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Journal Articles

Simulation of a jet flow rectified by a grating-type structure using immersed boundary methods

Hirose, Yoshiyasu; Abe, Satoshi; Ishigaki, Masahiro*; Shibamoto, Yasuteru; Hibiki, Takashi*

Progress in Nuclear Energy, 169, p.105085_1 - 105085_13, 2024/04

Journal Articles

Critical heat flux for downward flows in vertical round pipes

Hirose, Yoshiyasu; Shibamoto, Yasuteru; Hibiki, Takashi*

Progress in Nuclear Energy, 168, p.105027_1 - 105027_17, 2024/03

Journal Articles

Inverse estimation scheme of radioactive source distributions inside building rooms based on monitoring air dose rates using LASSO; Theory and demonstration

Shi, W.*; Machida, Masahiko; Yamada, Susumu; Yoshida, Toru*; Hasegawa, Yukihiro*; Okamoto, Koji*

Progress in Nuclear Energy, 162, p.104792_1 - 104792_19, 2023/08

Predicting radioactive source distributions inside reactor building rooms based on monitoring air dose rates is one of the most essential steps towards decommissioning of nuclear power plants. However, the attempt is rather a difficult task, because it can be generally mapped onto mathematically ill-posed problem. Then, in order to successfully perform the inverse estimations on radioactive source distributions even in such ill-posed conditions, we suggest that a machine learning method, least absolute shrinkage and selection operator (LASSO) minimizing the loss function, $$||CP-Q||_2^2+lambda||_1$$ is a promising scheme. For the purpose of its feasibility demonstrations in real building rooms, we employ PHITS code to make LASSO input as the above matrix C connecting the radioactive source vector P defined on surface meshes of structural materials with the air dose rate vector Q measured at internal positions inside the rooms. We develop a mathematical criterion on the number of monitoring points to correctly predict source distributions based on the theory of Candes and Tao. Then, we confirm that LASSO actually shows extremely high possibility for source distribution reconstructions as far as the number of detection points satisfies our criterion. Moreover, we verify that radioactive hot spots can be truly reconstructed in an experiment setup. At last, we examine an influence factor like detector-source distance to enhance the predicting possibility in the inverse estimation. From the above demonstrations, we propose that LASSO scheme is a quite useful way to explore hot spots as seen in damaged nuclear power plants like Fukushima Daiichi nuclear power plants.

Journal Articles

Generalized extreme value analysis of criticality tallies in Monte Carlo calculation

Ueki, Taro

Progress in Nuclear Energy, 159, p.104630_1 - 104630_9, 2023/05

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

In this work, the methodology of Generalized Extreme Value (GEV) is applied to criticality tallies in Monte Carlo fission source cycles in order to evaluate the utility value of the distribution tail ends. Numerical results obtained under a sufficiently large number of particles per cycle show that the extreme value index (EVI) in GEV falls within the range of Weibull distribution including the EVI of Gumbel distribution as the role of a boundary value layer. GEV is also applied to a historically-challenging loosely-coupled system for demonstrating population diagnosis under an insufficient number of particles per cycle. It turns out that the transition from one equilibrium to other equilibrium makes the EVIs of upper and lower distribution tail ends depart from each other so that one of them falls in the range of Weibull distribution and the other in that of Frechet distribution.

Journal Articles

CFD analysis on stratification dissolution and breakup of the air-helium gas mixture by natural convection in a large-scale enclosed vessel

Hamdani, A.; Abe, Satoshi; Ishigaki, Masahiro; Shibamoto, Yasuteru; Yonomoto, Taisuke

Progress in Nuclear Energy, 153, p.104415_1 - 104415_16, 2022/11

 Times Cited Count:3 Percentile:71.05(Nuclear Science & Technology)

Journal Articles

Liquid decontamination using acidic electrolyzed water for various uranium-contaminated steel surfaces in dismantled centrifuge

Sakasegawa, Hideo; Nomura, Mitsuo; Sawayama, Kengo; Nakayama, Takuya; Yaita, Yumi*; Yonekawa, Hitoshi*; Kobayashi, Noboru*; Arima, Tatsumi*; Hiyama, Toshiaki*; Murata, Eiichi*

