Aoki, Takeshi; Isaka, Kazuyoshi; Sato, Hiroyuki; Ohashi, Hirofumi
Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08
The flow distribution analysis performed in the HTGR design has to take into account the interaction thermal and radiation deformations of the graphite structure, and the gaps between the graphite structures forming unintended flow. In the present study, a user-friendly flow network calculation code (FNCC) has been developed on the basis of experiences of High Temperature engineering Test Reactor (HTTR) design for HTGR design with enhanced compatibility with other HTGR design codes and with considering graphite block deformation in iteration process without manual control. The validation of FNCC was performed for the one-column flow distribution test. The analytical results using FNCC showed good agreement with the experimental results. It is concluded that FNCC was validate for the analysis of distributions of flowrate and pressure for the flow network model including the unintended flow paths in prismatic-type HTGRs.
Inaba, Yoshitomo; Isaka, Kazuyoshi; Shibata, Taiju
JAEA-Data/Code 2017-002, 74 Pages, 2017/03
In order to ensure the thermal integrity of fuel in High Temperature Gas-cooled Reactors (HTGRs), it is necessary that the maximum fuel temperature in normal operation is to be lower than a thermal design target. In the core thermal-hydraulic design of block-type HTGRs, the maximum fuel temperature should be evaluated considering data such as core geometry and specifications, power density and neutron fluence distributions, and core coolant flow distribution. The fuel temperature calculation code used in the design stage of the High Temperature engineering Test Reactor (HTTR) presupposes to run on UNIX systems, and its operation and execution procedure are complicated and are not user-friendly. Therefore, a new fuel temperature calculation code, named FTCC, which has a user-friendly system such as a simple and easy operation and execution procedure, was developed. This report describes the calculation objects and models, the basic equations, the strong points (improvement points from the HTTR design code), the code structure, the using method of FTCC, and the result of a validation calculation with FTCC. The calculation result obtained by FTCC provides good agreement with that of the HTTR design code, and then FTCC will be used as one of the design codes for high temperature gas-cooled reactors. In addition, the effect of hot spot factors and fuel cooling forms on reducing the maximum fuel temperature is investigated with FTCC. As a result, it was found that the effect of center hole cooling for hollow fuel compacts and gapless cooling with monolithic type fuel rods on reducing the temperature is very high.
Nomoto, Yasunobu; Aihara, Jun; Nakagawa, Shigeaki; Isaka, Kazuyoshi; Ohashi, Hirofumi
JAEA-Data/Code 2015-008, 39 Pages, 2015/06
HTFP is a calculation code for amount of additionally released fission product (FP) from fuel rods of pin-in-type according to transient of core temperature at the accident of high temperature gas-cooled reactors (HTGRs). This code analyzes FP release inventory from core according to the transient of core temperature at the accident as an input data and considering FP release rate from a fuel compact and a graphite sleeve and radioactive decay of FP. This report describes the outline of HTFP code and its input data. The computed solutions using the HTFP code were compared to those of HTCORE code, which was used for the design of the High Temperature Engineering Test Reactor (HTTR) to validate the analysis models of the HTFP code. The comparison of HTFP code results with HTCORE code results showed the good agreement.
Inaba, Yoshitomo; Isaka, Kazuyoshi; Fukaya, Yuji; Tachibana, Yukio
JAEA-Data/Code 2014-023, 64 Pages, 2015/01
The Japan Atomic Energy Agency has performed the conceptual designs of small-sized High Temperature Gas-cooled Reactor (HTGR) systems, aiming for the deployment of the systems to overseas such as developing countries. The small-sized HTGR systems can provide power generation by steam turbine, high temperature steam for industry process and/or low temperature steam for district heating. In the core thermal and hydraulic designs of HTGRs, it is important to evaluate the maximum fuel temperature so that the thermal integrity of the fuel is ensured. In order to calculate and evaluate the fuel temperature on personal computers (PCs) in a convenient manner, the calculation file based on the Microsoft Excel were developed. In this report, the basic equations used in the calculation file, the calculation method and procedure, and the results of the validation calculation are described.
