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Development of a DDA+PGA-combined non-destructive active interrogation system in "Active-N"

古高 和禎; 大図 章; 藤 暢輔

Nuclear Engineering and Technology, 55(11), p.4002 - 4018, 2023/11

An integrated neutron interrogation system has been developed for non-destructive assay of highly radioactive special nuclear materials, to accumulate knowledge of the method through developing and using it. The system combines a differential die-away (DDA) measurement system for the quantification of nuclear materials and a prompt gamma-ray analysis (PGA) system for the detection of neutron poisons which disturb the DDA measurements; a common D-T neutron generator is used. A special care has been taken for the selection of materials to reduce the background gamma rays produced by the interrogation neutrons. A series of measurements were performed to test the basic performance of the system. The results show that the DDA system can quantify plutonium of as small as 20~mg and it is not affected by intense neutron background up to 4.2~TBq and gamma ray of 2.2~TBq. As a result of the designing of the combined system as a whole, the gamma-ray background counting rate at the PGA detector was reduced down to $$3.9times10^{3}$$ s$$^{-1}$$ even with the use of the D-T neutron generator. The test measurements show that the PGA system is capable of detecting less than 1~g of boron compound and about 100~g of gadolinium compound in~30 min. This research was implemented under the subsidy for nuclear security promotion of MEXT.


Large-eddy simulation on gas mixing induced by the high-buoyancy flow in the CIGMA facility

安部 諭; 柴本 泰照

Nuclear Engineering and Technology, 55(5), p.1742 - 1756, 2023/05

 被引用回数:0 パーセンタイル:0.02(Nuclear Science & Technology)

The hydrogen behavior in a nuclear containment vessel is a significant issue when discussing the potential of hydrogen combustion during a severe accident. After the Fukushima-Daiichi accident in Japan, we have investigated in-depth the hydrogen transport mechanisms by utilizing experimental and numerical approaches. Computational fluid dynamics is a powerful tool for better understanding the transport behavior of gas mixtures, including hydrogen. This paper describes a large-eddy simulation of gas mixing driven by a high-buoyancy flow. We focused on the interaction behavior of heat and mass transfers driven by the horizontal high-buoyant flow during density stratification. For validation, the experimental data of the Containment InteGral effects Measurement Apparatus (CIGMA) facility were used. With a high-power heater for the gas-injection line in the CIGMA facility, a high temperature flow of approximately 390$$^{circ}$$C was injected into the test vessel. By using the CIGMA facility, we can extend the experimental data to the high temperature region. The phenomenological discussion in this paper help understand the heat and mass transfer induced by the high-buoyancy flow in the containment vessel during a severe accident.


Phase analysis of simulated nuclear fuel debris synthesized using UO$$_{2}$$, Zr, and stainless steel and leaching behavior of the fission products and matrix elements

頓名 龍太郎*; 佐々木 隆之*; 児玉 雄二*; 小林 大志*; 秋山 大輔*; 桐島 陽*; 佐藤 修彰*; 熊谷 友多; 日下 良二; 渡邉 雅之

Nuclear Engineering and Technology, 55(4), p.1300 - 1309, 2023/04

 被引用回数:0 パーセンタイル:0.02(Nuclear Science & Technology)

UO$$_{2}$$・Zr・ステンレス鋼を出発物質として模擬デブリを合成し、形成された固相の分析と浸漬試験を行った。主要なU含有相は合成条件に依存し、不活性雰囲気下・1473KではUO$$_{2}$$相が維持されていた。1873Kでは(U,Zr)O$$_{2}$$固溶体相の形成が観測された。酸化性雰囲気では、1473Kの場合にはU$$_{3}$$O$$_{8}$$と(Fe,Cr)UO$$_{4}$$相の混合物が得られ、1873Kでは(U,Zr)O$$_{2}$$が形成された。浸漬試験により金属イオンの溶出挙動を調べるため、中性子照射により核分裂生成物を導入する、もしくは出発物質への添加によりその安定同位体を導入する処理を行った。試験の結果、Uの溶出挙動は、模擬デブリの性状や浸漬液の液性に依存することが確認された。CsやSr, Baは模擬デブリの固相組成に依存せず顕著な溶出を示した。一方で、多価イオンとなるEuとRuの溶出は抑制されることが観測され、模擬デブリ中でウラン相に固溶ないしは包含されたことによる影響が推察される。


