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Journal Articles

A 3D particle-based analysis of molten pool-to-structural wall heat transfer in a simulated fuel subassembly

Zhang, T.*; Morita, Koji*; Liu, X.*; Liu, W.*; Kamiyama, Kenji

Extended abstracts of the 2nd Asian Conference on Thermal Sciences (Internet), 2 Pages, 2021/10

For the Japanese sodium cooled fast reactor, a fuel subassembly with an inner duct structure (FAIDUS) was designed to avoid the re-criticality by preventing the large-scale pool formation. In the present study, using the finite volume particle method, the EAGLE ID1 test which was an in-pile test performed to demonstrate the effectiveness of FAIDUS was numerically simulated and the thermal-hydraulic mechanisms underlying the heat transfer process were analyzed.

Journal Articles

Water experiments on thermal striping phenomena at the core outlet of an advanced sodium-cooled fast reactor, 1; Proposal of countermeasures to mitigate temperature fluctuations around control rods

Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Kurihara, Akikazu; Tanaka, Masaaki

Hozengaku, 20(3), p.89 - 96, 2021/10

Hot sodium from the fuel assembly can mix with cold sodium from the control rod (CR) channel and the blanket assemblies at the bottom plate of the Upper Internal Structure (UIS) of Advanced-SFR. Temperature fluctuation due to mixing of the fluids at different temperature between the core outlet and cold channel may cause high cycle thermal fatigue on the structure around the bottom of UIS. A water experiment using a 1/3 scale 60 degree sector model simulating the upper plenum of the Advanced-SFR has been conducted to examine countermeasures for the significant temperature fluctuation generated around the bottom of UIS. We focused on the temperature fluctuations near the primary and backup control rod channels, and studied the countermeasure structure to mitigate the temperature fluctuation through temperature distribution and flow velocity distribution measurements. As a result, effectiveness of the countermeasure to mitigate the temperature fluctuation intensity was confirmed.

Journal Articles

Water experiments on thermal striping phenomena at the core outlet of an advanced sodium-cooled fast reactor, 2; Proposal of countermeasures to mitigate temperature fluctuations around radial blanket fuel assemblies

Kobayashi, Jun; Aizawa, Kosuke; Ezure, Toshiki; Kurihara, Akikazu; Tanaka, Masaaki

Hozengaku, 20(3), p.97 - 101, 2021/10

Focusing on the thermal striping phenomena that occurs at a bottom of the internal structure of an advanced sodium-cooled fast reactor (Advanced-SFR) that has been designed by the Japan Atomic Energy Agency, a water experiment using a 1/3 scale 60 degree sector model simulating the upper plenum of the Advanced-SFR has been conducted to examine countermeasures for the significant temperature fluctuation generated around the bottom of Upper Internal Structure (UIS). In the previous paper, we reported the effect of measures to mitigate temperature fluctuations around the control rod channels. In this paper, the same test section was used, and a water experiment was conducted to obtain the characteristics of temperature fluctuations around the radial blanket fuel assembly. And the shape of the Core Instrumentation Support Plate (CIP) was modified, and it was confirmed that it was highly effective in alleviating temperature fluctuations around the radial blanket fuel assembly.

Journal Articles

Modelling concrete degradation by coupled non-linear processes

Oda, Chie; Kawama, Daisuke*; Shimizu, Hiroyuki*; Benbow, S. J.*; Hirano, Fumio; Takayama, Yusuke; Takase, Hiroyasu*; Mihara, Morihiro; Honda, Akira

Journal of Advanced Concrete Technology, 19(10), p.1075 - 1087, 2021/10

Concrete in a transuranic (TRU) waste repository is considered a suitable material to ensure safety, provide structural integrity and retard radionuclide migration after the waste containers fail. In the current study, coupling between chemical, mass-transport and mechanical, so-called non-linear processes that control concrete degradation and crack development were investigated by coupled numerical models. Application of such coupled numerical models allows identification of the dominant non-linear processes that will control long-term concrete degradation and crack development in a TRU waste repository.

