Thermal-hydraulic analyses of the High-Temperature engineering Test Reactor for loss of forced cooling at 30% reactor power

Takamatsu, Kuniyoshi

Annals of Nuclear Energy, 106, p.71 - 83, 2017/08

http://doi.org/10.1016/j.anucene.2017.03.032

The HTTR, which is the only HTGR having inherent safety features in Japan, conducted a safety demonstration test involving a loss of both reactor reactivity control and core cooling. The paper shows thermal-hydraulics during the LOFC test at an initial power of 30% reactor power (9 MW), when the insertion of all control rods was disabled and all gas circulators were tripped to reduce the coolant flow rate to zero. The analytical results could show that the downstream of forced convection caused by the HPS pushes down the upstream by natural convection in the fuel assemblies; however, the forced convection has little influence on the core thermal-hydraulics without the reactor outlet coolant temperature. As a result, the three-dimensional thermal-phenomena inside the RPV during the LOFC test could be understood qualitatively.

Applicability study of nuclear graphite material IG-430 to VHTR

Osaki, Hirotaka; Shimazaki, Yosuke; Sumita, Junya; Shibata, Taiju; Konishi, Takashi; Ishihara, Masahiro

Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 8 Pages, 2015/05

For the design on the VHTR graphite components, it is desirable to employ graphite material with higher strength. IG-430 graphite has been developed as an advanced candidate for VHTR. However, the new developed IG-430 does not have enough databases for the design of HTGR. In this paper, the compressive strength (Cs) of IG-430, one of important strengths for design data, is statistically evaluated. The component reliability is evaluated based on the safety factors defined by the graphite design code, and the applicability as the VHTR graphite material is discussed. It was found that IG-430 has higher strength (about 11%) and lower standard deviation (about 27%) than IG-110 which is one of traditional graphites used for HTGR, because the crack in IG-430 would not easy to propagate rather than IG-110. Since fracture probability for IG-430 is low, the higher reliability of core-component will be achieved using IG-430. It is expected that IG-430 is applicable for VHTR graphite material.

Improvement of core dynamics analysis of control rod withdrawal test in HTGR

Takamatsu, Kuniyoshi; Nakagawa, Shigeaki

Nihon Genshiryoku Gakkai Wabun Rombunshi, 5(1), p.45 - 56, 2006/03

http://doi.org/10.3327/taesj2002.5.45

The HTTR (High Temperature Engineering Test Reactor), which has thermal output of 30MW, coolant inlet temperature of 395C and coolant outlet temperature of 850C/950C, is a first high temperature gas-cooled reactor (HTGR) in Japan. The HTGR has a high inherent safety potential to accident condition. Safety demonstration tests using the HTTR are underway in order to demonstrate such excellent inherent safety features of the HTGR. A one-point core dynamics approximation with one fuel channel model had applied to this analysis. It was found that the analytical model for core dynamics couldn't simulate the reactor power behavior accurately. This report proposes an original method using temperature coefficients of some regions in the core. It is crucial to evaluate this method precisely to simulate a performance of HTGR during the test.

Anisotropic deformation effect on the fracture of core components made of two-dimensional C/C composite

Sumita, Junya; Shibata, Taiju; Ishihara, Masahiro; Iyoku, Tatsuo; Tsuji, Nobumasa*

Key Engineering Materials, 297-300, p.143 - 147, 2005/11

no abstracts in English

Helium chemistry in high-temperature gas-cooled reactors; Chemistry control for avoiding Hastelloy XR corrosion in the HTTR-IS system

Sakaba, Nariaki; Hirayama, Yoshiaki*

Proceedings of International Conference on Nuclear Energy System for Future Generation and Global Sustainability (GLOBAL 2005) (CD-ROM), 6 Pages, 2005/10

The high-temperature gas-cooled reactor (HTGR) is capable of producing a massive quantity of hydrogen with no carbon dioxide emission during its production by a thermo chemical IS (Iodine-Sulphur) process. The HTTR (High Temperature Engineering Test Reactor), which is the first high-temperature gas-cooled reactor in Japan, will be connected to some heat utilization system in the near future. The thermo chemical IS process is one of the progressive candidates. The metallic material of the heat transfer tube of the intermediate heat exchanger (IHX) and liner in the concentric hot gas duct in the HTTR-IS system, which allows usage in high-temperature conditions, is the nickel-based high-temperature alloy Hastelloy XR. Since the coolant helium contains small amounts of impurities, it is necessary to control the chemical composition in order to minimize corrosion of the Hastelloy XR. Major corrosion phenomena of the Hastelloy XR are carburization, decarburization, oxidation, and carbon deposition depending upon the particular gas composition and its temperature. The carburization and decarburization phenomena can be restricted by controlling the carbon activity and oxygen partial pressure. This paper describes the effect of each coolant impurity for the carburization and decarburization. Also a chemical composition limit was evaluated to avoid the Hastelloy XR corrosion.