Progress in Nuclear Energy, 153, p.104396_1 - 104396_9, 2022/11

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

When dismantling centrifuges in uranium-enrichment facilities, decontamination techniques must be developed to remove uranium-contaminated surfaces of dismantled parts selectively. Dismantled uranium-contaminated parts can be disposed of as nonradioactive wastes or recycled after decontamination appropriate for clearance. previously, we developed a liquid decontamination technique using acidic electrolyzed water to remove uranium-contaminated surfaces. However, further developments are still needed for its actual application. Dismantled parts have various uranium-contaminated surface features due to varied operational conditions, inhomogeneous decontamination using iodine heptafluoride gas, and changes in long-term storage conditions after dismantling. Here, we performed liquid decontamination on specimens with varying uranium-contaminated surfaces cut from a centrifuge made of low-carbon steel. From the results, the liquid decontamination can effectively remove the uranium-contaminated surfaces, and radioactive concentrations fell below the target value within twenty minutes. Although the required time should also depend on dismantled parts' sizes and shapes in their actual application, we demonstrated that it could be an effective decontamination technique for uranium-contaminated steels of dismantled centrifuges.

Journal Articles

Weierstrass function methodology for uncertainty analysis of random media criticality with spectrum range control

Ueki, Taro

Progress in Nuclear Energy, 144, p.104099_1 - 104099_7, 2022/02

 Times Cited Count:0 Percentile:0.01(Nuclear Science & Technology)

Randomized Weierstrass function (RWF) has been under development for evaluating the uncertainty of random media criticality due to the material mixture under disorder. In this work, the modelling capability of RWF is refined so that the spectrum range can be controlled by specifying its lower and upper ends of the frequency domain variable. As a result, it becomes possible to make fair criticality comparison among replicas of random media under inverse power law power spectra. Technically, the infinite sum of trigonometric terms in RWF is extended to cover the arbitrarily low frequency domain and then truncated to finite terms for the sole purpose of spectrum range control. This means that the refinement is free of the convergence issue towards a fractal characteristic of Weierstrass function and thus termed Incomplete Randomized Weierstrass function (IRWF). As a demonstration, a three-dimensional version of IRWF is applied to the mixture of three fuels with different burnups in a water-moderated environment. Monte Carlo criticality calculations are carried out to evaluate the uncertainty of neutron effective multiplication factor due to the indeterminacy of the fuel mixture formation.

Journal Articles

Oxidative decomposition of ammonium ion with ozone in the presence of cobalt and chloride ions for the treatment of radioactive liquid waste

Aihara, Haruka; Watanabe, So; Shibata, Atsuhiro; Mahardiani, L.*; Otomo, Ryoichi*; Kamiya, Yuichi*

Progress in Nuclear Energy, 139, p.103872_1 - 103872_9, 2021/09

 Times Cited Count:1 Percentile:32.89(Nuclear Science & Technology)

Journal Articles

Interaction between caesium iodide particles and gaseous boric acid in a flowing system through a thermal gradient tube (1030 K-450 K) and analysis with ASTEC/SOPHAEROS

Gou$"e$llo, M.*; Hokkinen, M.*; Suzuki, Eriko; Horiguchi, Naoki; Barrachin, M.*; Cousin, F.*

Progress in Nuclear Energy, 138, p.103818_1 - 103818_10, 2021/08

 Times Cited Count:3 Percentile:58.27(Nuclear Science & Technology)

The present work aimed to study the transport of caesium iodide particles through a Thermal Gradient Tube (TGT) from 1023 K to 453 K. Retention inside the tube was evaluated for laminar flowrates composed of argon and steam. Higher retention of particles was highlighted for the experiments using higher steam content and lower flowrate. The second phase of the experiment aimed at identifying the possible revaporization or/and resuspension processes after the deposition. Three atmosphere compositions (Ar/H$$_{2}$$O, Ar/H$$_{2}$$ and Ar/Air) were investigated. The particles removed from what was deposited on the surface walls during the sampling phase exhibited a similar GMD in Ar/H$$_{2}$$O and Ar/H$$_{2}$$ and a bigger diameter in Ar/Air. The experimental results were then analysed with the SOPHAEROS module of the ASTEC code. Overall, the results obtained during the first phase were in agreement with the measured experimental results and during second phase led to no resuspension process.