Journal Articles

Analysis and mapping of detailed inner information of crystalline grain by wavelength-resolved neutron transmission imaging with individual Bragg-dip profile-fitting analysis

Sakurai, Yosuke*; Sato, Hirotaka*; Adachi, Nozomu*; Morooka, Satoshi; Todaka, Yoshikazu*; Kamiyama, Takashi*

Applied Sciences (Internet), 11(11), p.5219_1 - 5219_17, 2021/06

 Times Cited Count:1 Percentile:76.99(Chemistry, Multidisciplinary)

Journal Articles

Incorporation of U, Pb and rare earth elements in calcite through crystallisation from amorphous calcium carbonate; Simple preparation of reference materials for microanalysis

Miyajima, Yusuke*; Saito, Ayaka*; Kagi, Hiroyuki*; Yokoyama, Tatsunori; Takahashi, Yoshio*; Hirata, Takafumi*

Geostandards and Geoanalytical Research, 45(1), p.189 - 205, 2021/03

 Times Cited Count:0 Percentile:0.01(Geochemistry & Geophysics)

Uncertainty for elemental and isotopic analyses of calcite by LA-ICP-MS is largely controlled by the homogeneity of the reference materials (RMs) used for normalization and validation. In order to produce calcite RMs with homogeneous elemental and isotopic compositions, we incorporated elements including U, Pb, and rare earth elements into calcite through heat- and pressure-induced crystallization from amorphous calcium carbonate that was precipitated from element-doped reagent solution. X-ray absorption spectra showed that U was present as U(VI) in the synthesized calcite, probably with a different local structure from that of aqueous uranyl ions. The uptake rate of U by our calcite was higher in comparison to synthetic calcite of previous studies. Variations of element mass fractions in the calcite were better than 12% 2RSD, mostly within 7%. The $$^{207}$$Pb/$$^{206}$$Pb ratio in the calcite showed $$<$$1% variations, while the $$^{238}$$U/$$^{206}$$Pb ratio showed 3-24% variations depending on element mass fractions. Using the synthetic calcite as primary RMs, we could date a natural calcite RM, WC-1, with analytical uncertainty as low as $$<$$3%. The method presented can be useful to produce calcite with controlled and homogeneous element mass fractions, and is a promising alternative to natural calcite RMs for U-Pb geochronology.

Journal Articles

High temperature gas-cooled reactors

Takeda, Tetsuaki*; Inagaki, Yoshiyuki; Aihara, Jun; Aoki, Takeshi; Fujiwara, Yusuke; Fukaya, Yuji; Goto, Minoru; Ho, H. Q.; Iigaki, Kazuhiko; Imai, Yoshiyuki; et al.

High Temperature Gas-Cooled Reactors; JSME Series in Thermal and Nuclear Power Generation, Vol.5, 464 Pages, 2021/02

As a general overview of the research and development of a High Temperature Gas-cooled Reactor (HTGR) in JAEA, this book describes the achievements by the High Temperature Engineering Test Reactor (HTTR) on the designs, key component technologies such as fuel, reactor internals, high temperature components, etc., and operational experience such as rise-to-power tests, high temperature operation at 950$$^{circ}$$C, safety demonstration tests, etc. In addition, based on the knowledge of the HTTR, the development of designs and component technologies such as high performance fuel, helium gas turbine and hydrogen production by IS process for commercial HTGRs are described. These results are very useful for the future development of HTGRs. This book is published as one of a series of technical books on fossil fuel and nuclear energy systems by the Power Energy Systems Division of the Japan Society of Mechanical Engineers.

