Conceptual design of pilot ADS

Nishihara, Kenji; Fukushima, Masahiro; Abe, Takumi; Katano, Ryota; Yee-Rendon, B.; Iwamoto, Hiroki; Sugawara, Takanori; Obayashi, Hironari; Saito, Shigeru

JAEA-Research 2025-013, 125 Pages, 2026/03

JAEA-Research-2025-013.pdf:4.68MB

A conceptual design for a pilot Accelerator Driven subcritical System (ADS) was developed as a precursor to a commercial ADS aimed at partitioning and transmutation of minor actinides. The output of the pilot ADS was set at 200 MW. Based on safety assessment results, the design incorporates deep subcriticality and safety rods. Core design, accelerator design, target design, and in-vessel equipment design were performed, clarifying the specific concept.

The JUPITER experiments; HIGH-240 plutonium metal plates separated by lead and reflected by copper

Bess, J. D.*; Shepherd, D.*; McSpaden, A. T.*; Fukushima, Masahiro; Oizumi, Akito

International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook (2025 edition); NEA/NSC/DOC(95)03/I, Vol.I; PU-MET-FAST-050 (Internet), 544 Pages, 2026/02

This report presents a series of experiments conducted to obtain integral benchmark data for lead cross sections, which are essential for the design of Accelerator-Driven Systems (ADS). In this experimental campaign, systematic data were acquired using various fuel compositions, including highly enriched uranium, low-enriched uranium, and plutonium, under fast neutron spectrum conditions moderated by lead. All experiments were jointly conducted by JAEA and Los Alamos National Laboratory (LANL) at the National Criticality Experiments Research Center (NCERC). Two experimental cores for plutonium fuel were constructed. The Jupiter core used plutonium metal fuel containing 95 wt.% Pu and 4.5 wt.% Pu, while the High-240 Pu Jupiter core employed plutonium metal fuel with a higher Pu content, 72.8 wt.% Pu and 22 wt.% Pu. This benchmark report summarizes the results of High-240 Pu Jupiter experiments for registration in the International Criticality Safety Benchmark Evaluation Project (ICSBEP).

Design policy of pilot plant for accelerator-driven system

Nishihara, Kenji; Sugawara, Takanori; Fukushima, Masahiro; Iwamoto, Hiroki; Katano, Ryota; Abe, Takumi

Progress in Nuclear Science and Technology (Internet), 8, p.185 - 190, 2025/09

https://doi.org/10.15669/pnst.8.185

A pilot plant for the accelerator-driven system is proposed as a scaled-down version of a lead-bismuth cooled ADS with 800 MW thermal output for transmutation of minor actinides. In this article, the design policy of the pilot plant is elaborated for each design area: safety, subcritical management, sub-critical core, accelerator, target and in-core components. The conceptual design of the pilot ADS is also presented: 200 MW of thermal output, keff below 0.95, introduction of control rod, 1.0 GeV-10mA accelerator and roughly-estimated size of in-core components.

Uncertainty quantification of Np, Am, and Am reaction rates in highly enriched uranium fuel cores at Kyoto University Critical Assembly

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

Nuclear Science and Engineering, 199(3), p.429 - 444, 2025/03

https://doi.org/10.1080/00295639.2024.2380624

Experimental analyses of neptunium-237 (Np), americium-241 (Am), and Am fission and Np capture reaction rates are conducted by the Serpent 2 code together with ENDF/B-VIII.0 and JENDL-5, using experimental data at neutron spectra of thermal and intermediate regions obtained in the solid-moderated and solid-reflected cores with highly-enriched uranium fuel at the Kyoto University Critical Assembly. Also, uncertainty quantification of fission and capture reaction rate ratios of test samples of Np, Am and Am with reference samples of uranium-235 (U) and gold-197 (Au) are evaluated by the MARBLE code system. In terms of fission reaction rate ratios of Np/U, Am/U and Am/U, a comparison between experiments and Serpent 2 calculations shows an accuracy about 5, 15 and 10%, respectively, together with ENDF/B-VIII.0 and JENDL-5. For capture reaction rate ratios of Np/Au, Serpent 2 calculations reveal a fairly good accuracy at the thermal neutron spectrum. The total uncertainties of Np/U, Am/U and Am/U fission reaction rate ratios by MARBLE with the covariance data of ENDF/B-VIII.0 and JENDL-5 are found to be about 4% at most in all cores, except for about 8% of Am/U with ENDF/B-VIII.0 at the intermediate neutron spectrum.

Identification of effective integral experiments for nuclear data-induced uncertainty reduction through sparse modeling

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

Nuclear Science and Engineering, 11 Pages, 2025/00

https://doi.org/10.1080/00295639.2025.2586196

Proposed here is a novel method involving integral experiments effective in reducing nuclear-data-induced uncertainty. The method is formulated by the Extended Bias Factor method and the L1-norm-based sparse modeling to address computational difficulty in combinational optimization. The pseudo-design parameters only sensitive to specific microscopic reactions are defined for the validation of nuclear data. The method is applied to two pseudo parameters: neptunium-237 capture and bismuth-209 inelastic scattering cross sections, while considering integral experimental data used in ADJ2017 together with sample worth measurements made at the Kyoto University Critical Assembly. The results indicate that the proposed method successfully identifies a small subset of effective integral experiments.

