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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

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:7 Percentile:94.84(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.

JAEA Reports

Neutron flux estimation and neutronics characteristics calculation in post-JMTR conceptual study

Oizumi, Akito; Akie, Hiroshi

JAEA-Technology 2023-017, 93 Pages, 2023/12

JAEA-Technology-2023-017.pdf:8.45MB

After the decision of decommissioning JMTR (Japan Materials Testing Reactor), Japan Atomic Energy Agency investigated the possibility to construct a new irradiation test reactor to succeed JMTR (post-JMTR), and the final report of the investigated result was submitted to the Ministry of Education, Culture, Sports, Science and Technology on March 30th 2021. This investigation was carried out in 4 steps of (1) selection of reactor type, (2) reactor core plans studies, (3) neutronic studies, (4) thermal studies, and was finally (5) considered and evaluated. This JAEA-Technology report summarizes the process and the results of (3) neutronic studies. Neutron fluxes were calculated at irradiation sample positions in the investigated cores, the standard core and the compact core, and the calculated fluxes satisfied the required irradiation capability. It was also evaluated the two investigated cores' continuous reactor operation time in days in one refueling cycle, and the results guaranteed an operation days equality with that of existing JMTR. In addition, neutronic characteristics of the cores were estimated, such as power distribution in the core, control rod reactivity worth, reactivity coefficients, distribution of fuel burnup rate of each fuel element, and kinetics parameters. The evaluated neutronic characteristics were used in the post-JMTR final investigation report to confirm the neutronic feasibility by comparing with the neutronic limiting values of existing JMTR, and to estimate the cooling capability to make the core thermally feasible.

Journal Articles

Misuse scenario analysis of accelerator-driven system and the effects of dual C/S and design information verification

Oizumi, Akito; Sagara, Hiroshi*

Dai-44-Kai Nihon Kaku Busshitsu Kanri Gakkai Nenji Taikai Kaigi Rombunshu (Internet), 4 Pages, 2023/11

Research and development of transuranium (TRU) fuel cycles with accelerator drive systems (ADSs) transmuting minor actinides separated from the commercial cycles has been continuously conducted to reduce the high-level radioactive waste contained in spent fuel discharged from nuclear power plants. Since ADS could be misused to illegally produce Pu by using neutrons generated by the accelerator, a different approach from a conventional nuclear reactor would be needed. In this study, we have analyzed possible misuse scenarios of ADS quantitatively evaluated Pu that can be illegally produced within the design tolerance of ADS, and evaluated the effects of the Dual Containment and Surveillance(C/S) and the design information verification methods. As a result, it was quantitatively clarified that 10-60 kg of Pu could be generated clandestinely, and the dual C/S and design information verification with monitoring of the operation history of both accelerators and reactors could detect and prevent all the misuse scenarios effectively.

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:65.72(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

Comparison of calculated bare critical masses between two versions of the Japanese Evaluated Nuclear Data Library, JENDL-5 and JENDL-4.0

Oizumi, Akito

Proceedings of 12th International Conference on Nuclear Criticality Safety (ICNC2023) (Internet), 10 Pages, 2023/10

Research and development of the transuranium fuel cycle with accelerator-drive systems (ADSs) transmuting minor actinides (MAs) separated from commercial cycles has been continuously conducted by the Japan Atomic Energy Agency (JAEA) to reduce the high-level radioactive waste contained in the spent fuel discharged from nuclear power plants. To transmute MA with high efficiency, the ADS fuel contains a large amount of MA. The criticality safety design of facilities within the ADS cycle must rely on calculated values using the nuclear data library because there are no experimental or measured values for the critical mass of many MAs. Thus, it is crucial to figure out the impact of updating the evaluated nuclear data library on the critical mass calculation. This study compared the differences in the calculated critical masses of a bare metal sphere (BCMs) for each actinide isotope between two version of Japanese Evaluated Nuclear Data Library, JENDL-5 (released in December 2021) and JENDL-4.0 as a basic assessment. The study found that the differences in BCMs between JENDL-5 and JENDL-4.0 were less than 1% for $$^{233}$$U, $$^{235}$$U, $$^{237}$$Np, and $$^{239}$$Pu, which have the integral experimental data for metallic spheres registered in the International Criticality Safety Benchmark Evaluation Project (ICSBEP). On the other hands, the difference in BCMs between two nuclear data libraries was found to be almost 7-40% for nuclides such as $$^{241}$$Am, $$^{243}$$Am, $$^{244}$$Cm, and $$^{246}$$Cm, which have relatively limited or no integral experimental data registered in ICSBEP and International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP). Furthermore, as a result of analyzing the nuclear data that influenced the difference in BCM, for example in the case of $$^{244}$$Cm, it was clarified that the update of fission reaction and prompt $$nu$$ gave a significant contribution.

