ARKADIA; For the innovation of advanced nuclear reactor design

Ohshima, Hiroyuki; Asayama, Tai; Furukawa, Tomohiro; Tanaka, Masaaki; Uchibori, Akihiro; Takata, Takashi; Seki, Akiyuki; Enuma, Yasuhiro

Journal of Nuclear Engineering and Radiation Science, 9(2), p.025001_1 - 025001_12, 2023/04

https://doi.org/10.1115/1.4054726

This paper describes the outline and development plan for ARKADIA to transform advanced nuclear reactor design to meet expectations of a safe, economic, and sustainable carbon-free energy source. ARKADIA will realize Artificial Intelligence (AI)-aided integrated numerical analysis to offer the best possible solutions for the design and operation of a nuclear plant, including optimization of safety equipment. State-of-the-art numerical simulation technologies and a knowledge base that stores data and insights from past nuclear reactor development projects and R&D are integrated with AI. In the first phase of development, ARKADIA-Design and ARKADIA-Safety will be constructed individually, with the first target of sodium-cooled reactor. In a subsequent phase, everything will be integrated into a single entity applicable not only to advanced rectors with a variety of concepts, coolants, configurations, and output levels but also to existing light-water reactors.

Void reactivity in lead and bismuth sample reactivity experiments at Kyoto University Critical Assembly

Pyeon, C. H.*; Katano, Ryota; Oizumi, Akito; Fukushima, Masahiro

Nuclear Science and Engineering, 18 Pages, 2023/00

https://doi.org/10.1080/00295639.2023.2172311

Sample reactivity and void reactivity experiments are carried out in the solid-moderated and solid-reflected cores at the Kyoto University Critical Assembly (KUCA) with the combined use of aluminum (Al), lead (Pb) and bismuth (Bi) samples, and Al spacers simulating the void. MCNP6.2 eigenvalue calculations together with JENDL-4.0 provide good accuracy of sample reactivity with the comparison of experimental results; also experimental void reactivity is attained by using MCNP6.2 together with JENDL-4.0 and ENDF/B-VII.1 with a marked accuracy of relative difference between experiments and calculations. Uncertainty quantification of sample reactivity and void reactivity is acquired by using the sensitivity coefficients based on MCNP6.2/ksen and covariance library data of SCALE6.2 together with ENDF/B-VII.1, arising from the impact of uncertainty induced by Al, Pb and Bi cross sections. A series of reactivity analyses with the Al spacer simulating the void demonstrates the means of analyzing the void in the solid-moderated and solid-reflected cores at KUCA

Sensitivity coefficient evaluation of an accelerator-driven system using ROM-Lasso method

Katano, Ryota; Yamamoto, Akio*; Endo, Tomohiro*

Nuclear Science and Engineering, 196(10), p.1194 - 1208, 2022/10

https://doi.org/10.1080/00295639.2022.2067447

In this study, we propose the ROM-Lasso method that enables efficient evaluation of sensitivity coefficients of neutronics parameters to cross-sections. In the proposed method, a vector of sensitivity coefficients is expanded by subspace bases, so-called Active Subspace (AS) based on the idea of Reduced Order Modeling (ROM). Then, the expansion coefficients are evaluated by the Lasso linear regression between cross-sections and neutronics parameters obtained by the random sampling. The proposed method can be applied in the case where the adjoint method is difficult to be applied since the proposed method uses only forward calculations. In addition, AS is an effective subspace that can expand the vector of sensitivity coefficients with the lower number of dimension. Thus, the number of unknows is reduced from the original number of input parameters and the calculation cost is dramatically improved compared to the Lasso regression without AS. In this paper, we conducted ADS burnup calculations as a verification. We have shown how AS bases are obtained and the applicability of the proposed method.

Reactor physics experiment on a graphite-moderated core to construct integral experiment database for HTGR

Okita, Shoichiro; Fukaya, Yuji; Sakon, Atsushi*; Sano, Tadafumi*; Takahashi, Yoshiyuki*; Unesaki, Hironobu*

Nuclear Science and Engineering, 7 Pages, 2022/00

https://doi.org/10.1080/00295639.2022.2087836

Visualization of radioactive substances using a freely moving gamma-ray imager based on Structure from Motion

Sato, Yuki; Minemoto, Kojiro*; Nemoto, Makoto*; Torii, Tatsuo

Journal of Nuclear Engineering and Radiation Science, 7(4), p.042003_1 - 042003_12, 2021/10

https://doi.org/10.1115/1.4048843

Feasibility study of the one-dimensional radiation distribution sensing method using an optical fiber sensor based on wavelength spectrum unfolding

Terasaka, Yuta; Watanabe, Kenichi*; Uritani, Akira*; Yamazaki, Atsushi*; Sato, Yuki; Torii, Tatsuo; Wakaida, Ikuo

Journal of Nuclear Engineering and Radiation Science, 7(4), p.042002_1 - 042002_7, 2021/10

https://doi.org/10.1115/1.4048841

For the application in the measurement of the high dose rate hot spots inside the Fukushima Daiichi Nuclear Power Station (FDNPS) buildings, we propose a novel one-dimensional radiation distribution sensing method using an optical fiber sensor based on wavelength spectrum unfolding. The proposed method estimates the incident position of radiation to the fiber by the unfolding of the wavelength spectrum output from the fiber edge using the fact that the attenuation length of light along the fiber depends on the wavelength. Because this method measures the integrated light intensity, this method can avoid the problem of counting loss and signal pile-up, which occurs in the radiation detector with pulse counting mode under high dose rate field. Through basic experiments using the ultraviolet light source and Sr/Y radioactive point source, basic properties of source position detection were confirmed.