ZEUS: Fast-spectrum critical assembly with a mixed core of highly enriched and natural uranium containing lead surrounded by a copper reflector

Oizumi, Akito; Fukushima, Masahiro; Gunji, Satoshi; McKenzie, G.*; Amundson, K.*

International Criticality Safety Benchmark Evaluation Project (ICSBEP) Handbook (2022/23 edition) (Internet) , 313 Pages, 2024/11

This benchmark report was compiled to register a critical experiment using the lower-enriched uranium (LEU) system core to the International Criticality Safety Evaluation Project (ICSBEP). The LEU experiment was one of a series of joint experimental project with the Los Alamos Laboratory in the United States from 2015 to 2019 aimed at improving the design accuracy of the accelerator driven system (ADS). This core was loaded alternating highly-enriched uranium (HEU) and natural uranium (NU) to simulate LEU. In addition, a fast neutron spectrum system was constructed with not only HEU and NU but also lead which is part of coolant in the ADS. In this evaluation, it was clarified that the experimental uncertainty for the effective multiplication factor was almost 100 pcm. Moreover, the C/E-1 values of almost -70 pcm and -145 pcm were obtained by the calculation with the continuous energy Monte Carlo code MCNP and the nuclear data ENDF/B-VIII.0 and JENDL-4.0, respectively.

Impact of uncertainty reduction on lead-bismuth coolant in accelerator-driven system using sample reactivity experiments

Katano, Ryota; Oizumi, Akito; Fukushima, Masahiro; Pyeon, C. H.*; Yamamoto, Akio*; Endo, Tomohiro*

Nuclear Science and Engineering, 198(6), p.1215 - 1234, 2024/06

https://doi.org/10.1080/00295639.2023.2246779

In this study, we have demonstrated that data assimilation using lead and bismuth sample reactivities measured in the Kyoto University Critical Assembly A-core can successfully reduce the uncertainty of the coolant void reactivity in accelerator-driven systems derived from inelastic-scattering cross-sections of lead and bismuth. We re-evaluated and highlighted the experimental uncertainties and correlations of the sample reactivities for the data assimilation formula. We used the MCNP6.2 code to evaluate the sample reactivities and their uncertainties, and performed data assimilation using the reactor analysis code system MARBLE. The high-sensitivity coefficients of the sample reactivities to lead and bismuth allowed us to reduce the cross-section-induced uncertainty of the void reactivity of the accelerator-driven system from 6.3% to 4.8%, achieving a provisional target accuracy of 5% in this study. Furthermore, we demonstrated that the uncertainties arising from other dominant factors, such as minor actinides and steel, can be effectively reduced by using integral experimental data sets for the unified cross-section dataset ADJ2017.

JENDL-5 benchmarking for fission reactor applications

Tada, Kenichi; Nagaya, Yasunobu; Taninaka, Hiroshi; Yokoyama, Kenji; Okita, Shoichiro; Oizumi, Akito; Fukushima, Masahiro; Nakayama, Shinsuke

Journal of Nuclear Science and Technology, 61(1), p.2 - 22, 2024/01

https://doi.org/10.1080/00223131.2023.2197439

The new version of the Japanese evaluated nuclear data library, JENDL-5, was released in December 2021. This paper demonstrates the validation of JENDL-5 for fission reactor applications. Benchmark calculations are performed with the continuous-energy Monte Carlo codes MVP and MCNP and the deterministic code system MARBLE. The benchmark calculation results indicate that the performance of JENDL-5 for fission reactor applications is better than that of the former library JENDL-4.0.

Japanese Evaluated Nuclear Data Library version 5; JENDL-5

Iwamoto, Osamu; Iwamoto, Nobuyuki; Kunieda, Satoshi; Minato, Futoshi; Nakayama, Shinsuke; Abe, Yutaka*; Tsubakihara, Kosuke*; Okumura, Shin*; Ishizuka, Chikako*; Yoshida, Tadashi*; et al.

Journal of Nuclear Science and Technology, 60(1), p.1 - 60, 2023/01

https://doi.org/10.1080/00223131.2022.2141903

Measurement of Np and Am fission reaction rates in lead region at A-core of KUCA

Oizumi, Akito; Katano, Ryota; Kojima, Ryohei; Fukushima, Masahiro; Tsujimoto, Kazufumi; Pyeon, C. H.*

KURNS Progress Report 2020, P. 104, 2021/08

In the nuclear transmutation system such as ADS, the nuclear data validation of MA is required to reduce the uncertainty caused by the nuclear data of MA. This study aims to measure the fission reaction rate ratios (FRRRs) of Neptunium-237 (Np) or Americium-243 (Am) to Uranium-235 (U) by using a single fission chambers in the KUCA. The results showed that the measured FRRRs of Np/U and Am/U were 0.0480.003 and 0.0420.004, respectively. The measured values will be used for verification of evaluated nuclear data by conducting detailed analyses.