JAERI's hot stuff

Sakaba, Nariaki; Tachibana, Yukio; Onuki, Kaoru; Komori, Yoshihiro; Ogawa, Masuro

Nuclear Engineering International, 50(612), p.20 - 22, 2005/07

The HTTR (High Temperature Engineering Test Reactor) at Japan Atomic Energy Research Institute's Oarai Research Establishment attained its maximum reactor-outlet coolant temperature of 950C in April 2004 and ready to connect nuclear heat for industrial applications. The hydrogen production system by thermochemical water-splitting Iodine Sulphur cycle is also developing and succeeded to produce 30 normal L/h hydrogen in a closed cycle in June 2004.

Heat exchanger performance in main cooling system on high temperature test operation at High Temperature Gas-Cooled Reactor 'HTTR'

Tochio, Daisuke; Nakagawa, Shigeaki; Furusawa, Takayuki*

Nihon Genshiryoku Gakkai Wabun Rombunshi, 4(2), p.147 - 155, 2005/06

https://doi.org/10.3327/taesj2002.4.147

High Temperature Engineering Test Reactor (HTTR) of high temperature gas-cooled reactor at JAERI achieved the reactor outlet coolant temperature of 950C for the first time in the world at Apr. 19, 2004. To remove of generated heat at reactor core and to hold reactor inlet coolant temperature as specified temperature, heat exchangers in HTTR main cooling system should have designed heat exchange performance. In this report, heat exchanger performance is evaluated based on measurement data in high temperature test operation. And it is confirmed the adequacy of heat exchanger designing method by comparison of evaluated value with designed value.

Coolant chemistry of the High Temperature Gas-Cooled Reactor 'HTTR'

Sakaba, Nariaki; Nakagawa, Shigeaki; Furusawa, Takayuki*; Emori, Koichi; Tachibana, Yukio

Nihon Genshiryoku Gakkai Wabun Rombunshi, 3(4), p.388 - 395, 2004/12

https://doi.org/10.3327/taesj2002.3.388

Chemistry control is important for the helium coolant of High Temperature Gas-cooled Reactors (HTGRs) because impurities cause oxidation of the graphite used in the core and corrosion of high temperature materials used in the heat exchanger. In the High Temperature Engineering Test Reactor (HTTR) which is the first HTGR in Japan, the chemical impurity concentration is restricted and its behaviour is monitored during reactor operations. The impurity is reduced by the helium purification system and the concentration is measured by the helium sampling system installed to the primary and secondary helium system, continuously. This paper describes the impurity behaviour during the rise-to-power test which is the initial power-up of the HTTR. Also, the amount of the emitted impurity to the primary circuit from the graphite component and insulator used at the concentric hot gas duct are evaluated. During the power up, any abnormal impurity increases were not obtained and the chemical composition of the primary circuit is sufficiently in the stability area to avoid carbon deposition.

Safety demonstration tests using high temperature engineering test reactor

Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Tachibana, Yukio; Sakaba, Nariaki; Iyoku, Tatsuo

Nuclear Engineering and Design, 233(1-3), p.301 - 308, 2004/10

https://doi.org/10.1016/j.nucengdes.2004.08.016

Safety demonstration tests using the High Temperature Engineering Test Reactor (HTTR) are conducted for demonstrating inherent safety features of High Temperature Gas-cooled Reactors (HTGRs) as well as for providing core and plant transient data for validation of HTGR safety analysis codes. The safety demonstration tests are divided to the first phase and second phase tests. In the first phase tests, simulation tests of anticipated operational occurrences and anticipated transients without scram (ATWS) are conducted. The second phase tests will simulate accidents such as a depressurization accident (loss of coolant accident). The first phase tests simulating reactivity insertion events and coolant flow reduction events started in FY 2002. The first phase safety demonstration tests will continue until FY 2005, and the second phase tests will be carried out from FY 2006.

Structural integrity assessment of helium component during safety demonstration test using HTTR, 1 (Contract Research)

Sakaba, Nariaki; Nakagawa, Shigeaki; Furusawa, Takayuki; Tachibana, Yukio

JAERI-Tech 2004-045, 67 Pages, 2004/04

JAERI-Tech-2004-045.pdf:4.74MB

Safety demonstration tests using the HTTR are now underway in order to verify the inherent safety features and to improve the safety design and evaluation technologies for HTGRs, as well as to contribute to research and development for the VHTR, which is one of the Generation IV reactors. In the safety demonstration tests, the coolant flow reduction test by tripping one or two out of three gas circulators is being performed between FY2002 and FY 2005 and by tripping all the three gas circulators will be conducted after FY2006. This paper describes the structural integrity assessment of the primary pressurised water cooler after one and two gas circulators run down. Also, the possibility of natural convection in the primary coolant after all the three gas circulator stopped was evaluated by the operation data of the reactor-scram test performed during the rise-to-power tests.