Raman spectroscopy of eutectic melting between boride granule and stainless steel for sodium-cooled fast reactors

深井 尋史*; 古谷 正裕*; 山野 秀将

Nuclear Engineering and Technology, 55(3), p.902 - 907, 2023/03

 被引用回数:0 パーセンタイル:0.02(Nuclear Science & Technology)

本論文は、炭化ホウ素(B$$_{4}$$C)とステンレス鋼(SS)の共晶溶融・固化反応に関する反応生成物及びその分布を扱う。B$$_{4}$$C-SS共晶反応への炭素の存在の影響を調べるため、ホウ化鉄(FeB)とSSの反応を比較して、多変量スペクトル解析を用いたラマン分光分析を実施した。走査電子顕微鏡とエネルギー分散型X線分析も実施し、Cr, Ni, Feのような純金属の要素情報を調べた。B$$_{4}$$C-SS試料では、界面層に非結晶カーボンやFeB, Fe$$_{2}$$Bが見られた。それに対して、FeB-SS試料では、界面にはそのような界面層が見られなかった。


Study on the effect of long-term high temperature irradiation on TRISO fuel

Shaimerdenov, A.*; Gizatulin, S.*; Dyussambayev, D.*; Askerbekov, S.*; 植田 祥平; 相原 純; 柴田 大受; 坂場 成昭

Nuclear Engineering and Technology, 54(8), p.2792 - 2800, 2022/08

 被引用回数:4 パーセンタイル:85.55(Nuclear Science & Technology)

In the core of the WWR-K reactor, a long-term irradiation of tri-structural isotopic (TRISO)-coated fuel particles (CFPs) with a UO$$_{2}$$ kernel was carried out under normal operating conditions of the high-temperature gas-cooled reactor (HTGR). This TRISO fuel was attained at the temperature of 950 to 1,100 $$^{circ}$$C, and the uranium burnup of 9.9% FIMA (fission per initial metal atom) during the irradiation. The release of the gaseous fission product from the fuel was measured in-pile, and its release-to-birth (R/B) ratio was evaluated using the model developed in the High-Temperature Engineering Test Reactor (HTTR) project. After the irradiation test, fuel compacts were subjected to electric dissociation and nondestructive inspections such as X-ray radiography and gamma spectrometry. Finally, it was concluded that integrity of the TRISO fuel irradiated at approximately 9.9% FIMA was demonstrated, and a low fuel failure fraction and a low R/B measured with krypton-88 indicated good performance and reliability of the high burnup TRISO fuel.


Development and validation of fuel stub motion model for the disrupted core of a sodium-cooled fast reactor

川田 賢一; 鈴木 徹*

Nuclear Engineering and Technology, 53(12), p.3930 - 3943, 2021/12

 被引用回数:0 パーセンタイル:0.01(Nuclear Science & Technology)



Uncertainty analysis of ROSA/LSTF test by RELAP5 code and PKL counterpart test concerning PWR hot leg break LOCAs

竹田 武司; 大津 巌

Nuclear Engineering and Technology, 50(6), p.829 - 841, 2018/08

 被引用回数:13 パーセンタイル:81.28(Nuclear Science & Technology)

An experiment was conducted for OECD/NEA ROSA-2 Project using LSTF, which simulated 17% hot leg intermediate-break LOCA in PWR. Core uncovery started simultaneously with liquid level drop in crossover leg downflow-side before loop seal clearing, and water remaining occurred on upper core plate. Results of uncertainty analysis with RELAP5/MOD3.3 code clarified influences of combination of multiple uncertain parameters on peak cladding temperature within defined uncertain ranges. An experiment was performed for OECD/NEA PKL-3 Project with PKL. The LSTF test simulated PWR 1% hot leg small-break LOCA with steam generator secondary-side depressurization as accident management measure and nitrogen gas inflow. Some discrepancies appeared between the LSTF and PKL test results for primary pressure, core collapsed liquid level, and cladding surface temperature probably due to effects of differences between LSTF and PKL in configuration, geometry, and volumetric size.


