Fukaya, Yuji; Goto, Minoru; Ohashi, Hirofumi
Annals of Nuclear Energy, 181, p.109534_1 - 109534_10, 2023/02
Feasibility of reprocessing of High Temperature Gas-cooled Reactor (HTGR) spent fuel by existing Plutonium Uranium Redox EXtraction (PUREX) plant and technology has been investigated. The spent fuel dissolved solution includes approximately 3 times amount of uranium-235 and 1.5 times amount of protonium because of the 3 times higher burnup compared with that of Light Water Reactor (LWR). Then, the heavy metal of the spent fuel is planned to be diluted to 3.1 times by depleted uranium to satisfy the limitation of Rokkasho Reprocessing Plant (RRP) plant. In the present study, recoverability of uranium and plutonium with the dilution is confirmed by a simulation with a reprocessing process calculation code. Moreover, the case without the dilution from the economic perspective is investigated. As a result, the feasibility is confirmed without the dilution, and it is expected that the reprocessed amount is reduced to 1/3 compared with a diluted case even though the facility should be optimized from the perspective of mass flow and criticality.
Hong, Z.*; Pellegrini, M.*; Erkan, N.*; Liao, H.*; Yang, H.*; Yamano, Hidemasa; Okamoto, Koji*
Annals of Nuclear Energy, 180, p.109462_1 - 109462_9, 2023/01
A series of experiments were conducted using BC material and SUS304 tubes as a simulant of the real control rods. Reaction rate constant data in the 1450K-1500K range were obtained, and are consistent with the reference values. The reaction layer microstructure observation and the associated chemical composition analysis were also carried onto the experiment samples.
Matsumoto, Toshinori; Kawabe, Ryuhei*; Iwasawa, Yuzuru; Sugiyama, Tomoyuki; Maruyama, Yu
Annals of Nuclear Energy, 178, p.109348_1 - 109348_13, 2022/12
The Japan Atomic Energy Agency extended the applicability of their fuel-coolant interaction analysis code JASMINE to simulate the relevant phenomena of molten core in a severe accident. In order to evaluate the total coolability, it is necessary to know the mass fraction of particle, agglomerated and cake debris and the final geometry at the cavity bottom. An agglomeration model that considers the fusion of hot particles on the cavity floor was implemented in the JASMINE code. Another improvement is introduction of the melt spreading model based on the shallow water equation with consideration of crust formation at the melt surface. For optimization of adjusting parameters, we referred data from the agglomeration experiment DEFOR-A and the under-water spreading experiment PULiMS conducted by KTH in Sweden. The JASMINE analyses reproduced the most of the experimental results well with the common parameter set, suggesting that the primary phenomena are appropriately modelled.
Simanullang, I. L.*; Nakagawa, Naoki*; Ho, H. Q.; Nagasumi, Satoru; Ishitsuka, Etsuo; Iigaki, Kazuhiko; Fujimoto, Nozomu*
Annals of Nuclear Energy, 177, p.109314_1 - 109314_8, 2022/11
Yokoyama, Kaoru; Ohashi, Yusuke
Annals of Nuclear Energy, 175, p.109240_1 - 109240_7, 2022/09
Decommissioning is planned at nuclear facilities that have been discontinued. We examined the evaluation method of uranium radioactivity for concrete waste generated by the decommissioning of nuclear facilities. Since the peaks of Ac-228, Tl-208, and K- 40 are derived from concrete waste, it is difficult to distinguish the 1001 keV peak emitted from the uranium source. We have derived a formula to correct gamma rays from concrete and the environment, and the amount of uranium was quantified. When the weight of concrete waste is about 300 kg, if the weight of uranium is 3 g or more, it can be quantified within a relative error of about 30%. Measurement tests were performed using homogeneous simulated concrete waste. Since uranium contamination is on the concrete surface at the uranium processing facility and small chunks generated by scraping the concrete surface will be stored in a drum and measured, it seems that the test of homogeneous concrete reflects the actual waste.