Journal Articles

Numerical simulation of heat transfer behavior in EAGLE ID1 in-pile test using finite volume particle method

Zhang, T.*; Funakoshi, Kanji*; Liu, X.*; Liu, W.*; Morita, Koji*; Kamiyama, Kenji

Annals of Nuclear Energy, 150, p.107856_1 - 107856_10, 2021/01

 Times Cited Count:1 Percentile:81.22(Nuclear Science & Technology)

JAEA Reports

Decommissioning of the Uranium Enrichment Laboratory

Kokusen, Junya; Akasaka, Shingo*; Shimizu, Osamu; Kanazawa, Hiroyuki; Honda, Junichi; Harada, Katsuya; Okamoto, Hisato

JAEA-Technology 2020-011, 70 Pages, 2020/10


The Uranium Enrichment Laboratory in the Japan Atomic Energy Agency (JAEA) was constructed in 1972 for the purpose of uranium enrichment research. The smoke emitting accident on 1989 and the fire accident on 1997 had been happened in this facility. The research on uranium enrichment was completed in JFY1998. The decommissioning work was started including the transfer of the nuclear fuel material to the other facility in JFY2012. The decommissioning work was completed in JFY2019 which are consisting of removing the hood, dismantlement of wall and ceiling with contamination caused by fire accident. The releasing the controlled area was performed after the confirmation of any contamination is not remained in the target area. The radioactive waste was generated while decommissioning, burnable and non-flammable are 1.7t and 69.5t respectively. The Laboratory will be used as a general facility for cold experiments.

Journal Articles

Validation of analysis models on relocation behavior of molten core materials in sodium-cooled fast reactors based on the melt discharge experiment

Igarashi, Kai*; Onuki, Ryoji*; Sakai, Takaaki*; Kato, Shinya; Matsuba, Kenichi; Kamiyama, Kenji

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 6 Pages, 2020/08

Journal Articles

Development of ex-vessel phenomena analysis model for multi-scenario simulation system, spectra

Uchibori, Akihiro; Aoyagi, Mitsuhiro; Takata, Takashi; Ohshima, Hiroyuki

Proceedings of 2020 International Conference on Nuclear Engineering (ICONE 2020) (Internet), 7 Pages, 2020/08

The multi-scenario simulation system named SPECTRA has been developed for integrated analysis of in- and ex-vessel phenomena during a severe accident in sodium-cooled fast reactors. The base module computing ex-vessel compressible gas behavior by a lumped mass model and a sodium-concrete interaction module were verified through the basic analyses individually. A validity of the system including the base module and the individual physical module such as the sodium-concrete interaction module was confirmed through the analysis assuming sodium leakage from a reactor vessel and a primary cooling loop.

Journal Articles

Preliminary analysis of sodium experimental apparatus PLANDTL-2 for development of evaluation method for thermal-hydraulics in reactor vessel of sodium fast reactor under decay heat removal system operation condition

Ono, Ayako; Tanaka, Masaaki; Miyake, Yasuhiro*; Hamase, Erina; Ezure, Toshiki

Mechanical Engineering Journal (Internet), 7(3), p.19-00546_1 - 19-00546_11, 2020/06

Fully natural circulation decay heat removal systems (DHRSs) are to be adopted for sodium fast reactors, which is a passive safety feature without any electrical pumps. It is required to grasp the thermal-hydraulic phenomena in the reactor vessel and evaluate the coolability of the core under the natural circulation not only for the normal operating condition but also for severe accident conditions. In this paper, the numerical results of the preliminary analysis for the sodium experimental condition with the PLANDTL-2 are discussed to establish an appropriate numerical models for the reactor core including the gap region among the subassemblies and the DHX. From these preliminary analyses, the characteristics of the thermal-hydraulics behavior in the PLANDTL-2 to be focused are extracted.

