Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*
Annals of Nuclear Energy, 151, p.107867_1 - 107867_11, 2021/02
A new RCCS with passive safety features consists of two continuous closed regions. One is a region surrounding RPV. The other is a cooling region with heat transferred to the ambient air. The new RCCS needs no electrical or mechanical driving devices. We compared the RCCS using atmospheric radiation with that using atmospheric natural circulation in terms of passive safety features and control methods for heat removal. The magnitude relationship for passive safety features is heat conduction radiation natural convection. Therefore, the magnitude for passive safety features of the former RCCS can be higher than that of the latter RCCS. In controlling the heat removal, the former RCCS changes the heat transfer area only. On the other hand, the latter RCCS needs to change the chimney effect. It is necessary to change the air resistance in the duct. Therefore, the former RCCS can control the heat removal more easily than the latter RCCS.
Fukaya, Yuji; Ueta, Shohei; Goto, Minoru; Ohashi, Hirofumi
Annals of Nuclear Energy, 151( ), p.107937_1 - 107937_9, 2021/02
Feasibility study on Burnable Poison (BP) credit concept to High Temperature Gas-cooled Reactor (HTGR) fuel fabrication has been performed. By mixing BP into fuel material in the first place of fuel fabrication, criticality safety is ensured in the all fuel fabrication process even with high enrichment fuel such as 14 wt% used in commercial HTGR. However, the poison effect also prevents the criticality even in the HTGR core, and it may shorten cycle length and achievable burn-up of the core. Therefore, the effect is evaluated by whole core burn-up calculation. As a BP, boron, gadolinium, erbium, and hafnium are investigated. As a result, it is found that boron and gadolinium suit this concept and the 14 wt% fuel can be fabricated in the plant fabricating 9.9 wt% High Temperature engineering Test Reactor (HTTR) fuel. With the boron and gadolinium, the commercial HTGR fuel can be fabricated with the safety measure as same as Light Water Reactor (LWR) fuel facility to treat the fuel with the enrichment up to 5 wt%. Especially, gadolinium is significantly suitable to this concept due to the dependency to spectrum, and more enhanced safety measure is feasible as well.
Johnson, M.*; Journeau, C.*; Matsuba, Kenichi; Emura, Yuki; Kamiyama, Kenji
Annals of Nuclear Energy, 151, p.107881_1 - 107881_13, 2021/02
High-resolution X-ray imaging was employed at the JAEA MELT facility to visualize a kilogram-scale interaction between a jet of high temperature molten stainless steel and sodium. A novel software, SPECTRA, has been developed for the quantitative characterization of jet quenching and fragmentation. Tracking and 3D reconstruction of the melt phase traversing the imaging window enabled the detection of 72% of the debris mass recovered post-experiment. The rebounding of melt fragments confirmed a solid outer crust at the melt-coolant interface, while a thermal fragmentation event induced rapid vapor expansion. Jet fragmentation is best explained by the vaporization of coolant entrained within the melt jet generating an internal over-pressure sufficient for fragmentation of the crust. Thermal fragmentation produced a bimodal debris size distribution of coarse jet shells and finer fragments.
Zhang, T.*; Funakoshi, Kanji*; Liu, X.*; Liu, W.*; Morita, Koji*; Kamiyama, Kenji
Annals of Nuclear Energy, 150, p.107856_1 - 107856_10, 2021/01
Annals of Nuclear Energy, 149, p.107818_1 - 107818_7, 2020/12
Treatment of surplus plutonium has been one of the most important issues in the utilization of nuclear power in Japan. This study investigates a molten salt accelerator-driven system (ADS) to transmute transuranic (TRU) nuclides to address the issue. MARDS (Molten salt Accelerator Driven System) concept employs lead chloride (PbCl) as a fuel salt to achieve a hard spectrum. Since the fuel salt is used as a spallation target, a dedicated spallation target is not required in this concept. Furthermore, a beam window which is a boundary between an accelerator and subcritical core is designed to avoid touching the fuel salt. It mitigates the difficulties of the beam design for ADS. Neutronics calculation for the MARDS concept was performed for a condition of 400 MW thermal power with 800 MeV proton beam. The calculation results showed that the proton beam current was about 7 mA and about 4400 kg plutonium could be transmuted during 40-year operation.
