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Oxidation behavior of high-burnup advanced fuel cladding tubes in high-temperature steam

成川 隆文; 天谷 政樹

Journal of Nuclear Science and Technology, 56(7), p.650 - 660, 2019/07

To evaluate the oxidation behavior of high-burnup advanced fuel cladding tubes in high-temperature steam, laboratory-scale isothermal oxidation tests were conducted using the following advanced fuel cladding tubes with burnups of up to 85 GWd/t: M-MDA$textsuperscript{texttrademark}$, low-tin ZIRLO$textsuperscript{texttrademark}$, M5$textsuperscript{textregistered}$, and Zircaloy-2 (LK3). These oxidation tests were performed in steam-flowing conditions at temperatures ranging from 1173 to 1473 K for durations between 120 and 4000 s, and the oxidation kinetics was evaluated. The oxidation kinetics of the high-burnup advanced fuel cladding tube specimens estimated by assuming the parabolic rate law was comparable to or slower than that of the unirradiated Zircaloy-4 cladding tube specimens reported in a previous study. It is considered that the protective effect of the corrosion layer hindered oxidation. Furthermore, no increase in the oxidation kinetics because of the pre-hydriding was observed. The onset times of the breakaway oxidations of these cladding tube specimens were comparable to those of the unirradiated Zircaloy-4 cladding tubes reported in previous studies. Therefore, it is considered that the burnup extension up to 85 GWd/t and the use of the advanced fuel cladding tubes do not significantly increase the oxidation kinetics and do not significantly reduce the onset time of the breakaway oxidation.


Study on B$$_{4}$$C decoupler with burn-up reduction aiming at 1-MW pulsed neutron source

大井 元貴; 勅使河原 誠; 原田 正英; 池田 裕二郎

Journal of Nuclear Science and Technology, 56(7), p.573 - 579, 2019/07



Measurements of the $$^{243}$$Am neutron capture and total cross sections with ANNRI at J-PARC

木村 敦; 中村 詔司; 寺田 和司*; 中尾 太郎*; 水山 一仁*; 岩本 信之; 岩本 修; 原田 秀郎; 片渕 竜也*; 井頭 政之*; et al.

Journal of Nuclear Science and Technology, 56(6), p.479 - 492, 2019/06

Neutron total and capture cross sections of $$^{243}$$Am have been measured in Accurate Neutron Nucleus Reaction measurement Instrument at Materials and Life Science Experimental Facility of Japan Proton Accelerator Research Complex with a neutron TOF method. The neutron capture cross section in the energy region from 10 meV to 100 eV was determined using an array of Ge detectors. Three samples with different activities were used for measurements of the capture cross section. The neutron total cross section in the energy region from 4 meV to 100 eV was measured using Li-glass detectors. Derived cross-section value at neutron energy of 0.0253 eV is 87.7$$pm$$5.4 b for the capture cross section and 101$$pm$$11 b for the total cross section.


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

土井 大輔; 清野 裕; 宮原 信哉*; 宇埜 正美*

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

In severe accident scenarios for sodium-cooled fast reactors, it is desirable to gradually consume hydrogen generated by various ex-vessel phenomena without posting a challenge to containment integrity. An effective means is combustion of hydrogen jets containing sodium vapor and mist, but previous studies have been limited to determining ignition thresholds experimentally. The aim of this study was to visualize the ignition process in detail to investigate the ignition mechanism of hydrogen-sodium mixed jets. The ignition experiments of the hydrogen jet containing sodium mist were carried out under a condition of little turbulence. The ignition process was measured with an optical measurement system comprised of a high-speed camera and an image intensifier, and a spatial distribution of luminance was analyzed by image processing. Detail observation revealed that sodium mist particles burned as scattering sparks inside the jet and that hydrogen ignited around the mist particles. Additionally, the experimental results and a simple heat balance calculation indicated that the combustion heat of sodium mist particles could ignite the hydrogen as the heterogeneous ignition source in the fuel temperature range where the mist particle formation was promoted.


