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Wozniak, N.*; Ohgama, Kazuya; Doda, Norihiro; Ota, Hirokazu*; Shemon, E.*; Feng, B.*; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; et al.
Progress in Nuclear Science and Technology (Internet), 8, p.165 - 170, 2025/09
To enhance the accuracy of the safety evaluations for sodium-cooled fast reactors, it is necessary to develop a method that can realistically evaluate the reactivity changes induced by core deformation. In this context, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments using multiple ducts from a Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and experimental results demonstrated that the codes used by both countries were able to reasonably predict the thermal bowing of a row of assemblies in multiple duct configuration.
Ohgama, Kazuya; Doda, Norihiro; Ota, Hirokazu*; Wozniak, N.*; Uwaba, Tomoyuki; Futagami, Satoshi; Tanaka, Masaaki; Yamano, Hidemasa; Ogata, Takanari*; Shemon, E.*; et al.
Progress in Nuclear Science and Technology (Internet), 8, p.160 - 164, 2025/09
To enhance the accuracy of the safety evaluations for sodium-cooled fast reactors, it is necessary to develop a method that can realistically evaluate the reactivity changes induced by core deformation. In this context, Japan and the United States jointly conducted a benchmark analysis of thermal bowing experiments using a single duct from a Joyo-type fuel assembly. The aim was to confirm the validity of the core bowing analysis codes. Comparisons of analysis and experimental results demonstrated that the codes used by both countries were able to reasonably predict the axial distribution of horizontal duct displacement caused by thermal bowing as well as the contact load on the duct pad.
Nagasumi, Satoru; Hasegawa, Toshinari; Nakagawa, Shigeaki; Kubo, Shinji; Iigaki, Kazuhiko; Shinohara, Masanori; Saikusa, Akio; Nojiri, Naoki; Saito, Kenji; Furusawa, Takayuki; et al.
JAEA-Research 2025-005, 23 Pages, 2025/07
JAEA-Research-2025-005.pdf:2.68MB
A safety demonstration test under abnormal operating conditions using the HTTR (High Temperature Engineering Test Reactor) was conducted to demonstrate safety features of the HTGRs (High Temperature Gas-cooled Reactors). Under a simulation of a control rod shutdown failure, all primary helium gas circulators were intentionally stopped during a steady-state operation at 100% reactor thermal power (30 MW), temporal changes of the reactor power and temperatures around the reactor pressure vessel (RPV) were obtained after the complete loss of forced heat removal from the reactor core. After the event (primary coolant flow stopped), the reactor power quickly decreased due to the negative reactivity feedback associated with the core temperature rise, and then the reactor power spontaneously shifted to a stable state of low power (about 1.2%) even after a recriticality. Heat dissipation from RPV surface to a surrounding vessel cooling system (water-cooled panels) ensured the amount of heat removal required to maintain the reactor temperature constant in the low power state. In this way, the transition from the event occurrence to the stable and safety state, i.e., inherent safety features of HTGRs, were demonstrated in the case of core forced cooling loss without active shutdown operations.
Ota, Hirokazu*; Ogata, Takanari*; Yamano, Hidemasa; Futagami, Satoshi; Shimada, Sadae*; Yamada, Yumi*
Proceedings of 30th International Conference on Nuclear Engineering (ICONE30) (Internet), 8 Pages, 2023/05
Yoshimura, Kazuo; Doda, Norihiro; Igawa, Kenichi*; Tanaka, Masaaki; Yamano, Hidemasa
Journal of Nuclear Engineering and Radiation Science, 9(2), p.021601_1 - 021601_9, 2023/04
Feedback reactivity automatically caused by radial expansion of the core is known as one of the inherent safety features in a sodium-cooled fast reactor (SFR). In order to validate the evaluation models of the reactivity feedback equipped in the in-house plant dynamics analysis code named Super-COPD, the benchmark analyses for the unprotected loss of heat sink (ULOHS) tests of BOP-302R and BOP-301 in an experimental SFR, EBR-II were conducted and the applicability of the evaluation method for the reactivity feedback was indicated during the ULOHS even, by comparing the numerical results and the experimental data.
Kitano, Akihiro; Nakajima, Ken*
Proceedings of 2018 International Congress on Advances in Nuclear Power Plants (ICAPP 2018) (CD-ROM), p.1205 - 1210, 2018/04
The feedback reactivity is taken into account in fast reactor core design especially in order to make the power coefficient negative, which is required to be confirmed in the operation. In the feedback reactivity experiment, the positive reactivity was inserted in the critical state at zero power, and the thermal data, such as reactor power and the R/V inlet temperature, was acquired until the power got stable by the feedback reactivity. In the conventional study, only two critical points in an experiment are available for evaluation of the feedback reactivity coefficients. This method needs three days for evaluation. The advanced method based on the inverse kinetics is newly applied in this work using the more extensive data. It is confirmed that this approach can evaluate the feedback reactivity coefficients in one experiment.
Takamatsu, Kuniyoshi; Nakagawa, Shigeaki
Nihon Genshiryoku Gakkai Wabun Rombunshi, 5(1), p.45 - 56, 2006/03
The HTTR (High Temperature Engineering Test Reactor), which has thermal output of 30MW, coolant inlet temperature of 395
C 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.
Iguchi, Tadashi; Shibamoto, Yasuteru; Asaka, Hideaki; Nakamura, Hideo
Proceedings of 10th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-10), 16 Pages, 2003/10
Thermal-hydraulic and neutronic dynamics are always interrelated in BWR core. This is called thermal-hydraulic and neutronic (T/N) coupling. Channel stability experiments with T/N coupling under non-nuclear condition are very limited. This is mainly due to the difficulties in the real-time simulation of neutron dynamics and in the fast-response void fraction measurement under high-pressure and temperature conditions. Authors have developed techniques to solve the above difficulties, and have succeeded in experimentally simulating T/N coupling under non-nuclear conditions with the THYNC facility. Using THYNC facility, T/N coupling effect on channel stability was investigated. Experiments were performed under Pressure=2-7MPa, Subcooling=10-40K, and Mass flux=270-660kg/m
s. THYNC results indicated T/N coupling lowered the channel stability threshold. The reduction of channel stability threshold due to T/N coupling was small within 10% at 7MPa in the present THYNC experiment, although the experimental condition was set to be more severe than that supposed in a reactor.
Nakajima, Ken; ;
Proceedings of 6th International Conference on Nuclear Criticality Safety (ICNC '99), 3, p.1286 - 1292, 1999/00
no abstracts in English
optimal filter with a nonlinear termSuzuki, Katsuo;
Nuclear Technology, 113, p.145 - 154, 1996/02
Times Cited Count：4 Percentile：38.42(Nuclear Science & Technology)no abstracts in English
;
JAERI-M 85-212, 56 Pages, 1986/01
no abstracts in English