Shimazaki, Yosuke; Isaka, Kazuyoshi; Nomoto, Yasunobu; Seki, Tomokazu; Ohashi, Hirofumi
JAEA-Technology 2014-038, 51 Pages, 2014/12
The analytical models for the evaluation of graphite oxidation were implemented into the THYTAN code, which employs the mass balance and a node-link computational scheme to evaluate tritium behavior in the High Temperature Gas-cooled Reactor (HTGR) systems for hydrogen production, to analyze the graphite oxidation during the air or water ingress accidents in the HTGR systems. This report describes the analytical models of the THYTAN code in terms of the graphite oxidation analysis and its verification and validation (V&V) results. Mass transfer from the gas mixture in the coolant channel to the graphite surface, diffusion in the graphite, graphite oxidation by air or water, chemical reaction and release from the primary circuit to the containment vessel by a safety valve were modeled to calculate the mass balance in the graphite and the gas mixture in the coolant channel. The computed solutions using the THYTAN code for simple questions were compared to the analytical results by a hand calculation to verify the algorithms for each implemented analytical model. A representation of the graphite oxidation experimental was analyzed using the THYTAN code, and the results were compared to the experimental data and the computed solutions using the GRACE code, which was used for the safety analysis of the High Temperature Engineering Test Reactor (HTTR), in regard to corrosion depth of graphite and oxygen concentration at the outlet of the test section to validate the analytical models of the THYTAN code. The comparison of THYTAN code results with the analytical solutions, experimental data and the GRACE code results showed the good agreement.
Aihara, Jun; Goto, Minoru; Inaba, Yoshitomo; Isaka, Kazuyoshi; Ohashi, Hirofumi; Tachibana, Yukio
JAEA-Technology 2014-009, 29 Pages, 2014/05
Japan Atomic Energy Agency (JAEA) is carrying out conceptual design of a 50 MWt small-sized high temperature gas cooled reactor (HTGR), HTR50S. In this report, integrity of coated fuel particles (CFPs) is evaluated for core of HTR50S of 1st. step of phase I (first core of HTR50S) under normal operation. CFPs are considered to be failed by fuel kernel migration by temperature gradient in CFPs or corrosion of SiC layer by fission product Pd (Pd corrosion) or increase in internal pressure under normal operation. In this report, integrity of CFPs is to be maintained for each phenomenon.
Ohashi, Hirofumi; Sato, Hiroyuki; Tazawa, Yujiro; Aihara, Jun; Nomoto, Yasunobu; Imai, Yoshiyuki; Goto, Minoru; Isaka, Kazuyoshi; Tachibana, Yukio; Kunitomi, Kazuhiko
JAEA-Technology 2013-017, 71 Pages, 2014/02
Japan Atomic Energy Agency (JAEA) has started a conceptual design of a 50 MWt small-sized high temperature gas cooled reactor (HTGR) for steam supply and electricity generation (HTR50S). Though the safety design of HTR50S was determined based on that of the High Temperature Engineering Test Reactor (HTTR) for the early deployment of HTR50S, the shutdown cooling system, which is the forced cooling heat removal system, was categorized as non-safety class to optimize the protection to provide the highest level of safety that can reasonably be achieved, and the vessel cooling system, which is categorized as the safety class system, was designed as a passive safety features. The preliminary safety analysis of HTR50S for the rupture of co-axial hot gas duct in primary cooling system and the tube rupture of steam generator was conducted to confirm the adequacy of the safety design. It was confirmed that the analysis results satisfied the acceptance criteria.
Kubo, Shinji; Ijichi, Masanori*; Hodotsuka, Masatoshi; Yoshida, Mitsunori*; Kasahara, Seiji; Isaka, Kazuyoshi; Tanaka, Nobuyuki; Imai, Yoshiyuki; Onuki, Kaoru
Proceedings of 2007 AIChE Annual Meeting (CD-ROM), 7 Pages, 2007/11
no abstracts in English
Ijichi, Masanori; Yoshida, Mitsunori; Isaka, Kazuyoshi; Tanaka, Nobuyuki; Kasahara, Seiji; Okuda, Hiroyuki; Hodotsuka, Masatoshi; Kanagawa, Akihiro; Imai, Yoshiyuki; Noguchi, Hiroki; et al.
no journal, ,
no abstracts in English
Ijichi, Masanori; Kasahara, Seiji; Yoshida, Mitsunori; Isaka, Kazuyoshi; Tanaka, Nobuyuki; Imai, Yoshiyuki; Kubo, Shinji; Onuki, Kaoru; Hino, Ryutaro
no journal, ,
High efficiency flowsheet for thermochemical hydrogen production IS (Iodine-Sulfur) process is investigating. In order to establish the reference flowsheet that is concidering the recent experimental data and engineering investigation, steady state process simulation was done using process simulation code PRO/2 and thermochemical model/database(developed by OLI Systems, Inc.) which was extended for IS process. The results of parameter analyses and established reference flowsheet are reported.
Yamagishi, Isao; Kato, Chiaki; Nagaishi, Ryuji; Arisaka, Makoto; Uruga, Kazuyoshi*; Tsukada, Takeshi*
no journal, ,
no abstracts in English