Measurements of thermal-neutron capture cross-section and resonance integral of neptunium-237

中村 詔司; 北谷 文人; 木村 敦; 上原 章寛*; 藤井 俊行*

Journal of Nuclear Science and Technology, 56(6), p.493 - 502, 2019/06

放射化法により$$^{237}$$Np(n,$$gamma$$)$$^{238}$$Np反応の熱中性子捕獲断面積($$sigma_{0}$$)及び共鳴積分(I$$_{0}$$)を測定した。$$^{237}$$Npの0.489eVにある第一共鳴に注意を払い、カドミウム差法において、ガドリニウムフィルタを用いて、カットオフエネルギーを0.133eVに設定して$$sigma_{0}$$を測定した。ネプツニウム237試料を、京都大学複合原子力科学研究所の研究炉にて照射した。照射位置における熱中性子束、及び熱外ウェストコット因子を決定するために、金合金線モニタ、及びコバルト合金線モニタも一緒に照射した。照射したネプツニウム237試料及びモニタ試料の生成放射能を、ガンマ線分光により測定した。ウェストコットの理論に基づき、$$sigma_{0}$$とI$$_{0}$$を、それぞれと186.9$$pm$$6.2 barn、及び1009$$pm$$90 barnと導出した。


Model updates and performance evaluations on fuel performance code FEMAXI-8 for light water reactor fuel analysis

宇田川 豊; 天谷 政樹

Journal of Nuclear Science and Technology, 56(6), p.461 - 470, 2019/06



Melting behavior and thermal conductivity of solid sodium-concrete reaction product

河口 宗道; 宮原 信哉; 宇埜 正美*

Journal of Nuclear Science and Technology, 56(6), p.513 - 520, 2019/06

本研究はナトリウム-コンクリート反応(SCR)によって発生する生成物について、融点及び熱伝導率を明らかにしたものである。試料は次の2種類の方法で作製した。1つ目は加熱炉内でナトリウムとコンクリート粉末の混合物を加熱したものである。2つ目はSCR実験を行い、その堆積物をサンプリングしたものである。前者は、過去の実験からナトリウムとコンクリートの混合割合を決定しており、後者は温度履歴やナトリウムとコンクリートの分布等、より現実的な条件を模擬している。熱重量・示唆熱(TG-DTA)測定から、試料の融点は865-942$$^{circ}$$Cであることが示されたが、金属ナトリウムを含む試料の融点は明確には分からなかった。そこで、より現実的な2つの試料については加熱炉内におけるその圧縮成型体の観察を行った。その観察により軟化温度は800-840$$^{circ}$$C、融点は840-850$$^{circ}$$Cであることが分かった。融点はTG-DTAの結果から10-20$$^{circ}$$C低い温度となった。FactSage 7.2による熱力学計算から、融解が始まる温度はNa$$_{2}$$SiO$$_{3}$$やNa$$_{4}$$SiO$$_{4}$$等の構成物質の融解により起きることが分かった。反応生成物の熱伝導率は$$lambda$$=1-3W/m-Kとなった。これは、xNa$$_{2}$$O-1-xSiO$$_{2}$$ (x=0.5, 0.33, 0.25)の熱伝導率と同程度であった。700$$^{circ}$$Cにおけるこの熱伝導率は非架橋酸素数(NBO/T)の式によって説明されることが分かった。


Experimental investigation of decontamination factor dependence on aerosol concentration in pool scrubbing

孫 昊旻; 柴本 泰照; 岡垣 百合亜; 与能本 泰介

Science and Technology of Nuclear Installations, 2019, p.1743982_1 - 1743982_15, 2019/06

Because a pool scrubbing is important for reducing radioactive aerosols to the environment for a nuclear reactor in a severe accident situation, many researches have been performed. However, decontamination factor (DF) dependence on aerosol concentration was seldom considered. DF dependence in the pool scrubbing with 2.4 m water submergence was investigated by light scattering aerosol spectrometers. It was observed that DF increased monotonically as decreasing particle number concentration in a constant thermohydraulic condition. Two validation experiments were conducted to confirm the observed DF dependence. In addition, characteristics of the DF dependence in different water submergences were investigated experimentally. It was found the DF dependence became more significant in higher water submergence.


