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

https://doi.org/10.1016/j.pnucene.2022.104415

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

https://doi.org/10.1016/j.pnucene.2021.104099

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.

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

Goullo, M.*; Hokkinen, M.*; Suzuki, Eriko; Horiguchi, Naoki; Barrachin, M.*; Cousin, F.*

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

https://doi.org/10.1016/j.pnucene.2021.103818

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.

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

https://doi.org/10.1016/j.pnucene.2021.103689

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.

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

https://doi.org/10.1016/j.pnucene.2019.103091

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.

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

https://doi.org/10.1016/j.pnucene.2018.02.007

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.

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

https://doi.org/10.1016/j.pnucene.2017.03.002

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.

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

https://doi.org/10.1016/j.pnucene.2016.02.023

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.

Azimuthal flux distribution measurements around fuel rods in reduced-moderation LWR lattices

Yoshioka, Kenichi*; Kitada, Takanori*; Nagaya, Yasunobu

Progress in Nuclear Energy, 82, p.7 - 15, 2015/07

https://doi.org/10.1016/j.pnucene.2014.07.037

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.

Oxidation and carburizing of FBR structural materials in carbon dioxide

Furukawa, Tomohiro; Rouillard, F.*

Progress in Nuclear Energy, 82, p.136 - 141, 2015/07

https://doi.org/10.1016/j.pnucene.2014.07.038

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.