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相馬 秀; 石垣 将宏*; 安部 諭; 柴本 泰照
Nuclear Engineering and Technology, 56(7), p.2524 - 2533, 2024/07
被引用回数：3 パーセンタイル：75.80(Nuclear Science & Technology)When analyzing containment thermal-hydraulics, computational fluid dynamics (CFD) is a powerful tool because multi-dimensional and local analysis is required for some accident scenarios. According to the previous study, neglecting steam bulk condensation in the CFD analysis leads to a significant error in boundary layer profiles. Validating the condensation model requires the experimental data near the condensing surface, however, available boundary layer data is quite limited. It is also important to confirm whether the heat and mass transfer analogy (HMTA) is still valid in the presence of bulk condensation. In this study, the boundary layer measurements on the vertical condensing surface in the presence of air were performed with the rectangular channel facility WINCS, which was designed to measure the velocity, temperature, and concentration boundary layers. We set the laminar flow condition and varied the Richardson number (1.0-23) and the steam volume fraction (0.35-0.57). The experimental results were used to validate CFD analysis and HMTA models. For the former, we implemented a bulk condensation model assuming local thermal equilibrium into the CFD code and confirmed its validity. For the latter, we validated the HMTA-based correlations, confirming that the mixed convection correlation reasonably predicted the sum of wall and bulk condensation rates.
山本 智彦; 加藤 篤志; 早川 雅人; 下山 一仁; 荒 邦章; 畠山 望*; 山内 和*; 江田 優平*; 由井 正弘*
Nuclear Engineering and Technology, 56(3), p.893 - 899, 2024/03
被引用回数：1 パーセンタイル：51.66(Nuclear Science & Technology)In a secondary cooling system of a sodium-cooled fast reactor (SFR), rapid detection of hydrogen due to sodium-water reaction (SWR) caused by water leakage from a heat exchanger tube of a steam generator (SG) is important in terms of safety and property protection of the SFR. For hydrogen detection, the hydrogen detectors using atomic transmission phenomenon of hydrogen within Ni-membrane were used in Japanese proto-type SFR "Monju." However, during the plant operation, detection signals of water leakage were observed even in the situation without SWR concerning temperature up and down in the cooling system. For this reason, the study of a new hydrogen detector has been carried out to improve stability, accuracy and reliability. In this research, the authors focus on the difference in composition of hydrogen and the difference between the background hydrogen under normal plant operation and the one generated by SWR and theoretically estimate the hydrogen behavior in liquid sodium by using ultra-accelerated quantum chemical molecular dynamics (UA-QCMD). Based on the estimation, dissolved H or NaH, rather than molecular hydrogen (H
), is the predominant form of the background hydrogen in liquid sodium in terms of energetical stability. On the other hand, it was found that hydrogen molecules produced by the sodium-water reaction can exist stably as a form of a fine bubble concerning some confinement mechanism such as a NaH layer on their surface. In parallel, we observed experimentally that the fine bubbles of H stably existed in the liquid sodium than expected before. This paper describes the comparison between the theoretical estimation and experimental results based on hydrogen form in sodium in the development of the new hydrogen detector in Japan.