Journal Articles

Cost-reduced depletion calculation including short half-life nuclides for nuclear fuel cycle simulation

Okamura, Tomohiro*; Katano, Ryota; Oizumi, Akito; Nishihara, Kenji; Nakase, Masahiko*; Asano, Hidekazu*; Takeshita, Kenji*

Journal of Nuclear Science and Technology, 60(6), p.632 - 641, 2023/06

 Times Cited Count:3 Percentile:52.93(Nuclear Science & Technology)

The Okamura explicit method (OEM) for depletion calculation was developed by modifying the matrix exponential method for dynamic nuclear fuel cycle simulation. The OEM suppressed the divergence of the calculation for short half-life nuclides, even for long time steps. The computational cost of the OEM was small, equivalent to the Euler method, and it maintained sufficient accuracy for the fuel cycle simulation.

Journal Articles

General-purpose nuclear data library JENDL-5 and to the next

Iwamoto, Osamu; Iwamoto, Nobuyuki; Kunieda, Satoshi; Minato, Futoshi; Nakayama, Shinsuke; Kimura, Atsushi; Nakamura, Shoji; Endo, Shunsuke; Nagaya, Yasunobu; Tada, Kenichi; et al.

EPJ Web of Conferences, 284, p.14001_1 - 14001_7, 2023/05

 Times Cited Count:1 Percentile:77.10(Nuclear Science & Technology)

Journal Articles

Measurement of double-differential neutron yields for iron, lead, and bismuth induced by 107-MeV protons for research and development of accelerator-driven systems

Iwamoto, Hiroki; Nakano, Keita; Meigo, Shinichiro; Satoh, Daiki; Iwamoto, Yosuke; Sugihara, Kenta*; Nishio, Katsuhisa; Ishi, Yoshihiro*; Uesugi, Tomonori*; Kuriyama, Yasutoshi*; et al.

EPJ Web of Conferences, 284, p.01023_1 - 01023_4, 2023/05

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

For accurate prediction of neutronic characteristics for accelerator-driven systems (ADS) and a source term of spallation neutrons for reactor physics experiments for the ADS at Kyoto University Critical Assembly (KUCA), we have launched an experimental program to measure nuclear data on ADS using the Fixed Field Alternating Gradient (FFAG) accelerator at Kyoto University. As part of this program, the proton-induced double-differential thick-target neutron-yields (TTNYs) and cross-sections (DDXs) for iron, lead, and bismuth have been measured with the time-of-flight (TOF) method. For each measurement, the target was installed in a vacuum chamber on the beamline and bombarded with 107-MeV proton beams accelerated from the FFAG accelerator. Neutrons produced from the targets were detected with stacked, small-sized neutron detectors for several angles from the incident beam direction. The TOF spectra were obtained from the detected signals and the FFAG kicker magnet's logic signals, where gamma-ray events were eliminated by pulse shape discrimination. Finally, the TTNYs and DDXs were obtained from the TOF spectra by relativistic kinematics. The measured TTNYs and DDXs were compared with calculations by the Monte Carlo transport code PHITS with its default physics model of INCL version 4.6 combined with GEM and those with the JENDL-4.0/HE nuclear data library.

