Validation of fuel behavior analysis code FEMAXI-8 using fast reactor MOX fuel irradiation tests

Ikusawa, Yoshihisa; Nagayama, Masahiro*

JAEA-Data/Code 2023-006, 24 Pages, 2023/07

JAEA-Data-Code-2023-006.pdf:1.42MB

Core fuels with stainless steel cladding and high plutonium content mixed oxide (MOX) fuel in a water-cooled environment, such as supercritical water-cooled reactors (SCWR) and reduced-moderation water reactors (RMWR), have been studied. In order to contribute to the research and development of such a core fuel concept, the fuel performance code "FEMAXI-8" was verified based on the results of post irradiation examinations of MOX fuel irradiated in the experimental fast reactor "JOYO". FEMAXI-8 is the latest version of the behavior analysis code developed by JAEA to analyze the behavior of light water reactor fuels under normal operation and transient conditions. This latest code has been improved and developed to allow the selection of stainless steel cladding property models to analyze improved fuels such as accident tolerant fuels. The purpose of this report is to confirm the prediction accuracy of FEMAXI-8 for the irradiation behavior of the new type of core fuel that is currently being developed. As a result of the verification, it was confirmed that FEMAXI-8 has sufficient analysis accuracy for the irradiation behavior of sodium-cooled fast reactor MOX fuel with stainless steel cladding, which exceeds the plutonium content and irradiation conditions of light water reactors. In the future, the analysis accuracy of FEMAXI-8 could be improved by adopting the O/M ratio dependence of MOX fuel thermal conductivity and the irradiation behavior evaluation model at high temperature.

Influence of plutonium content in dissolver solutions derived from irradiated fast reactor fuels on plutonium stripping in multistage countercurrent liquid-liquid extraction with acid split flowsheet

Nakahara, Masaumi; Shibata, Atsuhiro

Journal of Nuclear Science and Technology, 60(7), p.849 - 858, 2023/07

https://doi.org/10.1080/00223131.2022.2162137

To develop the acid split method which has highly nuclear proliferation resistance, influence of Pu content in dissolver solutions derived from irradiated fast reactor fuel on the Pu stripping was investigated in experiments and a calculation. The Pu content in the U/Pu and U products increased with increasing the Pu content in the dissolver solution. Moreover, the calculated results indicate that the Pu leakage into the U product is suppressed with the Pu stripping solution only at low temperature.

Investigation of the core neutronics analysis conditions for evaluation of burn-up nuclear characteristics of the next-generation fast reactors

Takino, Kazuo; Oki, Shigeo

JAEA-Data/Code 2023-003, 26 Pages, 2023/05

JAEA-Data-Code-2023-003.pdf:1.66MB

Since next-generation fast reactors aim to achieve a higher core discharge burn-up than conventional reactors do, core neutronics design methods must be refined. Therefore, a suitable analysis condition is required for the analysis of burn-up nuclear characteristics to accomplish sufficient estimation accuracy while maintaining a low computational cost. We investigated the effect of the analysis conditions on the accuracy of estimation of the burn-up nuclear characteristics of next-generation fast reactors in terms of neutron energy groups, neutron transport theory, and spatial mesh. This study treated the following burn-up nuclear characteristics: criticality, burn-up reactivity, control rod worth, breeding ratio, assembly-wise power distribution, maximum linear heat rate, sodium void reactivity, and Doppler coefficient for the equilibrium operation cycle. As a result, it was found that the following conditions were the most suitable: 18-energy-group structure, 6 spatial meshes per assembly with diffusion approximation. Additionally, these conditions should apply to correction factors for energy group structure, spatial mesh and transport effects.