Calculation of low-energy electron antineutrino spectra emitted from nuclear reactors with consideration of fuel burn-up

Riyana, E. S.*; 須田 翔哉*; 石橋 健二*; 松浦 秀明*; 片倉 純一*; Sun, G. M.*; 片野 好章

Journal of Nuclear Science and Technology, 56(5), p.369 - 375, 2019/05



The effect of hydride morphology on the failure strain of stress-relieved Zircaloy-4 cladding with an outer surface pre-crack under biaxial stress states

Li F.; 三原 武; 宇田川 豊; 天谷 政樹

Journal of Nuclear Science and Technology, 56(5), p.432 - 439, 2019/05

 パーセンタイル:100(Nuclear Science & Technology)

Hydride precipitates are considered to affect cladding integrity adversely during pellet-cladding mechanical interaction (PCMI) in a reactivity-initiated accident (RIA). This study aims to clarify the role of hydride precipitates in cladding failure under the biaxial stress condition. The amount and distribution of hydride precipitates (hydride morphology) were evaluated quantitatively and hydrogen content was measured to assess its effect on the decrease in outer surface hoop strain at failure (failure strain) of the samples. The decrease in failure strain of the hydrided samples was found to be more significant under lower strain ratios in the samples with shallower pre-crack. The failure strain of sample tended to be more sensitive to hydrogen content under the strain ratio with a higher axial component in the case of samples with hydrogen contents higher than ~150 wppm.


Phenomenological level density model with hybrid parameterization of deformed and spherical state densities

古立 直也*; 湊 太志; 岩本 修

Journal of Nuclear Science and Technology, 56(5), p.412 - 424, 2019/05



Estimating internal dose coefficients of short-lived radionuclides in accordance with ICRP 2007 Recommendations

真辺 健太郎; 佐藤 薫; 高橋 史明

Journal of Nuclear Science and Technology, 56(5), p.385 - 393, 2019/05

高エネルギー加速器施設では、高エネルギー粒子と施設構造物、施設内の空気等との核反応によって様々な放射性核種が生成され、施設作業者に対する潜在的な内部被ばく源となる。しかしながら、国際放射線防護委員会(ICRP)が公開しているICRP 2007年勧告に従う線量係数(放射性核種1Bq摂取当たりの預託実効線量)の中には、半減期が10分未満の短半減期核種は含まれていない。そこで、本研究では対応する元素の体内動態モデル等に基づき、このような短半減期核種の吸入摂取及び経口摂取に対するICRP 2007年勧告に従う線量係数を評価した。その結果をICRP 1990年勧告に従う線量係数と比較したところ、吸入摂取では線量係数が減少し、経口摂取では増加する傾向が見られた。こうした線量係数の変化は、線量計算手順の変更や消化管モデルの改訂等が原因であることが明らかになった。この結果は、高エネルギー加速器施設におけるICRP 2007年勧告に対応した放射線防護計画の立案に有用なものとなる。


Features of a control blade degradation observed ${it in situ}$ during severe accidents in boiling water reactors

Pshenichnikov, A.; 山崎 宰春; Bottomley, D.; 永江 勇二; 倉田 正輝

Journal of Nuclear Science and Technology, 56(5), p.440 - 453, 2019/05

In the present paper new results using ${it in situ}$ video, are presented regarding BWR control blade degradation up to 1750 K at the beginning of a nuclear severe accident. Energy-dispersive X-ray spectrometry (EDS) mapping indicated stratification of the absorber blade melt with formation of a chromium and boride-enriched layer. High content-B- and C-containing material with increased melting temperature acted like a shielding and was found to prevent further relocation of control blade claddings. The interacted layers around the B$$_{4}$$C granules prevented direct steam attack of residual B$$_{4}$$C. The results provide new insights for understanding of the absorber blade degradation mechanism under reducing conditions specific to Fukushima Dai-Ichi Unit 2 resulting from prolonged steam starvation.