Yamashita, Takuya; Sato, Takumi; Madokoro, Hiroshi; Nagae, Yuji
Annals of Nuclear Energy, 173, p.109129_1 - 109129_15, 2022/08
Ohira, Hiroaki*; Tanaka, Masaaki; Yoshikawa, Ryuji; Ezure, Toshiki
Annals of Nuclear Energy, 172, p.109075_1 - 109075_10, 2022/07
In order to evaluate the mist behavior in the cover gas region of Sodium-cooled Fast Reactors (SFRs) in good accuracy, turbulent model for Rayleigh-Bnard convection (RBC) was selected, and the Reynolds-averaged number density and momentum equations for mist behavior were developed and incorporated into the OpenFOAM code. In the first stage, the RBC in a simple parallel channel was calculated using Favre-averaged k- SST model. The average temperature and flow characteristics agreed well with results from DNS, LES, and experiments. Then the basic heat transfer experiment simulating the cover gas region of SFRs was calculated using this turbulent model and new mist models. The calculated average temperature distribution in the height direction and the mist mass concentration agreed well with the experimental results. We developed a method that could simulate the mist behavior in turbulent RBC environments and the cover gas region of SFRs with high accuracy.
Annals of Nuclear Energy, 171, p.109052_1 - 109052_8, 2022/06
The fracture threshold of the fuel decreases if the oxidized Zr alloy cladding is strongly constrained by the spacer grid during quenching in a loss-of-coolant accident. Therefore, the estimation of realistic levels of the axial constraint has been a subject of significant interest on fuel safety. In this study, a test assembly consisting of a PWR-type simulated fuel segment and a 33 grid piece was heated in steam, cooled, and quenched, and the axial constraint force on the fuel segment was measured. The constraint force of the Zircaloy grid gradually decreased with temperature. Once the Zircaloy grid was heated to 1060 K, the reduced constraint force had difficulty recovering, and thus the maximum constraint force during cooling and quenching was 10 N. The constraint force was clearly reduced at 1070 K during the tests with the Inconel grid. However, the reduced constraint force partially recovered during cooling. As a result, the maximum constraint force during cooling and quenching was 20 to 50 N for the Inconel grid. In conclusion, oxidation, ballooning, rupture, or eutectic formation would not generally cause an extremely strong constraint, as predicted by previous studies, at the grid position.
Kakiuchi, Kazuo; Amaya, Masaki; Udagawa, Yutaka
Annals of Nuclear Energy, 171, p.109004_1 - 109004_9, 2022/06
Oizumi, Akito; Sugawara, Takanori; Sagara, Hiroshi*
Annals of Nuclear Energy, 169, p.108951_1 - 108951_9, 2022/05
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.
Tuya, D.; Nagaya, Yasunobu
Annals of Nuclear Energy, 169, p.108919_1 - 108919_9, 2022/05
Estimating an effect of a perturbation in a fissile system on its -eigenvalue requires special technique called perturbation theory when the considered perturbation is small. In this study, we develop an adjoint-weighted correlated sampling (AWCS) method based on the exact perturbation theory without any approximation by combining the correlated sampling (CS) method with iterated-fission probability (IFP) based adjoint-weighting method. With the advantages of the CS method being good at providing very small uncertainty for small perturbations and the IFP-based adjoint-weighting method being suitable for continuous-energy Monte Carlo calculation, the developed AWCS method based on the exact perturbation theory offers a new rigorous approach for perturbation calculations. The obtained results by the developed AWCS method for verification problems involving Godiva and simplified STACY density perturbations showed good agreement with the reference calculations.
Kwon, Saerom*; Konno, Chikara; Ota, Masayuki*; Sato, Satoshi*
Annals of Nuclear Energy, 169, p.108932_1 - 108932_7, 2022/05
Recently, it was reported that one of three ENDF files of Be-9 in the TENDL-2017 alpha sub-library included strange neutron production data. Thus we have tested three ENDF files of Be-9 in the TENDL-2017 deuterium sub-library for nuclear designs of a new fusion neutron source A-FNS. As a result, we found out that neutron production cross sections and secondary neutron spectra were different among three ENDF files and specified reasons. We confirmed that the latest TENDL, TENDL-2019, still had some of the issues.