高松 邦吉; 舩谷 俊平*
Nuclear Engineering and Technology, 56(3), p.832 - 845, 2024/03
被引用回数：0 パーセンタイル：0.00(Nuclear Science & Technology)受動的安全性を持つRCCSは、大気を冷却材として使用するため、冷却材を喪失することはないが、大気の擾乱の影響を受けやすいという欠点がある。そのため、大気放射を利用したRCCSと、大気自然循環を利用したRCCSを実用化するためには、想定される自然災害や事故状態を含むあらゆる状況下で、原子炉からの発熱を常に除去できるのかについて安全評価を実施する必要がある。そこで本研究では、2種類の受動的RCCSについて、熱除去のための受動的安全性の余裕(裕度)について同一条件で比較した。その結果、提案した大気輻射を利用したRCCSは、外気(大気)の擾乱に対して原子炉圧力容器(RPV)の温度を安定的に維持できる利点を明らかにすることができた。さらに、RPV表面から放出される廃熱をすべて利用できる方法も提案した。
田上 浩孝; 飛田 吉春
Nuclear Engineering and Technology, 56(3), p.873 - 879, 2024/03
被引用回数：0 パーセンタイル：0.00(Nuclear Science & Technology)SFR炉心はsevere accidentを仮定した場合、炉心物質分布の変化を通じて、結果として発生するエネルギーが安全上重要となる。本論文では、3次元の時空間依存核計算と混相の熱流動計算をカップリングしたSIMMER-IVコードを用いて小型SFRのULOFの安全解析を実施した。このカップリングにより、SIMMER-IVコードは、ULOFの遷移過程における炉心損傷の拡大挙動と反応度への影響の計算に適用可能である。SIMMER-IVによる遷移過程解析ではいくつかの保守的な想定を行った。燃料挙動による反応度上昇の重要なメカニズムの一つは燃料と冷却材の接触(FCI)で発生するナトリウム蒸気圧であり、FCIに関わる不確かさを非常に保守的に想定した感度解析も実施した。この研究から、小型SFRのULOFの特徴が理解された。保守的な想定のもとでは臨界の発生はもっともらしい現象ではあるが、この時に発生するエネルギーは限定的である。
大沢 直樹*; Kim, S.-Y.*; 久保田 真彦*; Wu, H.*; 渡部 創; 伊藤 辰也; 永石 隆二
Nuclear Engineering and Technology, 56(3), p.812 - 818, 2024/03
被引用回数：2 パーセンタイル：75.80(Nuclear Science & Technology)An impregnated silica-based adsorbent was prepared by combining HONTA extractant, Crea extractant, and macroporous silica polymer composite particles (SiO-P). The performance of platinum-group metals adsorption and separation on prepared (HONTA + Crea)/SiO-P adsorbent was assessed by batch-adsorption and chromatographic separation studies. (HONTA + Crea)/SiO-P adsorbent showed high adsorption performance of Pd(II) owing to an affinity between Pd(II) and Crea extractant based on the Hard and Soft Acids and Bases theory. The chromatographic experiment showed that Pd(II) was recovered entirely from the feed solution using 0.2 M thiourea in 0.1 M HNO
. Possibility of recovery of Zr(IV), Mo(VI), and Re(VII) was also observed using the (HONTA + Crea)/SiO-P adsorbent.
Zablackaite, G.; 塩津 弘之; 城戸 健太朗; 杉山 智之
Nuclear Engineering and Technology, 56(2), p.536 - 545, 2024/02
被引用回数：2 パーセンタイル：75.80(Nuclear Science & Technology)Radioactive iodine is a representative fission product to be quantified for the safety assessment of nuclear facilities. In integral severe accident analysis codes, the iodine behavior is usually described by a multi-physical model of iodine chemistry in aqueous phase under radiation field and mass transfer through gas-liquid interface. The focus of studies on iodine source term evaluations using the combination approach is usually put on the chemical aspect, but each contribution to the iodine amount released to the environment has not been decomposed so far. In this study, we attempted the decomposition by revising the two-film theory of molecular-iodine mass transfer. The model involves an effective overall mass transfer coefficient to consider the iodine chemistry. The decomposition was performed by regarding the coefficient as a product of two functions of pH and the overall mass transfer coefficient for molecular iodine. The procedure was applied to the EPICUR experiment and suppression chamber in BWR.
古高 和禎; 大図 章; 藤 暢輔
Nuclear Engineering and Technology, 55(11), p.4002 - 4018, 2023/11
被引用回数：1 パーセンタイル：23.64(Nuclear Science & Technology)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 
s
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.
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：0.00(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
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.
, 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：6 パーセンタイル：84.29(Nuclear Science & Technology)UO・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相が維持されていた。1873Kでは(U,Zr)O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UO
と(Fe,Cr)UO
相の混合物が得られ、1873Kでは(U,Zr)O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。
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：6 パーセンタイル：84.29(Nuclear Science & Technology)本論文は、炭化ホウ素(BC)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。BC-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。BC-SS試料では、界面層に非結晶カーボンやFeB, FeB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。
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：9 パーセンタイル：81.25(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 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 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.