Journal Articles

Development of a membrane reactor with a closed-end silica membrane for nuclear-heated hydrogen production

Myagmarjav, O.; Tanaka, Nobuyuki; Nomura, Mikihiro*; Noguchi, Hiroki; Imai, Yoshiyuki; Kamiji, Yu; Kubo, Shinji; Takegami, Hiroaki

Progress in Nuclear Energy, 137, p.103772_1 - 103772_7, 2021/07

 Times Cited Count:6 Percentile:73.26(Nuclear Science & Technology)

Journal Articles

Part 3, Evaluating a small modular high temperature reactor design during control rod withdrawal and a depressurised loss of coolant accidents

Atkinson, S.*; Aoki, Takeshi; Litskevich, D.*; Merk, B.*; Yan, X.

Progress in Nuclear Energy, 134, p.103689_1 - 103689_10, 2021/04

 Times Cited Count:1 Percentile:16.97(Nuclear Science & Technology)

This article evaluates the safety features of the designed 10 MWth U-Battery concept with respect to a control rod withdrawal and a depressurised loss of coolant accident. This article provides the evaluation methodology for both transients, using a one-dimensional heat transfer model involving point reactor kinetic model to simulate reactor feedback in the control rod withdrawal. Overall, this work has shown that during the control rod withdrawal the fuel temperature rises by 110 K and at this point the excess reactivity is compensated by the negative temperature coefficient of the fuel. During the depressurised loss of coolant accident, the maximum fuel temperature reached 1455 K after 60 hours. This concludes that during both transients the temperatures maintained well below the maximum fuel operating temperature.

Journal Articles

The Working group on the analysis and management of accidents (WGAMA); A Historical review of major contributions

Herranz, L. E.*; Jacquemain, D.*; Nitheanandan, T.*; Sandberg, N.*; Barr$'e$, F.*; Bechta, S.*; Choi, K.-Y.*; D'Auria, F.*; Lee, R.*; Nakamura, Hideo

Progress in Nuclear Energy, 127, p.103432_1 - 103432_14, 2020/09

 Times Cited Count:2 Percentile:11.59(Nuclear Science & Technology)

Journal Articles

A Review of separation processes proposed for advanced fuel cycles based on technology readiness level assessments

Baron, P.*; Cornet, S. M.*; Collins, E. D.*; DeAngelis, G.*; Del Cul, G.*; Fedorov, Y.*; Glatz, J. P.*; Ignatiev, V.*; Inoue, Tadashi*; Khaperskaya, A.*; et al.

Progress in Nuclear Energy, 117, p.103091_1 - 103091_24, 2019/11

 Times Cited Count:54 Percentile:93.84(Nuclear Science & Technology)

The results of an international review of separation processes for spent nuclear fuel (SNF) recycling in future closed fuel cycles with the evaluation of Technology Readiness Level are reported. This study was made by the Expert Group on Fuel Recycling Chemistry (EGFRC) organised by the Nuclear Energy Agency (NEA) of the Organisation for Economic Co-operation and Development (OECD). A unique feature of this study was that processes were classified according to a hierarchy of separations aimed at different elements within spent fuel (uranium; uranium-plutonium co-recovery; minor actinides; high heat generating radionuclides) and also the Head-end processes, used to prepare the SNF for chemical separation, were included. Separation processes covered both wet (hydrometallurgical) and dry (pyro-chemical) processes.

Journal Articles

STRAD project for systematic treatments of radioactive liquid wastes generated in nuclear facilities

Watanabe, So; Ogi, Hiromichi*; Arai, Yoichi; Aihara, Haruka; Takahatake, Yoko; Shibata, Atsuhiro; Nomura, Kazunori; Kamiya, Yuichi*; Asanuma, Noriko*; Matsuura, Haruaki*; et al.