Journal Articles

Measurement of 107-MeV proton-induced double-differential thick target neutron yields for Fe, Pb, and Bi using a fixed-field alternating gradient accelerator at Kyoto University

Iwamoto, Hiroki; Nakano, Keita; Meigo, Shinichiro; Satoh, Daiki; Iwamoto, Yosuke; Sugihara, Kenta; Nishio, Katsuhisa; Ishi, Yoshihiro*; Uesugi, Tomonori*; Kuriyama, Yasutoshi*; et al.

Journal of Nuclear Science and Technology, 60(4), p.435 - 449, 2023/04

 Times Cited Count:3 Percentile:52.93(Nuclear Science & Technology)

Double-differential thick target neutron yields (TTNYs) for Fe, Pb, and Bi targets induced by 107-MeV protons were measured using the fixed-field alternating gradient accelerator at Kyoto University for research and development of accelerator-driven systems (ADSs) and fundamental ADS reactor physics research at the Kyoto University Critical Assembly (KUCA). Note that TTNYs were obtained with the time-of-flight method using a neutron detector system comprising eight neutron detectors; each detector has a small NE213 liquid organic scintillator and photomultiplier tube. The TTNYs obtained were compared with calculation results using Monte Carlo-based spallation models (i.e., INCL4.6/GEM, Bertini/GEM, JQMD/GEM, and JQMD/SMM/GEM) and the evaluated high-energy nuclear data library, i.e., JENDL-4.0/HE, implemented in the particle and heavy iontransport code system (PHITS). All models, including JENDL-4.0/HE, failed to predict high-energy peaks at a detector angle of 5$$^{circ}$$. Comparing the energy- and angle-integrated spallation neutron yields at energies of $$le$$20 MeV estimated using the measured TTNYs and the PHITS indicated that INCL4.6/GEM would be suitable for the Monte Carlo transport simulation of ADS reactor physics experiments at the 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:122 Percentile:99.98(Nuclear Science & Technology)

Journal Articles

Non-proliferation features in partitioning and transmutation cycle using accelerator-driven system, 3; Safeguards by design by using ${it Material Attractiveness}$ evaluation for TRU fuel cycle

Oizumi, Akito; Sagara, Hiroshi*

Dai-43-Kai Nihon Kaku Busshitsu Kanri Gakkai Nenji Taikai Kaigi Rombunshu (Internet), 4 Pages, 2022/11

Research and development of partitioning and transmutation cycle with accelerator drive systems (ADSs) transmuting minor actinides separated from the commercial cycles has been continuously conducted to reduce the high-level radioactive waste contained in spent fuel discharged from nuclear power plants. Since the chemical form and composition of the fuels are different from those of the current commercial cycles, it is necessary to examine the inspection goal of the safeguards and the design level of physical protections which are required for the partitioning and transmutation (P&T) cycle. In previous studies, the ${it Material Attractiveness}$ (${it Attractiveness}$) of the uranium (U) in the ADS fuel with a unique isotopic composition was evaluated as 2, the second highest on a 4-point scale, assuming state actors. In this study, reduction methods of potential nuclear proliferation were examined for the rationalization of the P&T cycle design considering nuclear non-proliferation. The amount of recovered U (RepU) added to the ADS fuel, which was required to increase the bare critical mass of U, was quantitatively evaluated as one of the reduction methods of potential nuclear proliferation risk. As a result, the addition of RepU, which was about 1.3- 2.7 times U in the ADS fuel, lowered the ${it Attractiveness}$ to 3 - 4. The rationalization of the P&T cycle design based on the safeguards by design can be expected by reviewing the U decontamination standards in the reprocessing steps of the commercial cycle based on these quantitative data.

Journal Articles

Measurement of 107-MeV proton-induced double-differential neutron yields for iron for research and development of accelerator-driven systems

Iwamoto, Hiroki; Nakano, Keita; Meigo, Shinichiro; Satoh, Daiki; Iwamoto, Yosuke; Ishi, Yoshihiro*; Uesugi, Tomonori*; Kuriyama, Yasutoshi*; Yashima, Hiroshi*; Nishio, Katsuhisa; et al.

