Neutronic design of neutron moderator on a reentrant-hole configuration for Kyoto University Accelerator-based Neutron Source (KUANS)

Okita, Shoichiro; Tasaki, Seiji*; Abe, Yutaka*

Nihon Genshiryoku Gakkai Wabun Rombunshi, 19(3), p.178 - 184, 2020/09

https://doi.org/10.3327/taesj.J19.028

The Kyoto University Accelerator-based Neutron Source (KUANS) is a compact neutron source that is mainly used for spectrometer and detector development. In addition, it is also suited for experiments to study the neutronic design of moderators owing to the relatively low neutron generation yield by Be(p,n). We present a neutronic design of the neutron moderator on a reentrant-hole configuration for KUANS to enhance the neutron emission, and some experiments are conducted at KUANS for verification. A polyethylene moderator on a reentrant-hole configuration is designed by PHITS calculation and is introduced to KUANS to obtain intense oblong neutron beams. The intensity of the pulsed neutron beam is experimentally measured. The results reveal that the intensity becomes approximately 1.9 times stronger than that of the conventional rectangular design. In addition, the ratio of its intensity to the conventional intensity increases to approximately threefold as the neutron wavelength increases. It is interesting to note that the longer the neutron wavelength, the more efficiently they are extracted from the inside of the moderator owing to the existence of the reentrant-hole configuration.

Development of fuel temperature calculation code "FTCC" for high temperature gas-cooled reactors

Inaba, Yoshitomo; Isaka, Kazuyoshi; Shibata, Taiju

JAEA-Data/Code 2017-002, 74 Pages, 2017/03

JAEA-Data-Code-2017-002.pdf:2.36MB

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.

Development of fuel temperature calculation file for high temperature gas-cooled reactors

Inaba, Yoshitomo; Isaka, Kazuyoshi; Fukaya, Yuji; Tachibana, Yukio

JAEA-Data/Code 2014-023, 64 Pages, 2015/01

JAEA-Data-Code-2014-023.pdf:7.15MB

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.

Design study around beam window of ADS

Oigawa, Hiroyuki; Tsujimoto, Kazufumi; Kikuchi, Kenji; Kurata, Yuji; Sasa, Toshinobu; Umeno, Makoto*; Nishihara, Kenji; Saito, Shigeru; Mizumoto, Motoharu; Takano, Hideki*; et al.

Proceedings of 4th International Workshop on the Utilisation and Reliability of High Power Proton Accelerators, p.325 - 334, 2005/11

The Japan Atomic Energy Research Institute (JAERI) is conducting the research and development (R&D) on the Accelerator-Driven Subcritical System (ADS) for the effective transmutation of minor actinides (MAs). The ADS proposed by JAERI is the 800 MWth, Pb-Bi cooled, tank-type subcritical reactor loaded with (MA+Pu) nitride fuel. The Pb-Bi is also used as the spallation target. In this study, the feasibility of the ADS was discussed with putting the focus on the design around the beam window. The partition wall was placed between the target region and the ductless-type fuel assemblies to keep the good cooling performance for the hot-spot fuel pin. The flow control nozzle was installed to cool the beam window effectively. The thermal-hydraulic analysis showed that the maximum temperature at the outer surface of the beam window could be repressed below 500 C even in the case of the maximum beam power of 30 MW. The stress caused by the external pressure and the temperature distribution of the beam window was also below the allowable limit.

Core thermal-hydraulic design

Takada, Eiji*; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tochio, Daisuke

Nuclear Engineering and Design, 233(1-3), p.37 - 43, 2004/10

https://doi.org/10.1016/j.nucengdes.2004.07.009

The core thermal-hydraulic design for the HTTR is carried out to evaluate the maximum fuel temperature at normal operation and anticipated operation occurrences. To evaluate coolant flow distribution and maximum fuel temperature, we use the experimental results such as heat transfer coefficient, pressure loss coefficient obtained by mock-up test facilities. Furthermore, we evaluated hot spot factors of fuel temperatures conservatively. As the results of the core thermal-hydraulic design, an effective coolant flow through the core of 88 % of the total flow is achieved at minimum. The maximum fuel temperature appears during the high temperature test operation, and reaches 1492 C for the maximum through the burn-up cycle, which satisfies the design limit of 1495 C at normal operation. It is also confirmed that the maximum fuel temperature at any anticipated operation occurrences does not exceed the fuel design limit of 1600 C in the safety analysis. On the other hand, result of re-evaluation of analysis condition and hot spot factors based on operation data of the HTTR, the maximum fuel temperature for 160 effective full power operation days is estimated to be 1463 C. It is confirmed that the core thermal-hydraulic design gives conservative results.

Mercury target thermal hydraulic design for JAERI spallation neutron source

Kaminaga, Masanori; Haga, Katsuhiro; Kinoshita, Hidetaka; Terada, Atsuhiko*; Koikegami, Hajime*; Hino, Ryutaro

Proceedings of 11th International Conference on Nuclear Engineering (ICONE-11) (CD-ROM), 7 Pages, 2003/04

1MW pulsed spallation neutron source using a mercury target will be constructed in order to produce high-intensity neutrons for use in the fields of life and material sciences. The mercury target is proposed using a high-energy (3.0GeV) proton beam with a current of 0.333mA and operating at 25 Hz with a pulse duration less than 1s. A cross flow type (CFT) mercury target has been designed in order to distribute mercury flow according to an axial heat generation distribution caused by spallation reaction, based on the thermal hydraulic analytical results of 3GeV, 1MW proton beam injection by using the STAR-CD code. This paper presents the final CFD analytical results. In the analysis, an inlet temperature of 50C and an inlet mercury velocity of 1.0m/s were assumed. As results, a maximum velocity of 2.48m/s was observed near the front end of the outlet plenum and a maximum of 125.5C was observed near the beam window where the volumetric heat generation rate was relatively large. The maximum temperature is far below the mercury saturation temperature of 356C under atmospheric pressure. This result satisfied the thermal-hydraulic design criteria of "Maximum mercury temperature in the target shall be less than 300C".

Experimental study on the critical heat flux (CHF) in a rectangular channel with micro ribs for a solid target and proton beam window design

Kaminaga, Masanori; Kinoshita, Hidetaka; Haga, Katsuhiro; Hino, Ryutaro; Sudo, Yukio

Proceedings of International Workshop on Current Status and Future Directions in Boiling Heat Transfer and Two-Phase Flow, p.135 - 141, 2000/00

no abstracts in English

Conceptual design of a 50-MW severe-accident-free HTR and the related test program of the HTTR

Kunitomi, Kazuhiko; Tachibana, Yukio; Saikusa, Akio; Sawa, Kazuhiro; L.M.Lidsky*

Nuclear Technology, 123(3), p.245 - 258, 1998/09

https://doi.org/10.13182/NT98-A2896

no abstracts in English

Development of intelligent code system to support conceptual design of nuclear reactor core

Kugo, Teruhiko; Nakakawa, Masayuki;

Journal of Nuclear Science and Technology, 34(8), p.760 - 770, 1997/08

https://doi.org/10.3327/jnst.34.760

no abstracts in English

JPSR (JAERI Passive Safety Reactor)

Murao, Yoshio

Nihon Genshiryoku Gakkai-Shi, 37(9), p.784 - 787, 1995/00

no abstracts in English