ROSA/LSTF test and RELAP5 analyses on PWR cold leg small-break LOCA with accident management measure and PKL counterpart test

竹田 武司; 大津 巌

Nuclear Engineering and Technology, 49(5), p.928 - 940, 2017/08

 被引用回数:4 パーセンタイル:38.41(Nuclear Science & Technology)

An experiment using PKL was performed for the OECD/NEA PKL-3 Project as a counterpart to a previous test with LSTF on a cold leg small-break loss-of-coolant accident with an accident management measure in a PWR. The rate of steam generator secondary-side depressurization was controlled to achieve a primary depressurization rate of 200 K/h as a common test condition. In both tests, rapid recovery started in the core collapsed liquid level after loop seal clearing. Some discrepancies appeared between the LSTF and PKL test results for the core collapsed liquid level, the cladding surface temperature, and the primary pressure. The RELAP5/MOD3.3 code indicated a remaining problem in the prediction of primary coolant distribution. Results of uncertainty analysis for the LSTF test clarified the influences of the combination of multiple uncertain parameters on peak cladding temperature within the defined uncertain ranges.


Bayesian optimization analysis of containment-venting operation in a Boiling Water Reactor severe accident

Zheng, X.; 石川 淳; 杉山 智之; 丸山 結

Nuclear Engineering and Technology, 49(2), p.434 - 441, 2017/03

 被引用回数:4 パーセンタイル:38.41(Nuclear Science & Technology)

Containment venting is one of essential measures to protect the integrity of the final barrier of a nuclear reactor during severe accidents, by which the uncontrollable release of fission products can be avoided. The authors seek to develop an optimization approach, from a simulation-based perspective, to the venting operations by using an integrated severe accident code, THALES2/KICHE. The effectiveness of containment venting strategies needs to be verified via numerical simulations based on various settings of venting conditions. The number of iterations, however, needs to be controlled for cumbersome computational burden of integrated codes. Bayesian optimization is an efficient global optimization approach. By using Gaussian process regression, a surrogate model of the "black-box" code is constructed. It can be updated simultaneously whenever new simulation results are acquired. With predictions via the surrogate model, upcoming locations of the most probable optimum can be revealed. The sampling procedure is adaptive. The number of code queries is largely reduced for the optimum finding, compared with pure random searches. One typical severe accident scenario of a boiling water reactor is chosen as an example. The research demonstrates the applicability of the Bayesian optimization approach to the design and establishment of containment-venting strategies during severe accidents.


A Preliminary evaluation of unprotected loss-of-flow accident for a prototype fast-breeder reactor

鈴木 徹; 飛田 吉春; 川田 賢一; 田上 浩孝; 曽我部 丞司; 松場 賢一; 伊藤 啓; 大島 宏之

Nuclear Engineering and Technology, 47(3), p.240 - 252, 2015/04

 被引用回数:25 パーセンタイル:90.56(Nuclear Science & Technology)

In the original licensing application for the prototype fast-breeder reactor, MONJU, the event progression during an unprotected loss-of-flow (ULOF), which is one of the technically inconceivable events postulated beyond design basis, was evaluated. Through this evaluation, it was confirmed that radiological consequences could be suitably limited even if mechanical energy was released. Following the Fukushima-Daiichi accident, a new nuclear safety regulation has become effective in Japan. The conformity of MONJU to this new regulation should hence be investigated. The objectives of the present study are to conduct a preliminary evaluation of ULOF for MONJU, reflecting the knowledge obtained after the original licensing application through CABRI experiments and EAGLE projects, and to gain the prospect of In-Vessel Retention (IVR) for the conformity of MONJU to the new regulation. The preliminary evaluation in the present study showed that no significant mechanical energy release would take place, and that thermal failure of the reactor vessel could be avoided by the stable cooling of disrupted-core materials. This result suggests that the prospect of IVR against ULOF, which lies within the bounds of the original licensing evaluation and conforms to the new nuclear safety regulation, will be gained.