Annals of Nuclear Energy, 144, p.107503_1 - 107503_7, 2020/09
Disposal scenario of High Level Waste (HLW) of Light Water Reactor (LWR) have been optimized to reduce waste volume and repository footprint in geological disposal. The optimization was performed with existing and demonstrated technology in the situation where the reprocessing will be delayed. In general, the scenario with Partitioning and Transmutation (P&T) is optimized to minimize waste package number generated in the situation where the spent fuel will be reprocessed immediately with the minimum cooling time. With considering the delay of reprocessing, it is found that the more simplified and effective optimization with the high-waste-loading glass and cold crucible induction melter technologies and without partitioning. The optimized case can achieve significant reduction of number of waste package generation and the repository footprint to half of those of non-P&T case with 100 years cooling.
Okada, Yuji; Amaya, Masaki
Annals of Nuclear Energy, 145, p.107539_1 - 107539_8, 2020/09
Chai, P.; Yamashita, Susumu; Yoshida, Hiroyuki
Annals of Nuclear Energy, 145, p.107606_1 - 107606_13, 2020/09
The eutectic reaction model in JUPITER code was validated against two series of experimental tests that performed by JAEA. An experiment that aimed to evaluate the eutectic reaction between Zircaloy and Stainless steel, was simulated by JUPITER code to validate its reliability on predicting the binary eutectic reaction phenomenon. A comparison of the simulation and experimental results demonstrates good agreement on the increase rate of the solution depth at various temperature environments. Another series of tests which aimed to predict the eutectic reaction between the control rod blade and channel box in BWR, were simulated by JUPITER code to test its applicability on predicting the eutectic reaction between multiple mixture components. Although the deviation could not be completely eliminated, the reaction performance in the experiment was reasonably reproduced. As a result, it could be concluded that JUPITER code is feasible to predict the eutectic reaction behavior in nuclear severe accident.
El-Asaad, H.*; Nagai, Haruyasu; Sagara, Hiroshi*; Han, C. Y.*
Annals of Nuclear Energy, 141, p.107292_1 - 107292_9, 2020/06
Atmospheric dispersion simulations can provide crucial information to assess radioactive plumes in the environment for nuclear emergency preparedness. However, it is a difficult and time-consuming task to make simulations assuming many possible scenarios and to derive data from a vast number of results. Therefore, an interface was developed to assist users in conveying characteristics of plumes from simulation results. The interface uses a large database that contains WSPEEDI-II simulations for the first 20-days of radioactive release from the Fukushima Daiichi Nuclear Power Plant, and it displays essential quantitative data to the user from the database. The user may conduct sensitivity analysis with the help of the interface by changing release condition to generate many different case scenarios.
Yokoyama, Kaoru; Ohashi, Yusuke
Annals of Nuclear Energy, 141, p.107299_1 - 107299_5, 2020/06
A large amount of general steel waste is generated during decommissioning and dismantling of nuclear facilities. Very low-contaminated radioactive waste, whose radioactivity is below clearance level, generated from the demolition process may be reused for general use. We examined the feasibility of the clearance verification system for uranium waste. The relative error of uranium determination was within 30% for 1 g of uranium when measuring steel materials (angle bar, channel steel, pipe steel, square steel tube, fragments of metal tube).
Aoki, Takeshi; Chirayath, S. S.*; Sagara, Hiroshi*
Annals of Nuclear Energy, 141, p.107325_1 - 107325_7, 2020/06
The proliferation resistance (PR) of an inert matrix fuel (IMF) in the transuranic nuclear fuel cycle (NFC) of a high temperature gas cooled reactor is evaluated relative to the uranium and plutonium mixed-oxide (MOX) NFC of a light water reactor using PRAETOR code and sixty-eight input attributes. The objective is to determine the impacts of chemical stability of IMF and fuel irradiation on the PR. Specific material properties of the IMF, such as lower plutonium content, carbide ceramics coating, and absence of U, contribute to enhance its relative PR compared to MOX fuel. The overall PR value of the fresh IMF (an unirradiated direct use material with a one-month diversion detection timeliness goal) is nearly equal to that of the spent MOX fuel (an irradiated direct use nuclear material with a three-month diversion detection timeliness goal). Final results suggest a reduced safeguards inspection frequency to manage the IMF.
Taniguchi, Yoshinori; Udagawa, Yutaka; Amaya, Masaki
Annals of Nuclear Energy, 139, p.107188_1 - 107188_7, 2020/05
Udagawa, Yutaka; Mihara, Takeshi; Taniguchi, Yoshinori; Kakiuchi, Kazuo; Amaya, Masaki
Annals of Nuclear Energy, 139, p.107268_1 - 107268_9, 2020/05
Fukaya, Yuji; Goto, Minoru; Ohashi, Hirofumi
Annals of Nuclear Energy, 138, p.107182_1 - 107182_9, 2020/04
The investigation on self-shielding effect of double heterogeneity for plutonium burner High Temperature Gas-cooled Reactor (HTGR) design has been performed. Plutonium burner HTGR designed in the previous study by using the advantage of double heterogeneity to control excess reactivity. In the present study, the mechanism of the self-shielding effect is elucidated by the analysis of burn-up calculation and reactivity decomposition based on exact perturbation theory. As a result, it is revealed that the characteristics of burn-up reactivity are determined by resonance cross section peak at 1 eV of Pu due to the surface term of background cross section, this is, the characteristics of neutron leakage from fuel lump and collision to a moderator. Moreover, significant spectrum shift is caused during the burn-up period, and it enhances reactivity worth of Pu and Pu in EOL.
