermal-hydraulic analysis code for transmutation system with lead-bismuth cooled accelerator-driven system (ADS) has been developed using the Japanese-version of Transient Reactor Analysis Code (J-TRAC) as the framework to apply the design studies of ADS. To identify the required capabilities of the thermal-hydraulic analysis code for ADS, previous thermal-hydraulic analyses of light water reactors, sodium-cooled fast reactor and ADS have been surveyed. To make up for insufficient capabilities of the J-TRAC code as a thermal-hydraulic analysis code of ADS, physical properties of lead-bismuth eutectic (LBE), argon gas and nitride nuclear fuel were implemented to the J-TRAC code. It was confirmed that the implemented capabilities worked as expected through verification calculations on (1) single-phase LBE flow, (2) heat transfer in a fuel assembly, and (3) heat transfer in a steam generator.

Development of dynamic analysis code for HTTR hydrogen production system (Contract research)

Maeda, Yukimasa; Nishihara, Tetsuo; Ohashi, Hirofumi; Sato, Hiroyuki; Inagaki, Yoshiyuki

JAERI-Data/Code 2005-001, 149 Pages, 2005/03

JAERI-Data-Code-2005-001.pdf:12.66MB

A heat and mass balance analysis code (N-HYPAC) has been developed to investigate transient behavior in the HTTR hydrogen production system. The code can analyze heat and mass transfer (temperature and mass and pressure distributions of process and helium gases) and behavior of the control system under both static state(case of steady operation) and dynamic state(case of transient operation). Analysis model of helium and process gases from IHX to secondary helium loop and hydrogen production system has been constructed. This report describes analytical flow sheet, construction of the code, basic equations, method to treat the input data, estimation of the preliminary analysis.

Analytical study on fire and explosion accidents assumed in HTGR hydrogen production system

Inaba, Yoshitomo; Nishihara, Tetsuo; Nitta, Yoshikazu*

Nuclear Technology, 146(1), p.49 - 57, 2004/04

https://doi.org/10.13182/NT04-A3486

One of the most important safety design issues for a hydrogen production system coupling with a High Temperature Gas-cooled Reactor (HTGR) is to ensure reactor safety against fire and explosion accidents because a large amount of combustible fluid is dealt with in the system. The Japan Atomic Energy Research Institute (JAERI) has a demonstration test plan of a hydrogen production system by steam reforming of methane coupling with the High Temperature engineering Test Reactor (HTTR). In the plan, we developed the P2A code system to analyze event sequences and consequences in detail on the fire and explosion accidents assumed in the HTGR or HTTR hydrogen production system. This paper described the three accident scenarios assumed in the system, the structure of P2A, the analysis procedure with P2A and the results of the numerical analyses based on the accident scenarios, and it was showed that P2A was a useful tool for the accident analysis in the system.

An EGS4 user code developed for design and optimization of -ray detection systems

Oishi, Tetsuya; Tsutsumi, Masahiro; Sugita, Takeshi*; Yoshida, Makoto

Journal of Nuclear Science and Technology, 40(6), p.441 - 445, 2003/06

https://doi.org/10.3327/jnst.40.441

An EGS4 user code has been developed to design and optimize ray detection systems for several types of radiation sources. The code is fundamentally based on the PRESTA-CG, which is improved on the electron transport in the EGS4 and specialized for the utilization of a combinatorial geometry (CG) method. The main additional functions in the present user code are classified into two parts of the definition of radiation sources and the transport of photons. The developed user code was applied to two types of detection systems in order to demonstrate its availability. As the result, it was found that the present code allows the detailed response analysis of complicated detection systems for several sources with just a simple handling.

Development of fire and explosion analysis code system -P2A- for HTGR hydrogen production system (Contract research)

Inaba, Yoshitomo; Nishihara, Tetsuo; Moriyama, Koichi*; Nakamura, Masashi*

JAERI-Data/Code 2002-014, 255 Pages, 2002/07

JAERI-Data-Code-2002-014.pdf:21.69MB

One of the most important safety design issues for an HTGR hydrogen production system is to ensure reactor safety against fire and explosion accidents in the hydrogen production plant because a large amount of combustible fluid is dealt with in the system. JAERI has the demonstration test plan to connect the hydrogen production system with the HTTR. In the plan, we considered effective measures against the fire and explosion accidents in the HTTR hydrogen production system, which were applicable to the HTGR hydrogen production system of a commercial base, and also developed the P2A code system to analyze event sequences and consequences in detail, on assumed fire and explosion accidents in the HTGR hydrogen production system and the HTTR hydrogen production system. The P2A can analyze the process of leakage, dispersion, ignition, and combustion including deflagration and detonation of the combustible fluid in the internal and external area of the reactor building. In this report, we describe the outline and the usage of the P2A, and the results of preliminary calculations.

Estimates of collective doses from a hypothetical accident of a nuclear submarine

Kobayashi, Takuya; Togawa, Orihiko; Odano, Naoteru; Ishida, Toshihisa

Journal of Nuclear Science and Technology, 38(8), p.658 - 663, 2001/08

https://doi.org/10.3327/jnst.38.658

The collective dose to the Japanese population has been estimated from a hypothetical accident of a nuclear submarine if it sinks in an offshore region around Japan. A computer code system DSOCEAN has been used for assessing the collective dose due to radionuclides released to the ocean from a sunken nuclear submarine. The maximum of the estimated collective effective dose equivalent by the annual intake of marine products after radionuclide releases for one year is approximately 0.5% of the annual average dose by the natural radiation that is reported by UNSCEAR.

A Probabilistic assessment code system for derivation of clearance levels of radioactive materials; PASCLR user's manual

Takahashi, Tomoyuki*; Takeda, Seiji; Kimura, Hideo

JAERI-Data/Code 2000-041, 108 Pages, 2001/01

JAERI-Data-Code-2000-041.pdf:4.86MB

no abstracts in English