Validation of feedback reactivity evaluation models for plant dynamics analysis code during unprotected loss of heat sink event in sodium-cooled fast reactors

Yoshimura, Kazuo; Doda, Norihiro; Igawa, Kenichi*; Tanaka, Masaaki; Yamano, Hidemasa

Journal of Nuclear Engineering and Radiation Science, 9(2), p.021601_1 - 021601_9, 2023/04

https://doi.org/10.1115/1.4055328

Feedback reactivity automatically caused by radial expansion of the core is known as one of the inherent safety features in a sodium-cooled fast reactor (SFR). In order to validate the evaluation models of the reactivity feedback equipped in the in-house plant dynamics analysis code named Super-COPD, the benchmark analyses for the unprotected loss of heat sink (ULOHS) tests of BOP-302R and BOP-301 in an experimental SFR, EBR-II were conducted and the applicability of the evaluation method for the reactivity feedback was indicated during the ULOHS even, by comparing the numerical results and the experimental data.

Study on actinide burning core concepts for the future phaseout of a fast reactor fuel cycle

Mori, Tetsuya; Naganuma, Masayuki; Oki, Shigeo

Nuclear Technology, 209(4), p.532 - 548, 2023/04

https://doi.org/10.1080/00295450.2022.2133514

This paper deals with a conceptual study on a plutonium (Pu) and minor actinide (MA) burning fast reactor core for the distant future phaseout of a fast-reactor fuel cycle after it is commercialized and used for a long time. This burning core aims to reduce the Pu and MA inventories contained in the fuel cycle through multiple recycling. A key point for the core design is the degradation of Pu and MA during multiple recycling. This degradation affects the core feasibility by increasing the sodium void reactivity and decreasing the absolute value of the Doppler constant. A feasible core concept was found by incorporating the following three factors to improve the reactivity coefficients: core flattening, fuel burnup reduction, and the use of silicon carbide (SiC) in the cladding and wrapper tubes. Notably, softening the neutron spectrum using the SiC structural material not only improved the reactivity coefficients but also indirectly mitigated the degradation of Pu and MA. Consequently, the designed core allowed for multiple recycling to continue until the Pu and MA reduced significantly, particularly by about 99% in a phaseout scenario starting from a fast-reactor fleet of 30-GWe nuclear power capacity. Fast reactors were found to have the potential to become self-contained energy systems that can minimize the inventories of Pu they produced themselves, as well as long-lived MA. Fast reactors can be among the important options for environmental burden reduction in the future.

Applicability evaluation of Akaike's Bayesian information criterion to covariance modeling in the cross-section adjustment method

Maruyama, Shuhei; Endo, Tomohiro*; Yamamoto, Akio*

EPJ Web of Conferences, 281, p.00008_1 - 00008_9, 2023/03

https://doi.org/10.1051/epjconf/202328100008

The applicability of Akaike's Bayesian Information Criterion (ABIC) to covariance modeling in the cross-section adjustment method was investigated. One of the most important things for a reliable cross-section adjustment method is giving a suitable covariance matrix. However, since we cannot know the true covariance matrix in advance, we usually estimate and assume it. To judge the goodness of the covariance matrix modeling, a metric is desirable. As a candidate for this metric, we focus on ABIC which is one of the information criteria in Bayesian inference, because the cross-section adjustment method is often discussed within the framework of Bayesian inference. In the conventional cross-section adjustment method, incorporation of the analysis model uncertainty in a covariance matrix still requires ad hoc treatment. In JAEA, the integral experimental database for fast reactors has been developed and the adjusted cross-section set ADJ2017 has been created based on this database. Many of the core characteristics in the database have been analyzed by a deterministic method. Therefore, the predicted core characteristics have non-negligible uncertainties with correlations due to some numerical approximations. However, the evaluations of the uncertainties and their correlations are still challenging issues. In addition, there would be unknown uncertainties that experimenters and analysts of reactor physics experiments could not recognize. To judge the goodness of the covariance matrix related to these uncertainties, the applicability of ABIC to the cross-section adjustment method was investigated.