Progress in Nuclear Energy, 117, p.103090_1 - 103090_8, 2019/11

AA2019-0193.pdf:1.29MB

 Times Cited Count:9 Percentile:78.07(Nuclear Science & Technology)

Journal Articles

Research and development activities for accelerator-driven system in JAEA

Sugawara, Takanori; Takei, Hayanori; Iwamoto, Hiroki; Oizumi, Akito; Nishihara, Kenji; Tsujimoto, Kazufumi

Progress in Nuclear Energy, 106, p.27 - 33, 2018/07

 Times Cited Count:13 Percentile:85.51(Nuclear Science & Technology)

The Japan Atomic Energy Agency (JAEA) has investigated an accelerator-driven system (ADS) to transmute minor actinides which will be partitioned from the high level waste. There are various inherent issues for the research and development on the ADS. The recent two activities to realize a feasible and reliable ADS concept are introduced in this paper. For the feasibility, the design of a beam window which is a boundary of the accelerator and the subcritical core, is one of the most important issues. To mitigate the design condition of the beam window, namely to reduce the proton beam current, the subcritical core concept with subcriticality adjustment rods were investigated. For the reliability, the beam-trip is the inherent and serious issue for the ADS design because it induces rapid temperature change to coolant and structures in the subcritical core. To improve the beam-trip frequencies, a double-accelerator concept was proposed and its beam-trip frequency was estimated.

Journal Articles

A Refined analysis on the power reactivity loss measurement in Monju

Taninaka, Hiroshi; Takegoshi, Atsushi; Kishimoto, Yasufumi*; Mori, Tetsuya; Usami, Shin

Progress in Nuclear Energy, 101(Part C), p.329 - 337, 2017/11

 Times Cited Count:2 Percentile:19.91(Nuclear Science & Technology)

The present paper describes the evaluation of the power reactivity loss data obtained in the Japanese prototype fast breeder reactor Monju. The most recent analysis on the power reactivity loss measurement (Takano, et al., 2008) is updated considering the following findings: (a) in-core temperature distribution effect, (b) crystalline binding effect, (c) logarithmic averaging of the fuel temperature, (d) localized fuel thermal elongation effect, (e) updated Japanese Evaluated Nuclear Data Library, JENDL-4.0, and (f) refined corrections on the measured value. The influences of the updates are quantitatively identified and the most precise and probable C/E value is derived together with a thorough uncertainty evaluation. As a result, it is revealed that the analysis overestimates the measurement by 4.6% for the measurement uncertainty of 2.0%. The discrepancy is reduced to as small as 1.1% when the core bowing effect is considered, which implies the importance of the core bowing effect in the calculation of the power reactivity loss.

Journal Articles

Study of experimental core configuration of the modified STACY for measurement of criticality characteristics of fuel debris

Gunji, Satoshi; Tonoike, Kotaro; Izawa, Kazuhiko; Sono, Hiroki

Progress in Nuclear Energy, 101(Part C), p.321 - 328, 2017/11

 Times Cited Count:3 Percentile:29.14(Nuclear Science & Technology)

Criticality safety of fuel debris, particularly MCCI (Molten-Core-Concrete-Interaction) products, is one of the major safety issues for decommissioning of Fukushima Daiichi Nuclear Power Station. Criticality or subcriticality condition of the fuel debris is still uncertain; its composition, location, neutron moderation, etc. are not yet confirmed. The effectiveness of neutron poison in cooling water is also uncertain for use as a criticality control of fuel debris. A database of computational models is being built by Japan Atomic Energy Agency (JAEA), covering a wide range of possible conditions of such composition, neutron moderation, etc., to facilitate assessing criticality characteristics once fuel debris samples are taken and their conditions are known. The computational models also include uncertainties which are to be clarified by critical experiments. These experiments are planned and will be conducted by JAEA with the modified STACY (STAtic experiment Critical facilitY) and samples to simulate fuel debris compositions. Each of the samples will be cladded by a zircalloy tube whose outer shape is compatible with the fuel rod of STACY and loaded into an array of the fuel rods. This report introduces a study of experimental core configurations to measure the reactivity worth of samples simulating MCCI products. Parameters to be varied in the computation models for the experimental series are:(1) Uranium dioxide with $$^{235}$$U enrichments of 3, 4, and 5 wt.%; (2) Concrete volume fraction in the samples of 0, 20, 40, 60, and 80%; and (3) Porosity of the samples filled from 0 to 80% where the sample void is filled with water. It is concluded that the measurement is feasible in both under- and over-moderated conditions. Additionally, the required amount of samples was estimated.