Pyke, C. K.*; Hiller, P. J.*; 駒 義和; 大木 恵一
Nuclear Engineering and Technology, 54(1), p.414 - 422, 2022/01
被引用回数：3 パーセンタイル：41.50(Nuclear Science & Technology)Presented in this paper is a methodology for combining a Bayesian statistical approach with the Data Quality Objectives structured decision-making methodology to provide increased levels of confidence in analytical data when approaching a waste boundary. Development of sampling and analysis plans for the characterisation of radioactive waste often use a simple, one pass statistical approach as underpinning for the sampling schedule. Using a Bayesian statistical approach introduces the concept of Prior information giving an adaptive sample strategy based on previous knowledge. This aligns more closely with the iterative approach demanded of the most commonly used structured decision-making tool in this area (Data Quality Objectives) and the potential to provide a more fully underpinned justification than the more traditional statistical approach. The approach described has been developed in a UK regulatory context but is translated to a waste stream from the Fukushima Daiichi nuclear power station to demonstrate how the methodology can be applied in this context to support decision making regarding the ultimate disposal option for radioactive waste in a more global context.
川田 賢一; 鈴木 徹*
Nuclear Engineering and Technology, 53(12), p.3930 - 3943, 2021/12
被引用回数：1 パーセンタイル：9.64(Nuclear Science & Technology)混合酸化物燃料を用いたナトリウム冷却高速増殖炉(SFR)の炉心損傷事象(CDA)の初期過程解析コードSAS4Aの解析能力の向上のために、著者らは前報において炉心流量喪失時炉停止機能喪失(ULOF)条件下での物理現象を詳細に検討した。その前報の研究成果として、燃料ピン崩壊後に残存した燃料ペレットが炉心中央部に移動する現象(燃料スタブモーション)が、適切に模擬すべき重要現象の一つとして選択された。本論文では、実験データの分析をもとに、スタブモーションに関わる挙動を評価し、概略を数値化し、従来のSAS4Aコードではモデル化されていなかった、燃料スタブの動きを表現するシンプルなモデルを新たに提案した。開発したモデルの適用性をCABRI試験の一連の解析を通じて検証し、崩壊炉心の反応性評価において、スタブモーションが合理的な保守性をもって再現されることを確認した。
竹田 武司; 大津 巌
Nuclear Engineering and Technology, 50(6), p.829 - 841, 2018/08
被引用回数：15 パーセンタイル：77.86(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.
竹田 武司; 大津 巌
Nuclear Engineering and Technology, 49(5), p.928 - 940, 2017/08
被引用回数：5 パーセンタイル：39.39(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.
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：5 パーセンタイル：39.39(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.
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：29 パーセンタイル：90.03(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.
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：39 パーセンタイル：93.09(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.
Smith, G.*; 加藤 智子
Nuclear Engineering and Technology, 42(1), p.1 - 8, 2010/02
放射性廃棄物地層処分においては、数千年もしくはそれ以上の超長期に渡って、Cl-36のような長半減期核種が、人間が普通にアクセスし利用する環境、すなわち生物圏に放出される。いかなる場合においても、処分場に起因して人間が受ける放射線量が放射線防護基準を満たすことを保証する必要がある。このような長期の時間枠における線量評価においては、地表環境や人間活動の変遷を考慮しなければならないという理由から、評価の枠組みを構築することは容易ではなく、長年に渡る国際共同プロジェクトによりこの問題が議論されてきた。本報では、放射線防護に関する国際的な勧告とサイト特有の評価におけるセーフティケース構築の準備に関して、国際協力により得られた成果及びJAEAを含む各国の研究アプローチについて概説する。
澤田 淳; 佐藤 久
Nuclear Engineering and Technology, 42(1), p.9 - 16, 2010/02
亀裂を対象とした平行平板モデルに用いられるパラメータ値の設定方法の検討に必要な単一亀裂内のデータ取得のために、亀裂の透明レプリカを用いた実験的検討を行った。光学的計測手法により亀裂開口幅分布やトレーサー試験時のトレーサー濃度データを高い空間解像度で定量的に取得した。亀裂開口幅分布の算術平均値,トレーサー試験から求めた開口幅,亀裂内体積測定から求めた平均開口幅などの異なる計測手法から求めた開口幅の値が一致することが示され、本試験データが良い精度で取得できていることを示している。亀裂開口幅データから局所的に三乗則が成り立つと仮定して実施した数値解析から得られる亀裂の透水量は透水試験の値より10%
100%大きな値となった。また、定量的なトレーサー濃度分布のデータは不均質亀裂内の移流分散の数値解析コードの検証にとても有用である。