Development of a robust nuclear data adjustment method to outliers

Fukui, Yuhei*; Endo, Tomohiro*; Yamamoto, Akio*; Maruyama, Shuhei

EPJ Web of Conferences, 281, p.00006_1 - 00006_9, 2023/03

https://doi.org/10.1051/epjconf/202328100006

We developed a new nuclear data adjustment method for experimental data containing outliers. This method mitigates the effect of outliers by applying M-estimation, a type of robust estimation, to the conventional nuclear data adjustment method using sensitivity coefficients. Based on the M-estimation, we derived a weighted nuclear data adjustment formula and developed a weight calculation method. The weighted nuclear data adjustment formula was derived by weighting the function to take the extremum of the conventional nuclear data adjustment. The weighting of each nuclear characteristic is calculated from the difference between the measured and calculated values of the nuclear characteristic. This weight calculation method can evaluate the validity of each nuclear characteristic by considering correlations between nuclear characteristics using singular value decomposition. The proposed method and the conventional method were compared and verified by twin experiments. In the twin experiments, the nuclear data were adjusted using experimental data that intentionally included outliers. As a result of twin experiments, it was confirmed that the nuclear data were adjusted robustly and appropriately even with the experimental data containing outliers.

Chapter 5, Sodium-cooled Fast Reactor (SFRs)/ Chapter 12, Generation-IV Sodium-cooled Fast Reactor (SFR) concepts in Japan

Kubo, Shigenobu; Chikazawa, Yoshitaka; Ohshima, Hiroyuki; Kamide, Hideki

Handbook of Generation IV Nuclear Reactors, Second Edition, p.173 - 194, 2023/03

Handbook of Generation IV Nuclear Reactors, Second Edition is a fully revised and updated comprehensive resource on the latest research and advances in generation IV nuclear reactor concepts. Editor Igor Pioro and his team of expert contributors have updated every chapter to reflect advances in the field since the first edition published in 2016. JAEA contributes to Chapter 5; Sodium-cooled Fast Reactors (SFRs) and Chapter 12; Generation-IV Sodium-cooled Fast Reactor (SFR) concepts in Japan. Major characteristics and current technology developments including safety enhancement were described in Chapter 5. Chapter 12 shows design activities of SFR. Innovative technology developments, and update of the Japan sodium-cooled fast reactor design with lessons learned from the TEPCO Fukushima Daiichi NPP accident.

Measurement of void fraction distribution in a sphere-packed bed using X-ray imaging

Yamamoto, Seishiro*; Odaira, Naoya*; Ito, Daisuke*; Ito, Kei*; Saito, Yasushi*; Imaizumi, Yuya; Matsuba, Kenichi; Kamiyama, Kenji

Konsoryu, 37(1), p.79 - 85, 2023/03

High temperature mechanical properties and microstructure in 9Cr or 12Cr oxide dispersion strengthened steels

Mitsuhara, Masatoshi*; Kurino, Koichi*; Yano, Yasuhide; Otsuka, Satoshi; Toyama, Takeshi*; Onuma, Masato*; Nakashima, Hideharu*

Tetsu To Hagane, 109(3), p.189 - 200, 2023/03

https://doi.org/10.2355/tetsutohagane.TETSU-2022-080

Oxide Dispersion Strengthened (ODS) ferritic steel, a candidate material for fast reactor fuel cladding, has low thermal expansion, good thermal conductivity, and excellent resistance to irradiation damage and high temperature strength. The origin of the excellent high-temperature strength lies in the dispersion of fine oxides. In this study, creep tests at 700 or 750C, which are close to the operating temperatures of fast reactors, and high-temperature tensile tests at 900 to 1350 C, which simulate accident conditions, were conducted on 9Cr ODS ferritic steels, M11 and MP23, and 12Cr ODS ferritic steel, F14, to confirm the growth behavior of oxides. In the M11 and F14 creep test samples, there was little oxide growth or decrease in number density from the initial state, indicating that dispersion strengthening by oxides was effective during deformation. After creep deformation of F14, the development of dislocation substructures such as dislocation walls and subgrain boundaries was hardly observed, and mobile dislocations were homogeneously distributed in the grains. The dislocation density increased with increasing stress during the creep test. In the high-temperature ring tensile tests of MP23 and F14, the strength of both steels decreased at higher temperatures. In MP23, elongation decreased with increasing test temperature from 900 to 1100 C, but increased at 1200 C, decreased drastically at 1250 C, and increased again at 1300 C. In F14, elongation decreased with increasing temperature. It was inferred that the formation of the -ferrite phase was responsible for this complex change in mechanical properties of MP23 from 1200 to 1300 C.