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Journal Articles

Uncertainty quantification of $$^{237}$$Np, $$^{241}$$Am, and $$^{243}$$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

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

Experimental analyses of neptunium-237 ($$^{237}$$Np), americium-241 ($$^{241}$$Am), and $$^{243}$$Am fission and $$^{237}$$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 $$^{237}$$Np, $$^{241}$$Am and $$^{243}$$Am with reference samples of uranium-235 ($$^{235}$$U) and gold-197 ($$^{197}$$Au) are evaluated by the MARBLE code system. In terms of fission reaction rate ratios of $$^{237}$$Np/$$^{235}$$U, $$^{241}$$Am/$$^{235}$$U and $$^{243}$$Am/$$^{235}$$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 $$^{237}$$Np/$$^{197}$$Au, Serpent 2 calculations reveal a fairly good accuracy at the thermal neutron spectrum. The total uncertainties of $$^{237}$$Np/$$^{235}$$U, $$^{241}$$Am/$$^{235}$$U and $$^{243}$$Am/$$^{235}$$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 $$^{243}$$Am/$$^{235}$$U with ENDF/B-VIII.0 at the intermediate neutron spectrum.

Journal Articles

Experiments on criticality and reactivity worths in the FCA-XXII-1 assembly simulating highly enriched MOX fueled tight lattice LWR cores

Fukushima, Masahiro; Ando, Masaki; Nagaya, Yasunobu

Nuclear Science and Engineering, 199(1), p.18 - 41, 2025/01

 Times Cited Count:1 Percentile:68.64(Nuclear Science & Technology)

A series of integral experiments were conducted at FCA of JAEA, simulating LWR cores with a tight lattice cell of highly enriched MOX fuel containing more than 15% fissile plutonium. The three experimental configurations were constructed using foamed polystyrene with different void fractions to clarify the prediction accuracy of neutronic calculation codes and nuclear data among various neutron spectra. The nuclear characteristics measured in the experiments were criticality, moderator void reactivity worths, and sample reactivity worths. The preliminary analyses on experiments were conducted using a deterministic calculation code conventionally used for fast reactors with JENDL-4.0. Most reactivity worth calculations correlated well with the experimental values. Specifically for the softer neutron spectra configurations, the treatment of ultrafine energy groups obviously improved the prediction accuracy of the deterministic calculations. Furthermore, reference calculations were performed with MVP3 code by modeling the experimental setup in detail, confirming the validity of the deterministic calculations.

Journal Articles

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.

Journal Articles

Design policy of pilot plant for accelerator-driven system

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

Proceedings of International Conference on Nuclear Fuel Cycle (GLOBAL2024) (Internet), 4 Pages, 2024/10

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 presentation, the design policy of the pilot plant is presented.

Journal Articles

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

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

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.

Journal Articles

TRU oxide sample reactivity worths measured in the FCA-IX assemblies with systematically changed neutron energy spectra

Fukushima, Masahiro; Okajima, Shigeaki*; Mukaiyama, Takehiko*

Journal of Nuclear Science and Technology, 61(4), p.478 - 497, 2024/04

 Times Cited Count:3 Percentile:67.98(Nuclear Science & Technology)

A series of integral experiments was conducted to evaluate the fission and the capture cross- sections of transuranic (TRU) nuclides at the fast critical facility FCA of the Japan Atomic Energy Agency (JAEA). The experiments were carried out using seven uranium-fueled assemblies of the FCA. The neutron energy spectra of the core regions were adjusted so as to change from an intermediate neutron spectrum to a fast neutron spectrum on an assembly-by-assembly basis. The integral data measured with these experimental configurations provide some neutron energy characteristics: 1) fission rate ratios (FRRs) of $$^{237}$$Np, $$^{238}$$Pu, $$^{242}$$Pu, $$^{241}$$Am, $$^{243}$$Am, and $$^{244}$$Cm relative to $$^{239}$$Pu by using absolutely calibrated fission chambers, 2) small sample reactivity worths (SRWs) of $$^{237}$$Np, $$^{238}$$Pu, $$^{240}$$Pu, $$^{241}$$Am, and $$^{243}$$Am where oxide powders of around 15 to 20 grams were used, 3) criticalities, and 4) spectral indices such as fission rate ratios of $$^{238}$$U relative to $$^{235}$$U. In this paper, details of the SRW measurements are reported, and the latest Japanese Evaluated Nuclear Data Library JENDL-5 is tested by using the integral data obtained in systematically varied neutron energy spectra.