Journal Articles

Advancement of elemental analytical model in LEAP-III code for tube failure propagation

Uchibori, Akihiro; Yanagisawa, Hideki*; Takata, Takashi; Li, J.*; Jang, S.*

Mechanical Engineering Journal (Internet), 7(3), p.19-00548_1 - 19-00548_11, 2020/06

Evaluation of occurrence possibility of tube failure propagation under sodium-water reaction accident is an important issue. In this study, a numerical analysis method to predict occurrence of failure propagation by overheating rupture was constructed to expand application range of an existing computer code. Applicability of the method was constructed through the numerical analysis of the experiment on water vapor discharging in liquid sodium. To improve the evaluation accuracy for the temperature distribution, a Lagrangian particle model for simulating reacting jet was also developed as an alternative method and its basic function was confirmed.

Journal Articles

A Conceptual design study of pool-type sodium-cooled fast reactor with enhanced anti-seismic capability

Kubo, Shigenobu; Chikazawa, Yoshitaka; Ohshima, Hiroyuki; Uchita, Masato*; Miyagawa, Takayuki*; Eto, Masao*; Suzuno, Tetsuji*; Matoba, Ichiyo*; Endo, Junji*; Watanabe, Osamu*; et al.

Mechanical Engineering Journal (Internet), 7(3), p.19-00489_1 - 19-00489_16, 2020/06

The authors are developing the design concept of pool-type sodium-cooled fast reactor (SFR) that addresses Japan's specific siting conditions such as earthquakes and meets safety design criteria (SDC) and safety design guidelines (SDGs) for Generation IV SFRs. The development of this concept will broaden not only options for reactor types in Japan but also the range and depth of international cooperation. A design concept of 1,500 MWt (650 MWe) class pool-type SFR was thought up by applying design technology obtained from the design of advanced loop-type SFR, named JSFR, equipped with safety measures that reflect results from the feasibility study on commercialized fast reactor cycle systems and fast reactor cycle technology development, improved maintainability and repairability, and lessons learned from the Fukushima Daiichi Nuclear Power Plants accident.

Journal Articles

Study on plutonium burner high temperature gas-cooled reactor in Japan; Introduction scenario, reactor safety and fabrication tests of the 3S-TRISO fuel

Ueta, Shohei; Mizuta, Naoki; Fukaya, Yuji; Goto, Minoru; Tachibana, Yukio; Honda, Masaki*; Saiki, Yohei*; Takahashi, Masashi*; Ohira, Koichi*; Nakano, Masaaki*; et al.

Nuclear Engineering and Design, 357, p.110419_1 - 110419_10, 2020/02

 Times Cited Count:1 Percentile:39.17(Nuclear Science & Technology)

The concept of a plutonium (Pu) burner HTGR is proposed to incarnate highly-effective Pu utilization by its inherent safety features. The security and safety fuel (3S-TRISO fuel) employs the coated fuel particle with a fuel kernel made of plutonium dioxide (PuO$$_{2}$$) and yttria stabilized zirconia (YSZ) as an inert matrix. This paper presents feasibility study of Pu burner HTGR and R&D on the 3S-TRISO fuel.

JAEA Reports

Code-B-2.5.2 for stress calculation for SiC-TRISO fuel particle

Aihara, Jun; Goto, Minoru; Ueta, Shohei; Tachibana, Yukio

JAEA-Data/Code 2019-018, 22 Pages, 2020/01


Concept of Pu-burner high temperature gas-cooled reactor (HTGR) was proposed for purpose of more safely reducing amount of recovered Pu. In Pu-burner HTGR concept, coated fuel particle (CFP), with ZrC coated yttria stabilized zirconia (YSZ) containing PuO$$_{2}$$ (PuO$$_{2}$$-YSZ) small particle and with tri-structural isotropic (TRISO) coating, is employed for very high burn-up and high nuclear proliferation resistance. ZrC layer is oxygen getter. On the other hand, we have developed Code-B-2.5.2 for prediction of pressure vessel failure probabilities of SiC-tri-isotropic (TRISO) coated fuel particles for HTGRs under operation by modification of an existing code, Code-B-2. The main purpose of modification is preparation of applying code for CFPs of Pu-burner HTGR. In this report, basic formulae are described.