Characteristics of self-leveling behavior of debris beds in a series of experiments

Cheng, S.; 山野 秀将; 鈴木 徹; 飛田 吉春; 中村 裕也*; Zhang, B.*; 松元 達也*; 守田 幸路*

Nuclear Engineering and Technology, 45(3), p.323 - 334, 2013/06

 被引用回数:35 パーセンタイル:93.35(Nuclear Science & Technology)

During a hypothetical core-disruptive accident (CDA) in a sodium-cooled fast reactor (SFR), degraded core materials can form conically-shaped debris beds over the core-support structure and/or in the lower inlet plenum of the reactor vessel from rapid quenching and fragmentation of core material pool. However, coolant boiling may lead ultimately to leveling of the debris bed that is crucial to the relocation of molten core and heat-removal capability of debris bed. To clarify the mechanisms underlying this self-leveling behavior, a great amount of experiments were performed within a variety of conditions in recent years under the constructive collaboration between Japan Atomic Energy Agency (JAEA) and Kyushu University (Japan). The present contribution synthesizes and gives detailed comparative analyses of those experiments. Effects of various experimental parameters that may have potential influence on the leveling process such as boiling mode (bottom-heated, depressurization boiling and gas injection), particle size, particle density, particle shape (spherical and non-spherical), boiling intensity (or gas flow rate), water depth along with column geometry, were investigated, thus, giving a large palette of favorable data for better understanding of CDAs and improved verifications of computer models developed in advanced fast reactor safety analysis codes.


International collaboration in assessment of radiological impacts arising from releases to the biosphere after disposal of radioactive waste into geological repositories

Smith, G.*; 加藤 智子

Nuclear Engineering and Technology, 42(1), p.1 - 8, 2010/02



A Study of hydraulic properties in a single fracture with in-plane heterogeneity; An Evaluation using optical measurements of a transparent replica

澤田 淳; 佐藤 久

Nuclear Engineering and Technology, 42(1), p.9 - 16, 2010/02



High temperature oxidation of Nb-containing Zr alloy cladding in LOCA conditions

中頭 利則; 永瀬 文久; 更田 豊志

Nuclear Engineering and Technology, 41(2), p.163 - 170, 2009/03



The Impact of fuel cycle options on the space requirements of a HLW repository

河田 東海夫

Nuclear Engineering and Technology, 39(6), p.683 - 690, 2007/12



A Next generation sodium-cooled fast reactor concept and its R&D program

一宮 正和; 水野 朋保; 小竹 庄司

Nuclear Engineering and Technology, 39(3), p.171 - 186, 2007/06



JAEA's VHTR for Hydrogen and Electricity Cogeneration; GTHTR300C

國富 一彦; Yan, X.; 西原 哲夫; 坂場 成昭; 毛利 智聡

Nuclear Engineering and Technology, 39(1), p.9 - 20, 2007/02



Key R&D activities supporting disposal of radioactive waste; Responding to the challenges of the 21st century

宮本 陽一; 梅木 博之; 大澤 英昭; 内藤 守正; 中野 勝志; 牧野 仁史; 清水 和彦; 瀬尾 俊弘

Nuclear Engineering and Technology, 38(6), p.505 - 534, 2006/08



Current status of thermal/hydraulic feasibility project for reduced-moderation water reactor, 2; Development of two-phase flow simulation code with advanced interface tracking method

吉田 啓之; 玉井 秀定; 大貫 晃; 高瀬 和之; 秋本 肇

Nuclear Engineering and Technology, 38(2), p.119 - 128, 2006/04



Safety studies on hydrogen production system with a high temperature gas-cooled reactor

武田 哲明

Nuclear Engineering and Technology, 37(6), p.537 - 556, 2005/12



Uncertainty and sensitivity studies with the probabilistic accident consequence assessment code OSCAAR

本間 俊充; 富田 賢一*; 波戸 真治*

Nuclear Engineering and Technology, 37(3), p.245 - 258, 2005/06


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