Hamdani, A.; Abe, Satoshi; Ishigaki, Masahiro; Shibamoto, Yasuteru; Yonomoto, Taisuke
Progress in Nuclear Energy, 153, p.104415_1 - 104415_16, 2022/11
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
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.
Progress in Nuclear Energy, 144, p.104099_1 - 104099_7, 2022/02
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.
Aihara, Haruka; Watanabe, So; Shibata, Atsuhiro; Mahardiani, L.*; Otomo, Ryoichi*; Kamiya, Yuichi*
Progress in Nuclear Energy, 139, p.103872_1 - 103872_9, 2021/09
Goullo, M.*; Hokkinen, M.*; Suzuki, Eriko; Horiguchi, Naoki; Barrachin, M.*; Cousin, F.*
Progress in Nuclear Energy, 138, p.103818_1 - 103818_10, 2021/08
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/HO, Ar/H 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/HO and Ar/H 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.
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
Atkinson, S.*; Aoki, Takeshi; Litskevich, D.*; Merk, B.*; Yan, X.
Progress in Nuclear Energy, 134, p.103689_1 - 103689_10, 2021/04
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.
Herranz, L. E.*; Jacquemain, D.*; Nitheanandan, T.*; Sandberg, N.*; Barr, 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
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
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.
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
Sugawara, Takanori; Takei, Hayanori; Iwamoto, Hiroki; Oizumi, Akito; Nishihara, Kenji; Tsujimoto, Kazufumi
Progress in Nuclear Energy, 106, p.27 - 33, 2018/07
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.
Taninaka, Hiroshi; Takegoshi, Atsushi; Kishimoto, Yasufumi*; Mori, Tetsuya; Usami, Shin
Progress in Nuclear Energy, 101(Part C), p.329 - 337, 2017/11
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.
Gunji, Satoshi; Tonoike, Kotaro; Izawa, Kazuhiko; Sono, Hiroki
Progress in Nuclear Energy, 101(Part C), p.321 - 328, 2017/11
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 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.
Arima, Tatsumi*; Idemitsu, Kazuya*; Inagaki, Yaohiro*; Kawamura, Katsuyuki*; Tachi, Yukio; Yotsuji, Kenji
Progress in Nuclear Energy, 92, p.286 - 297, 2016/09
Diffusion and adsorption behavior of uranyl (UO) species is important for the performance assessment of radioactive waste disposal. The diffusion behaviors of UO, K, CO and Cl and HO in the aqueous solutions were evaluated by molecular dynamics (MD) calculations. The diffusion coefficient (De) of UO is the smallest and is 26% less than the self-diffusion coefficient of HO. For the aqueous solution with high concentration of carbonate ions, uranyl carbonate complexes: UOCO and UO(CO) can be observed. For the clay (montmorillonite or illite)-aqueous solution systems, the adsorption and diffusion behaviors of UO and K were evaluated by MD calculations. The distribution coefficients (Kd) increase with the layer charge of clay, and Kd of UO 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.
Miwa, Shuhei; Yamashita, Shinichiro; Osaka, Masahiko
Progress in Nuclear Energy, 92, p.254 - 259, 2016/09
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 under steam atmosphere, while the formation of CsBO 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 formation would have resulted in a lesser formation of gaseous hydrogen iodine, HI, and a high volatile atomic I under steam-starvation atmosphere.
Shen, X.*; Sun, Haomin; Deng, B.*; Hibiki, Takashi*; Nakamura, Hideo
Progress in Nuclear Energy, 89, p.140 - 158, 2016/05
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.
Yoshioka, Kenichi*; Kitada, Takanori*; Nagaya, Yasunobu
Progress in Nuclear Energy, 82, p.7 - 15, 2015/07
A reduced-moderation LWR has been developed for the reduction of spent fuel and for the efficient utilization of uranium resources. The streaming channel concept to improve the negative void reactivity coefficient is one of the features of the reactor. This concept makes the fuel assembly more heterogeneous. The geometrical heterogeneity makes azimuthal neutron flux distribution of fuel rods steep. To validate azimuthal neutron flux distribution calculation, we measured the distribution around fuel rods in reduced moderation LWR lattices. These measurements were conducted in NCA with the foil activation method. The core consisted of the central triangular tight lattice zone and the outer driver zone of a square lattice. A pile of polystyrene plates for simulating void fraction was installed into the triangular tight lattice. Azimuthal neutron flux distributions were deduced from the activity of these small foils measured with plastic scintillators. Measurements were compared to calculations by the MVP code with JENDL-3.3. It was found that calculations agreed well with measurements.
Minato, Futoshi; Iwamoto, Osamu
Progress in Nuclear Energy, 82, p.112 - 117, 2015/07
-decay and -delayed neutron emission of neutron-rich spherical nuclei are investigated. Our formalism adopts a self-consistent QRPA approach for -decay and Hauser-Feshbach statistical model for particle evaporation from highly excited state of daughter (precursor) nucleus. In this work, we particularly pay attention to the effects of two contributions. One is tensor force, which is not taken into account in conventional self-consistent QRPA but is important for reproducing half-lives of closed-shell nuclei. And another is isospin finite range pairing. They play a significant role to reduce energy of excited state of precursor nuclei. We found that these effects reduce the number of decay branches above neutron threshold of precursor nuclei and consequently a predicted -delayed neutron yields become smaller than that without them. This work is planned to apply to nuclear data evaluation of -delayed neutron yield of fission fragments in future.
Furukawa, Tomohiro; Rouillard, F.*
Progress in Nuclear Energy, 82, p.136 - 141, 2015/07
The application of a supercritical carbon dioxide (SC-CO) turbine cycle to fast rectors has the potential to enhance reliability because the SC-CO turbine system is capable of replacing the steam generator turbine system of conventional sodium cooled fast reactors. A key problem in the application is the corrosion of structural material by SC-CO at high temperatures. The authors have performed corrosion test on high-chromium martensitic and austenitic stainless steels in CO under the pressure conditions from atmospheric pressure to 25 MPa at elevated temperature, and proposed corrosion allowances of the steels for preliminary design of the SC-CO system. This paper initially reports the results of metallurgical examination of the steels after 8010 hours in SC-CO which is the longest immersion data in our experiments, and then describes the behavior of the oxide growth from the view point of estimation of the corrosion allowance for the design.
Sato, Hiroyuki; Ohashi, Hirofumi; Nakagawa, Shigeaki; Tachibana, Yukio; Kunitomi, Kazuhiko
Progress in Nuclear Energy, 82, p.46 - 52, 2015/07
Safety requirements and design considerations for a HTGR hydrogen production system by IS process are examined. Requirements in order to construct hydrogen production plants under conventional chemical plant regulation are identified. In addition, safety requirements for the collocation of the nuclear facility and hydrogen production plant utilizing IS process are investigated. Furthermore, design considerations to comply with the requirements are suggested and the technical feasibility of the design considerations is evaluated. The evaluation results clarified that design considerations suggested for coupling IS plant to HTGR are reasonably practicable.