Ishigaki, Masahiro*; Abe, Satoshi; Hamdani, A.; Hirose, Yoshiyasu
Annals of Nuclear Energy, 168, p.108867_1 - 108867_20, 2022/04
Fukaya, Yuji; Okita, Shoichiro; Nakagawa, Shigeaki; Goto, Minoru; Ohashi, Hirofumi
Annals of Nuclear Energy, 168, p.108911_1 - 108911_7, 2022/04
A power distribution monitoring system by using a moving detector for a core with a long neutron flight path has been proposed. High Temperature Gas-cooled Reactor (HTGR) and Fast Reactor (FR) has a long neutron flight path and the neutrons reach to detector far from fuel assembly in the center of the core unlike Light Water Reactor (LWR). By using the feature, power distribution can be observed with a few detectors by moving the detector and computed tomography technology similar to X-ray Computed Tomography (CT). For a small-sized core, the power distribution can be evaluated only by an ex-core neutron detector. For a large-sized core with inner detectors, the power distribution can be observed with a small number of in-core detectors even if the deployment is limited due to material integrity conditions such as temperature environment. The feasibility is numerically confirmed by simulations of the HTGR core and its detector response. It is expected to observe the power distribution in the core of HTGR and FR, which is difficult continuously to deploy in-core detectors because of high temperature and/or high irradiation damage.
Segawa, Mariko; Toh, Yosuke; Kai, Tetsuya; Kimura, Atsushi; Nakamura, Shoji
Annals of Nuclear Energy, 167, p.108828_1 - 108828_5, 2022/03
Abe, Satoshi; Hamdani, A.; Ishigaki, Masahiro*; Shibamoto, Yasuteru
Annals of Nuclear Energy, 166, p.108791_1 - 108791_18, 2022/02
Shiotsu, Hiroyuki; Ito, Hiroto*; Sugiyama, Tomoyuki; Maruyama, Yu
Annals of Nuclear Energy, 163, p.108587_1 - 108587_9, 2021/12
Fukaya, Yuji; Goto, Minoru
Annals of Nuclear Energy, 164, p.108617_1 - 108617_6, 2021/12
A reasonable source term of a hypothetical criticality accident for fuel fabrication facility with solution system has been proposed. The public exposure must not exceed the limitation of 5 mSv during an accident. Then, we proposed the reasonable source term of the first burst peak due to the hydrogen gas generation by radiation decomposition of water. With the criticality control system composed of the Criticality Accident Alarm System (CAAS) and soluble neutron absorber, safety is ensured by the reduced fission number. We confirmed the effect by environmental impact assessment during a criticality accident by using site condition of a fuel fabrication facility in Tokai-mura, Japan. As a result, the public exposure is reduced at a site boundary from 68 mSv to 0.6 mSv under the current regulatory guideline.
Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*
Annals of Nuclear Energy, 162, p.108512_1 - 108512_10, 2021/11
The objectives of this study are as follows: to understand the characteristics, degree of passive safety features for heat removal were compared for RCCSs based on atmospheric radiation and based on atmospheric natural circulation under the same conditions. Next, simulations on accidental conditions, such as increasing average heat-transfer coefficient via natural convection due to natural disasters, were performed with STAR-CCM+, and methodology to control the amount of heat removal was discussed. As a result, a new RCCS based on atmospheric radiation is recommended because of the excellent degree of passive safety features/conditions, and the amount of heat removal by heat transfer surfaces which can be controlled. Finally, methodology to determine structural thickness of scaled-down heat removal test facilities for reproducing natural convection and radiation was developed, and experimental methods by using pressurized and decompressed chambers was also proposed.
Takino, Kazuo; Sugino, Kazuteru; Oki, Shigeo
Annals of Nuclear Energy, 162, p.108454_1 - 108454_7, 2021/11