Journal Articles

Diffusion and adsorption of uranyl ion in clays; Molecular dynamics study

Arima, Tatsumi*; Idemitsu, Kazuya*; Inagaki, Yaohiro*; Kawamura, Katsuyuki*; Tachi, Yukio; Yotsuji, Kenji

Progress in Nuclear Energy, 92, p.286 - 297, 2016/09

 Times Cited Count:10 Percentile:68.73(Nuclear Science & Technology)

Diffusion and adsorption behavior of uranyl (UO$$_2^{2+}$$) species is important for the performance assessment of radioactive waste disposal. The diffusion behaviors of UO$$_2^{2+}$$, K$$^{+}$$, CO$$_3^{2-}$$ and Cl$$^{-}$$ and H$$_{2}$$O in the aqueous solutions were evaluated by molecular dynamics (MD) calculations. The diffusion coefficient (De) of UO$$_2^{2+}$$ is the smallest and is 26% less than the self-diffusion coefficient of H$$_{2}$$O. For the aqueous solution with high concentration of carbonate ions, uranyl carbonate complexes: UO$$_{2}$$CO$$_{3}$$ and UO$$_{2}$$(CO$$_{3}$$)$$^{2-}$$ can be observed. For the clay (montmorillonite or illite)-aqueous solution systems, the adsorption and diffusion behaviors of UO$$_2^{2+}$$ and K$$^{+}$$ were evaluated by MD calculations. The distribution coefficients (Kd) increase with the layer charge of clay, and Kd of UO$$_2^{2+}$$ might be smaller than that of K$$^{+}$$. Further, their two-dimensional diffusion coefficients were relatively small in the adsorption layer and were extremely small for illite with higher layer charge.

Journal Articles

Prediction of the effects of boron release kinetics on the vapor species of cesium and iodine fission products

Miwa, Shuhei; Yamashita, Shinichiro; Osaka, Masahiko

Progress in Nuclear Energy, 92, p.254 - 259, 2016/09

 Times Cited Count:17 Percentile:84.5(Nuclear Science & Technology)

Cesium (Cs) and iodine (I) vapor species formed just after release from degraded fuels were predicted by means of the chemical equilibrium calculation with focuses on the effects of boron (B) release kinetics. Modified equations for the release kinetics of Cs, I and molybdenum (Mo) were utilized for evaluation of atmospheric dependences of their releases fractions. The release kinetics of B was evaluated considering the formation of iron (Fe)-B-O-H compounds. The release of B was enhanced above approximately 2250 K with the vapor species of CsBO$$_{2}$$ under steam atmosphere, while the formation of CsBO$$_{2}$$ was limited under steam-starvation atmosphere due to the much lower the release of B by the formation of low volatile Fe-B compounds. This limitation of CsBO$$_{2}$$ formation would have resulted in a lesser formation of gaseous hydrogen iodine, HI, and a high volatile atomic I under steam-starvation atmosphere.

Journal Articles

Gas-liquid bubbly flow structure in a vertical large-diameter square duct

Shen, X.*; Sun, Haomin; Deng, B.*; Hibiki, Takashi*; Nakamura, Hideo

Progress in Nuclear Energy, 89, p.140 - 158, 2016/05

 Times Cited Count:20 Percentile:88.97(Nuclear Science & Technology)

An experimental study was performed on the local structure of upward air-water two-phase flow in a vertical large diameter square duct by using a four-sensor probe. The four-sensor probe method classifying spherical and non-spherical bubbles was applied as a key measurement way to obtain local parameters such as 3-D bubble velocity vector, bubble diameter and interfacial area concentration. Both the local void fraction and interfacial area concentration indicated radial core-peak and wall-peak distributions at low and high liquid flow rates respectively. The 2 components of the bubble velocity vector in the cross-section revealed that there exists a rotating secondary flow in the octant symmetric triangular area and the magnitude of the rotating secondary flow increases with the liquid flow rate. Some of constitutive correlations of drift-flux model and interfacial area concentration are reviewed to study their predictabilities against the present data.

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