Okada, Yuji; Amaya, Masaki
Annals of Nuclear Energy, 136, p.107028_1 - 107028_9, 2020/02
Komeda, Masao; Toh, Yosuke
Annals of Nuclear Energy, 135, p.106993_1 - 106993_6, 2020/01
This paper presents a conceptual study of a novel active method using a neutron source. The main feature of this new method is the fast rotation of a neutron source in order to derive the fission neutron counts and applying the counts to detect the nuclear material. Irradiating neutrons to a container that involves nuclear material, the measurement data include both neutrons from the neutron source and fission neutrons. However, if the neutron source is rotated quite fast, the components of the irradiation neutrons and fission neutrons are separated. Since this novel method does not require an expensive D-T tube, this new system is expected to be affordable and easy to assemble.
Takamatsu, Kuniyoshi; Matsumoto, Tatsuya*; Liu, W.*; Morita, Koji*
Annals of Nuclear Energy, 133, p.830 - 836, 2019/11
A RCCS having passive safety features through radiation and natural convection was proposed. The RCCS design consists of two continuous closed regions: an ex-reactor pressure vessel region and a cooling region with a heat-transfer surface to ambient air. The RCCS uses a novel shape to remove efficiently the heat released from the RPV through as much radiation as possible. Employing air as the working fluid and ambient air as the ultimate heat sink, the RCCS design can strongly reduce the possibility of losing the working fluid and the heat sink for decay-heat-removal. Moreover, the authors started experiment research with using a scaled-down heat-removal test facility. Therefore, this study propose a comparative methodology between an actual RCCS and a scaled-down heat-removal test facility.
Sugino, Kazuteru; Numata, Kazuyuki*; Ishikawa, Makoto; Takeda, Toshikazu*
Annals of Nuclear Energy, 130, p.118 - 123, 2019/08
In MA sample irradiation test data calculations, the neutron fluence during irradiation period is generally scaled by using dosimetry data in order to improve calculation accuracy. In such a case, appropriate correction is required to burnup sensitivity coefficients obtained by the conventional generalized perturbation theory because some cancellations occur in the burnup sensitivity coefficients. Therefore, a new formula for the burnup sensitivity coefficient has been derived with the consideration of the neutron fluence scaling effect (NFS). In addition, the cross-section-induced uncertainty is evaluated by using the obtained burnup sensitivity coefficients and the covariance data based on the JENDL-4.0.
Sugawara, Takanori; Takei, Hayanori; Tsujimoto, Kazufumi
Annals of Nuclear Energy, 125, p.242 - 248, 2019/03
To realize the feasible accelerator-driven system (ADS) concept, the investigations for the reliable accelerator and conceptual plant design considering safety issues were performed. As the reliable accelerator concept, the double-accelerator concept was proposed to reduce the beam-trip frequency. The estimated beam-trip frequency with the double-accelerator concept using the J-PARC LINAC operation data showed that the beam-trip frequency was significantly improved with the comparison of the single accelerator result. The basic investigation of the primary reactor auxiliary coolant system (PRACS) was performed for the safety design of the LBE cooled ADS. The concept which the PRACS heat exchanger was integrated to the steam generator was proposed and the transient analysis in the loss of heat sink accident was carried out. The result presented that the decay heat removal was appropriate when the operation of the PRACS succeeded.
Atz, M.*; Salazar, A.*; Hirano, Fumio; Fratoni, M.*; Ahn, J.*
Annals of Nuclear Energy, 124, p.28 - 38, 2019/02
The likelihood for criticality in the far field of a repository was evaluated for direct disposal of commercial light water reactor used nuclear fuel. Two models were used in combination for this evaluation: (1) a neutronics model to estimate the minimum critical masses of spherical, water-saturated depositions of fuel material; (2) a transport model to simulate the dissolution of fuel material from multiple canisters and the subsequent transport of the solutes through host rock to a single accumulation location. The results suggest that accumulation of a critical mass is possible under conservative conditions but that these conditions are unlikely to occur, especially in the vicinity of a carefully-arranged repository.