Journal Articles

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

 Times Cited Count:10 Percentile:96.64(Nuclear Science & Technology)

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.

Journal Articles

Experiments on central reaction rate ratios and fission distributions in the FCA-XXII-1 assembly simulating highly enriched MOX fueled tight lattice LWR cores

Fukushima, Masahiro; Ando, Masaki; Nagaya, Yasunobu

Nuclear Science and Engineering, 15 Pages, 2024/00

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

Journal Articles

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, 197(11), p.2902 - 2919, 2023/11

 Times Cited Count:2 Percentile:51.90(Nuclear Science & Technology)

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

Journal Articles

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

 Times Cited Count:180 Percentile:99.98(Nuclear Science & Technology)

Journal Articles

Experimental analyses of $$^{243}$$Am and $$^{235}$$U fission reaction rates at Kyoto University Critical Assembly

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

Nuclear Science and Engineering, 195(11), p.1144 - 1153, 2021/11

 Times Cited Count:1 Percentile:10.83(Nuclear Science & Technology)

Measurements of $$^{243}$$Am and $$^{235}$$U fission reaction rates are conducted with the use of two single fission chambers in the solid-moderated and solid-reflected core at the Kyoto University Critical Assembly (KUCA). Critical irradiation experiments of $$^{243}$$Am and $$^{235}$$U foils are carried out, and the measured result of $$^{243}$$Am/$$^{235}$$U is 0.0424 $$pm$$ 0.0019; also, calculation/experiment values between calculated (MCNP6.1 with JENDL-4.0, ENDF/B-VIII.0, and JEFF-3.3) and measured results of $$^{243}$$Am/$$^{235}$$U range among 0.93 $$pm$$ 0.04, 0.94 $$pm$$ 0.04, and 0.93 $$pm$$ 0.04, respectively. Through a comparison between the measured and calculated results, the $$^{243}$$Am fission cross-section data of the three major nuclear data libraries are successfully validated, demonstrating the same accuracy as that of previous minor actinide irradiation experiments at KUCA. Importantly, the comparison also provides the complemental data of integral experiments of $$^{243}$$Am fission reaction rates that confirm the accuracy of the $$^{243}$$Am fission cross-section data.

Journal Articles

Measurement of $$^{237}$$Np and $$^{243}$$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 ($$^{237}$$Np) or Americium-243 ($$^{243}$$Am) to Uranium-235 ($$^{235}$$U) by using a single fission chambers in the KUCA. The results showed that the measured FRRRs of $$^{237}$$Np/$$^{235}$$U and $$^{243}$$Am/$$^{235}$$U were 0.048$$pm$$0.003 and 0.042$$pm$$0.004, respectively. The measured values will be used for verification of evaluated nuclear data by conducting detailed analyses.

Journal Articles

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

 Times Cited Count:6 Percentile:58.63(Nuclear Science & Technology)

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 $$^{27}$$Al and $$^{209}$$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.

Journal Articles

Void reactivity measurements of lead and bismuth in the KUCA-A core

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

KURNS Progress Report 2020, P. 102, 2021/07

For the design study of ADS, integral experimental data of LBE is necessary to validate cross sections of lead (Pb) and bismuth (Bi). In this study, we conducted Pb and Bi void reactivity measurements using aluminum (Al) void space in Kyoto University Critical Assembly (KUCA). We found that the calculations overestimate the void reactivities of Pb and Bi by about 20 pcm.

JAEA Reports

Critical mass evaluation of minor actinides in aqueous solution; Data for criticality safety assessment of separation process

Morita, Yasuji; Fukushima, Masahiro; Kashima, Takao*; Tsubata, Yasuhiro

JAEA-Data/Code 2020-013, 38 Pages, 2020/09

JAEA-Data-Code-2020-013.pdf:1.94MB

Critical Masses of Cm, Am and the mixture were calculated in metal-water mixtures with water reflector as a basic data for criticality safety assessment of minor actinide separation process. In the mixture of Cm-244 and Cm-245, higher ratio of Cm-245 gives smaller critical mass, but the amount of Cm-245 in the critical mass can be obtained by concentration of Cm-245 in the Cm mixture without depending on the Cm-245 ratio. Critical mass of Cm isotope mixture with 30% Cm-245 was smaller than that of Pu isotope mixture in the practical reprocessing (71% Pu-239 + 17% Pu-240 + 12% Pu-241). When Cm is separated from other element including Am and the solution is concentrated, measure for the critical accident has to be taken. Critical mass of Am-242m is smaller than that of Cm-245, but the ratio of Am-242m in the Am contained in practical spent fuel is small enough, about several percent, and therefore the critical accident by Am does not have to be considered. That by the mixture of Am and Cm does not either.