Uncertainty quantification of lead and bismuth sample reactivity worth at Kyoto University Critical Assembly

Pyeon, C. H.*; Yamanaka, Masao*; Fukushima, Masahiro

Nuclear Science and Engineering, 195(8), p.877 - 889, 2021/08

https://doi.org/10.1080/00295639.2020.1870861

Uncertainty quantification of lead (Pb) and bismuth (Bi) sample reactivity worth is numerically determined using the SCALE6.2 code system and experimental results obtained from the solid-moderated and solid-reflected core at the Kyoto University Critical Assembly (KUCA) to demonstrate the sensitivity coefficients of aluminum (Al) and Bi scattering reactions. From the results of the numerical analyses, the impact of Al and Bi scattering cross sections obtained using SCALE6.2/TSAR is disclosed on the Bi sample reactivity worth using Al reference and Bi test samples, although the uncertainty itself is small in the Bi sample reactivity worth.

Judgment on convergence-in-distribution of Monte Carlo tallies under autocorrelation

Ueki, Taro

Nuclear Science and Engineering, 194(6), p.422 - 432, 2020/06

https://doi.org/10.1080/00295639.2019.1710418

In Monte Carlo criticality calculation, the convergence-in-distribution check of the sample mean of tallies can be approached in terms of the influence range of autocorrelation. In this context, it is necessary to evaluate the attenuation of autocorrelation coefficients over lags. However, in just one replica of calculation, it is difficult to accurately estimate small ACCs at large lags because of the comparability with statistical uncertainty. This paper proposes a method to overcome such an issue. Its essential component is the transformation of a standardized time series of tallies so that the resulting series asymptotically converges in distribution to Brownian motion. The convergence-in-distribution check is constructed based on the independent increment property of Brownian motion. The judgment criterion is set by way of the spectral analysis of fractional Brownian motion. Numerical results are demonstrated for extreme and standard types of criticality calculation.

A Study on sodium-concrete reaction in presence of internal heating

Kawaguchi, Munemichi; Miyahara, Shinya*; Uno, Masayoshi*

Journal of Nuclear Engineering and Radiation Science, 6(2), p.021305_1 - 021305_9, 2020/04

https://doi.org/10.1115/1.4045898

Sodium-concrete reaction (SCR) is one of the important phenomena during severe accidents in sodium-cooled fast reactors (SFRs) owing to the generation of large sources of hydrogen and aerosols in the containment vessel. In this study, SCR experiments with an internal heater were performed to investigate the chemical reaction beneath the internal heater (800C), which was used to simulate the obstacle and heating effect on SCR. Furthermore, the effects of the internal heater on the self-termination mechanism were discussed. The internal heater on the concrete hindered the transport of Na into the concrete. Therefore, Na could start to react with the concrete at the periphery of the internal heater, and the concrete ablation depth at the periphery was larger than under the internal heater. The high Na pool temperature of 800C increased largely the Na aerosol release rate, which was explained by Na evaporation and hydrogen bubbling, and formed the porous reaction product layer, whose porosity was 0.54-0.59 from the mass balance of Si and image analyzing EPMA mapping. They had good agreement with each other. The porous reaction products decreased the amount of Na transport into the reaction front. The Na concentration around the reaction front became about 30wt.% despite the position of the internal heater. It was found that the Na concentration condition was one of the dominant parameters for the self-termination of SCR, even in the presence of the internal heater.

Evaluation of radiation effects on residents living around the NSRR under external hazards

Motome, Yuiko; Akiyama, Yoshiya; Murao, Hiroyuki

Journal of Nuclear Engineering and Radiation Science, 6(2), p.021115_1 - 021115_11, 2020/04

https://doi.org/10.1115/1.4045075

The nuclear safety research reactor (NSRR) is a research reactor of training research isotopes general atomics -annular core pulse reactor type. The NSRR facility has been utilized for fuel irradiation experiments to study the behaviors of nuclear fuels under reactivity-initiated accident conditions. Under the new regulation standards, which was established after the Fukushima Daiichi accident, research reactors are regulated based on the risk of the facilities. To apply the graded approach, the radiation effects on residents living around the NSRR under the external hazards were evaluated, and the level of the risk of the NSRR facility was investigated. This paper summarizes the result of the evaluation in the case where the safety functions are lost due to a tornado, an earthquake followed by a tsunami. All in all, the risk is confirmed to be relatively low, since the effective dose on the residents is found to be below 5 mSv per event due to the loss of the safety functions.