JAEA-Conf 2022-001, p.129 - 133, 2022/11

For accurate prediction of neutronic characteristics for accelerator-driven systems (ADS) and a source term of spallation neutrons for reactor physics experiments for the ADS at Kyoto University Critical Assembly (KUCA), we have launched an experimental program to measure nuclear data on ADS using the Fixed Field Alternating Gradient (FFAG) accelerator at Kyoto University. As part of this program, the proton-induced double-differential thick-target neutron-yields (TTNYs) and cross-sections (DDXs) for iron have been measured with the time-of-flight (TOF) method. For each measurement, the target was installed in a vacuum chamber on the beamline and bombarded with 107-MeV proton beams accelerated from the FFAG accelerator. Neutrons produced from the targets were detected with stacked, small-sized neutron detectors composed of the NE213 liquid organic scintillators and photomultiplier tubes, which were connected to a multi-channel digitizer mounted with a field-programmable gate array (FPGA), for several angles from the incident beam direction. The TOF spectra were obtained from the detected signals and the FFAG kicker magnet's logic signals, where gamma-ray events were eliminated by pulse shape discrimination applying the gate integration method to the FPGA. Finally, the TTNYs and DDXs were obtained from the TOF spectra by relativistic kinematics.

Journal Articles

Material attractiveness evaluation of fuel assembly of accelerator-driven system for nuclear security and non-proliferation

Oizumi, Akito; Sugawara, Takanori; Sagara, Hiroshi*

Annals of Nuclear Energy, 169, p.108951_1 - 108951_9, 2022/05

 Times Cited Count:1 Percentile:19.69(Nuclear Science & Technology)

Research and development of the partitioning and transmutation (P&T) cycle with accelerator-drive systems (ADSs) transmuting minor actinides separated from the commercial cycles have been continuously conducted to reduce the amount of high-level radioactive waste contained in spent fuel discharged from nuclear power plants. Because the chemical form and composition of the fuels are different from those of the current commercial cycles, it is necessary to examine the inspection goal of the safeguards (SGs) and the design level of physical protections (PPs) that are required for the P&T cycle. In this study, the material attractiveness was evaluated assuming the theft or diversion of fuel assemblies from the fuel storage pool of the ADS facility in terms of nuclear security and non-proliferation. According to the results, quantitative components based on the fundamental fuel property were created as an important factor to decide the inspection goal for SGs and the design level for PPs required for the ADS facility. Additionally, the attractiveness of mixed oxide (MOX) fuel assemblies stored in the commercial boiling water reactor (BWR) facility was compared with that of the ADS. With regard to nuclear security, the ADS fuel was less attractive than the BWR MOX in every cycle. Regarding nuclear non-proliferation, the ADS fuel assembly had less attractive plutonium (Pu) than the BWR MOX, and the uranium (U) in the ADS fuel assembly was as attractive as (or slightly more attractive than) that of the BWR MOX owing to low spontaneous fission neutron. Furthermore, new issues were identified through this evaluation. With the current regulations, it was difficult to decide whether the ADS fuel before irradiation should be treated as fresh or spent, because the ADS fresh fuel contained more transuranium and rare earth than U and contained U whose main component was U-234 instead of U-238.

JAEA Reports

User manual of NMB4.0

Okamura, Tomohiro*; Nishihara, Kenji; Katano, Ryota; Oizumi, Akito; Nakase, Masahiko*; Asano, Hidekazu*; Takeshita, Kenji*

JAEA-Data/Code 2021-016, 43 Pages, 2022/03

JAEA-Data-Code-2021-016.pdf:3.06MB

The quantitative prediction and analysis of the future nuclear energy utilization scenarios are required in order to establish the advanced nuclear fuel cycle. However, the nuclear fuel cycle consists of various processes from front- to back-end, and it is difficult to analyze the scenarios due to the complexity of modeling and the variety of scenarios. Japan Atomic Energy Agency and Tokyo Institute of Technology have jointly developed the NMB code as a tool for integrated analysis of mass balance from natural uranium needs to radionuclide migration of geological disposal. This user manual describes how to create a database and scenario input for the NMB version 4.0.