Journal Articles

Sample worth measurements of lead and bismuth in low-enriched uranium region at A-core of KUCA for ADS

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

KURNS Progress Report 2019, P. 143, 2020/08

For the design study of ADS, integral experimental data of nuclear characteristics of LBE is necessary to validate cross sections of lead (Pb) and bismuth (Bi). The calculation agree with experiment for the Bi sample worth. On the other hand, the calculation overestimates for the Pb sample worth.

Journal Articles

Measurement of $$^{243}$$Am fission rates in low-enriched uranium region at A-core of KUCA

Oizumi, Akito; Fukushima, Masahiro; Tsujimoto, Kazufumi; Yamanaka, Masao*; Pyeon, C. H.*

KURNS Progress Report 2019, P. 14, 2020/08

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 (FRRs) of Americium-243 ($$^{243}$$Am) to Uranium-235 ($$^{235}$$U) by using a single fission chambers in the KUCA. The result showed that the measured FRR of $$^{243}$$Am/$$^{235}$$U were 0.042 $$pm$$0.002. These measured values will be used for verification of evaluated nuclear data by conducting detailed analyses.

Journal Articles

Special issue on accelerator-driven system benchmarks at Kyoto University Critical Assembly

Pyeon, C. H.*; Talamo, A.*; Fukushima, Masahiro

Journal of Nuclear Science and Technology, 57(2), p.133 - 135, 2020/02

 Times Cited Count:4 Percentile:95.79(Nuclear Science & Technology)

Journal Articles

Systematic measurements and analyses for lead void reactivity worth in a plutonium core and two uranium cores with different enrichments

Fukushima, Masahiro; Goda, J.*; Oizumi, Akito; Bounds, J.*; Cutler, T.*; Grove, T.*; Hayes, D.*; Hutchinson, J.*; McKenzie, G.*; McSpaden, A.*; et al.

Nuclear Science and Engineering, 194(2), p.138 - 153, 2020/02

 Times Cited Count:7 Percentile:54.72(Nuclear Science & Technology)

To validate lead (Pb) nuclear cross sections, a series of integral experiments to measure lead void reactivity worth was conducted systematically in three fast spectra with different fuel compositions on the Comet critical assembly of the National Criticality Experiments Research Center. Previous experiments in a high-enriched uranium (HEU)/Pb and a low-enriched uranium (LEU)/Pb systems had been performed in 2016 and 2017, respectively. A follow-on experiment in a plutonium (Pu)/Pb system has been completed. The Pu/Pb system was constructed using lead plates and weapons grade plutonium plates that had been used in the Zero Power Physics Reactor (ZPPR) of Argonne National Laboratory until the 1990s. Furthermore, the HEU/Pb system was re-examined on the Comet critical assembly installed newly with a device that can guarantee the gap reproducibility with a higher accuracy and precision, and then the experimental data was re evaluated. Using the lead void reactivity worth measured in these three cores with different fuel compositions, the latest nuclear data libraries, JENDL 4.0 and ENDF/B VIII.0, were tested with the Monte Carlo calculation code MCNP version 6.1. As a result, the calculations by ENDF/B-VIII.0 were confirmed to agree with lead void reactivity worth measured in all the cores. It was furthermore found that the calculations by JENDL 4.0 overestimate by more than 20% for the Pu/Pb core while being in good agreements for the HEU/Pb and LEU/Pb cores.

Journal Articles

Recent activities in the field of reactor physics

Fukushima, Masahiro; Tojo, Masayuki*

Journal of Nuclear Science and Technology, 56(12), p.1061 - 1062, 2019/12

 Times Cited Count:0 Percentile:0.00(Nuclear Science & Technology)

Reactor Physics that treat the essentials of how fission nuclear reactors work fundamentally has important roles on safe operations and design studies of various types of nuclear reactors. From the latest activities in the field of reactor physics, this report summarizes some outstanding researches and developments published in scientific journals including the Journal of Nuclear Science and Technology.

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