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.

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

https://doi.org/10.1080/00223131.2022.2131646

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.

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

https://doi.org/10.1051/epjconf/202328401023

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.

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, 21 Pages, 2023/04

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.

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, 20 Pages, 2023/00

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.

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

https://doi.org/10.11484/jaea-conf-2022-001

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.

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.

Non-proliferation features in partitioning and transmutation cycle using accelerator-driven system, 2; Evaluation of 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, 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 U rich U.

Experimental analyses of Am and 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

https://doi.org/10.1080/00295639.2021.1932220

Measurements of Am and 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 Am and U foils are carried out, and the measured result of Am/U is 0.0424 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 Am/U range among 0.93 0.04, 0.94 0.04, and 0.93 0.04, respectively. Through a comparison between the measured and calculated results, the 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 Am fission reaction rates that confirm the accuracy of the Am fission cross-section data.

Material attractiveness evaluation of actinides in Pyroprocessing Facility for partitioning and transmutation cycle

Oizumi, Akito; Sagara, Hiroshi*

Proceedings of INMM & ESARDA Joint Virtual Annual Meeting (Internet), 7 Pages, 2021/08

Research and development of partitioning and transmutation (P&T) 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. The Japan Atomic Energy Agency has proposed a pyrochemical process for reprocessing ADS spent fuel with high decay heat and radioactivity due to the large amount of MA. 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 P&T cycle. In this study, the material attractiveness was evaluated assuming the diversion of the Cd cathode and the nitride powder from the pyroprocessing in terms of non-proliferation. Additionally, they were compared with the material attractiveness of the MOX fuel assemblies (fresh and spent fuels) for a conventional boiling water reactor (BWR). The Cd cathode used to recover actinides from ADS spent fuel by molten salt electrolysis in the pyroprocessing facility of P&T cycle was less attractive than the MOX fuel assembly for the BWR because the Cd cathode included Pu having high decay heat. The nitride powder electrorefined from the ADS spent fuel was also less attractiveness than the MOX fuel assembly for the BWR because of the same reasons of the Cd cathode.