JAEA Reports

Flow separation at inlet causing transition and intermittency in circular pipe flow

Ogawa, Masuro*

JAEA-Technology 2019-010, 22 Pages, 2019/07


Transition phenomena from laminar to turbulent flow are roughly classified into three categories. Circular pipe flow of the third category is linearly stable against any small disturbance, despite that flow actually transitions and transitional flow exhibits intermittency. These are among major challenges that are yet to be resolved in fluid dynamics. Thus, author proposes hypothesis as follows; "Flow in a circular pipe transitions from laminar flow because of vortices released from separation bubble forming in vicinity of inlet of pipe, and transitional flow becomes intermittent because vortex-shedding is intermittent." Present hypothesis can easily explain why linear stability theory has not been able to predict transition in circular pipe flow, why circular pipe flow actually transitions, why transitional flow actually exhibits intermittency even due to small disturbance, and why numerical analysis has not been able to predict intermittency of transitional flow in circular pipe.

Journal Articles

Study on analysis method for inert gas behavior in liquid metal flow with considering dissolution and entrainment at free surface

Matsushita, Kentaro; Ito, Kei*; Ezure, Toshiki; Tanaka, Masaaki

Dai-24-Kai Doryoku, Enerugi Gijutsu Shimpojiumu Koen Rombunshu (USB Flash Drive), 5 Pages, 2019/06

In the design study on a sodium-cooled fast reactor (SFR), a numerical simulation code named SYRENA has been developed in Japan Atomic Energy Agency to analyze the behavior of gas bubbles and/or dissolved gas in the primary coolant system. In the present study, the effect of the non-condensable gas entrainment at the free surface on the bubble and the dissolved gas behavior in the primary coolant system were investigated for a typical pool type reactor, and also effect of a dipped-plate (D/P) installed below the free surface in the reactor vessel to suppress the gas bubble entrainment into the primary coolant system was especially investigated. It was clarified that the D/P was influential to the non-condensable gas behavior and the molar flow rate of gas bubbles in the primary coolant system varies depending on the relationship between the gas entrainment rate at the free surface and the exchange flow rate through the D/P.

Journal Articles

Visualizing an ignition process of hydrogen jets containing sodium mist by high-speed imaging

Doi, Daisuke; Seino, Hiroshi; Miyahara, Shinya*; Uno, Masayoshi*

Journal of Nuclear Science and Technology, 56(6), p.521 - 532, 2019/06

 Times Cited Count:1 Percentile:23.13(Nuclear Science & Technology)

Journal Articles

Routing study of above core structure with mock-up experiment for ASTRID

Takano, Kazuya; Sakamoto, Yoshihiko; Morohoshi, Kyoichi*; Okazaki, Hitoshi*; Gima, Hiromichi*; Teramae, Takuma*; Ikarimoto, Iwao*; Botte, F.*; Dirat, J.-F.*; Dechelette, F.*

Proceedings of 2019 International Congress on Advances in Nuclear Power Plants (ICAPP 2019) (Internet), 8 Pages, 2019/05

ASTRID has the objective to integrate innovative options in order to prepare the 4th generation reactors. In ASTRID, large number of tubes are installed above each fuel subassembly to monitor the core. These instrumentations such as thermocouples (TC) and Failed Fuel Detection and Location (FFDL) systems are integrated into Above Core Structure (ACS) with various sizes tubes. In the present study, the routing study for TC tubes and FFDL tubes was performed with 3D modeling and mock-up experiment of the ACS designed for ASTRID with 1500 MW thermal power in order to clarify the integration process and secure the design hypotheses. Although some problems on fabricability were found in the mock-up experiment, the possible solutions were proposed. The present study gives manufacturing feedback to design team and will contribute to increase the knowledge for ACS design and fabricability.

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