An Interpretation of Fukushima-Daiichi Unit 3 plant data covering the two-week accident-progression phase based on correction for pressure data

佐藤 一憲

Journal of Nuclear Science and Technology, 56(5), p.394 - 411, 2019/05

福島第一3号機の圧力測定システムでは、運転中の蒸発/凝縮を補正するためにその一部に水柱が採用されている。これらの水柱の一部は事故条件下において蒸発し、正しい圧力データが示されていなかった。RPV(原子炉圧力容器), S/C(圧力抑制室)及びD/W(ドライウェル)の各圧力の比較を通し、水柱変化の効果を評価した。これによりRPV, S/C圧力データに対して水柱変化の効果の補正を行った。補正された圧力を用いて、事故進展中のRPV, S/C, D/W間のわずかな圧力差を評価した。この情報を、3号機の水位、CAMS(格納系雰囲気モニタリングシステム)および環境線量率などのデータとともに活用し、RPVおよびPCVの圧力上昇・下降および放射性物質の環境への放出に着目して事故進展挙動の解釈を行った。RPV内およびRPV外の燃料デブリのドライアウトはこれらの圧力低下を引き起こしている可能性がある一方、S/Cからペデスタルに流入したS/C水がペデスタルに移行した燃料デブリによって加熱されたことがPCV加圧の原因となっている。ペデスタル移行燃料デブリの周期的な再冠水とそのドライアウトは、最終的なデブリの再冠水まで数回の周期的な圧力変化をもたらしている。


Estimation of the release time of radio-tellurium during the Fukushima Daiichi Nuclear Power Plant accident and its relationship to individual plant events

高橋 千太郎*; 川島 茂人*; 日高 昭秀; 田中 草太*; 高橋 知之*

Nuclear Technology, 205(5), p.646 - 654, 2019/05

A simulation model was developed to estimate an areal (surface) deposition pattern of $$^{rm 129m}$$Te after the Fukushima Daiichi Nuclear Power Plant accident, and by using this model, timing and intensity of the release of $$^{rm 129m}$$Te were reversely estimated from the environmental monitoring data. The validation using data for $$^{137}$$Cs showed that the model simulated atmospheric dispersion and estimated surface deposition with relatively high accuracy. The estimated surface deposition pattern of $$^{rm 129m}$$Te was consistent with the actually measured one. The estimated time and activity of $$^{rm 129m}$$Te emission seemed to indicate that the $$^{rm 129m}$$Te was emitted mainly from Unit 3.


Effect of re-oxidation rate of additive cations on corrosion rate of stainless steel in boiling nitric acid solution

入澤 恵理子; 山本 正弘; 加藤 千明; 本岡 隆文; 伴 康俊

Journal of Nuclear Science and Technology, 56(4), p.337 - 344, 2019/04

 パーセンタイル:100(Nuclear Science & Technology)

The boiling nitric acid solution containing highly oxidizing cations dissolved from the spent nuclear fuels corrodes stainless steels because of the nobler corrosion potential and their fast reduction rate. The cations themselves are re-oxidized to higher oxidizing states in a bulk solution after the corrosion reaction. In this paper, the re-oxidation rate constants of typical cations, such as Cr, V, Pu, and Np, were analyzed, and discussed about the effect on time dependencies of the corrosion rate. It was indicated that the cations with a large re-oxidation rate constant, such as Np, could keep the corrosion rate at high level continuously for the long immersion duration.


Estimation method of systematic uncertainties in Monte Carlo particle transport simulation based on analysis of variance

橋本 慎太郎; 佐藤 達彦

Journal of Nuclear Science and Technology, 56(4), p.345 - 354, 2019/04

 パーセンタイル:100(Nuclear Science & Technology)



Visualized measurement of extremely high-speed droplets in Venturi scrubber

堀口 直樹; 吉田 啓之; 阿部 豊*

Journal of Nuclear Science and Technology, 56(3), p.278 - 290, 2019/03

 パーセンタイル:100(Nuclear Science & Technology)



The Effects of plutonium content and self-irradiation on thermal conductivity of mixed oxide fuel

生澤 佳久; 森本 恭一; 加藤 正人; 齋藤 浩介; 宇埜 正美*

Nuclear Technology, 205(3), p.474 - 485, 2019/03

 パーセンタイル:100(Nuclear Science & Technology)



Chemical reaction kinetics dataset of Cs-I-B-Mo-O-H system for evaluation of fission product chemistry under LWR severe accident conditions

宮原 直哉; 三輪 周平; 堀口 直樹; 佐藤 勇*; 逢坂 正彦

Journal of Nuclear Science and Technology, 56(2), p.228 - 240, 2019/02

 パーセンタイル:100(Nuclear Science & Technology)


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