Journal Articles

NMB4.0: Development of integrated nuclear fuel cycle simulation code

Okamura, Tomohiro*; Katano, Ryota; Oizumi, Akito; Nishihara, Kenji; Nakase, Masahiko*; Asano, Hidekazu*; Takeshita, Kenji*

Bulletin of the Laboratory for Advanced Nuclear Energy, 6, p.29 - 30, 2022/02

Takeshita Laboratory, Tokyo Institute of Technology, has been developing Nuclear Material Balance code version 4.0 (NMB4.0) in collaboration with Japan Atomic Energy Agency (JAEA). This report summarized the outline and functions of NMB4.0.

Journal Articles

Non-proliferation features in partitioning and transmutation cycle using accelerator-driven system, 2; Evaluation of ${it Material Attractiveness}$ of uranium in ADS fuel assembly

Oizumi, Akito; Sugawara, Takanori; Sagara, Hiroshi*

Dai-42-Kai Nihon Kaku Busshitsu Kanri Gakkai Nenji Taikai Kaigi Rombunshu (Internet), 4 Pages, 2021/11

Research and development of partitioning and transmutation cycle with accelerator drive systems (ADSs) transmuting minor actinides (MAs) separated from the commercial cycles has been continuously conducted to reduce the high-level radioactive waste (HLW) contained in spent fuel discharged from nuclear power plants. Since the chemical form and composition of the fuels are different from those of the current commercial cycles, it is necessary to examine the inspection goal of the safeguards (SGs) and the design level of physical protections (PPs) which are required for the partitioning and transmutation cycle. In this study, ${it Material Attractiveness (Attractiveness)}$ of the uranium (U) in the fuel assembly in the fuel storage pool in the ADS facility was evaluated and it was compared with the plutonium (Pu) in the MOX fuel assembly for a general boiling water reactor (BWR). As a result, it made clear that the U in the ADS fuel assembly had equal to or less attractive than the Pu in the BWR MOX fuel assembly. Moreover, a new issue has been extracted. It is difficult to determine whether the ADS fresh fuel should be considered as non-irradiated or irradiated fuel under the current regulatory standards because the ADS fresh fuel contains many MAs, rare-earths, and $$^{234}$$U rich U.

Journal Articles

NMB4.0: Development of integrated nuclear fuel cycle simulator from the front to back-end

Okamura, Tomohiro*; Katano, Ryota; Oizumi, Akito; Nishihara, Kenji; Nakase, Masahiko*; Asano, Hidekazu*; Takeshita, Kenji*

EPJ Nuclear Sciences & Technologies (Internet), 7, p.19_1 - 19_13, 2021/11

Nuclear Material Balance code version 4.0 (NMB4.0) has been developed through collaborative R&D between Tokyo Institute of Technology and JAEA. Conventional nuclear fuel cycle simulation codes mainly analyze actinides and are specialized for front-end mass balance analysis. However, quantitative back-end simulation has recently become necessary for considering R&D strategies and sustainable nuclear energy utilization. Therefore, NMB4.0 was developed to realize the integrated nuclear fuel cycle simulation from front- to back-end. There are three technical features in NMB4.0: 179 nuclides are tracked, more than any other code, throughout the nuclear fuel cycle; the Okamura explicit method is implemented, which contributes to reducing the numerical cost while maintaining the accuracy of depletion calculations on nuclides with a shorter half-life; and flexibility of back-end simulation is achieved. The main objective of this paper is to show the newly developed functions, made for integrated back-end simulation, and verify NMB4.0 through a benchmark study to show the computational performance.

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:12.48(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.

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