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"Fast reactors."
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Sodium fast reactors with closed fuel cycle
The authors cover research and development on the sodium cooled fast reactors. They deal with a wide range of topics in the domain of science and technology under topics like design aspects, safety, construction, fuel cycles, and more.
Book
Void Reactivity Coefficient for Hybrid Reactor Cooled Using Liquid Metal
A negative value of the void reactivity coefficient (αV) is one of the most important passive safety properties for the operation of nuclear reactor. Herein, are presented calculated values of the void reactivity coefficient for different geometries of reactors cooled by liquid lead (LFR) and sodium (SFR) with U-238-Pu-239 and Th-232-U-233 fuels. The calculations were carried out for the reactors filled with either one or two types of fuel assemblies. The most interesting results are obtained for reactor filled with two different types of fuel assemblies (hybrid reactor). Hybrid reactors consist of central and peripheral types of fuel assemblies using low enrichment fuel and high enrichment fuel, respectively. Both hybrid reactors based on the uranium cycle (U-cycle) and the thorium cycle (Th-cycle) can maintain a negative void reactivity coefficient value for wide range of reactor parameters. The calculation results of the hybrid reactor matched those from FBR-IME reactor.
Journal Article
Effects of spent nuclear fuel on neutron and physical characteristics of a fast reactor core
Nuclear power waste can be reduced by replacing depleted uranium with spent nuclear fuel (SNF) from high-power channel (RBMK) reactors in the fuel compositions of fast neutron reactors. To determine the possibility of using RBMK SNF, processed through a simplified reprocessing technique, as a feedstock for fast reactors. We have simulated a fuel campaign for a modernized 1200â¯MW BN-1200M sodium-cooled fast neutron reactor fueled with UâPu nitride fuel. Physical calculations were carried out for scenarios with both homogeneous and heterogeneous layouts of the fast reactor core; the layouts are planned to be used at the stage 1 and stages 2, 3 of operation, respectively. The results of physical calculations are presented for a scenario with a simplified technology of RBMK SNF reprocessing for the first active loading and makeup fuel: the maximum reactivity margin for the burnup of the core fuel has increased by 0.3 and 0.4% for a homogeneous and heterogeneous layout, respectively. The equivalent dose rate (EDR) of photon radiation from the fresh fuel assembly similarly increases by 250 and 40 times for a 30-day and 3âyear storage, respectively. The EDR of fission products for a fresh assembly decreases to 70 and 50 Sv/h at a purification factor of 1·10.sup.4 and its increased value, respectively. Reprocessed RBMK SNF used instead of depleted uranium fully realizes the energy potential of natural uranium, as well as reduces nuclear power waste and load on SNF storage facilities.
Journal Article
Neutronic and Thermal Hydraulic Analysis of Gas Cooled Fast Reactor in Indonesia – An Overview
Gas-cooled Fast Reactor (GFR) is one of the fourth-generation reactors that is still being developed. The GFR system combines the advantages of a fast spectrum system for long-term sustainability of uranium resources and waste reduction (through fuel reprocessing and long-lived actinide fission) with a high-temperature system (industrial use and high thermal cycle efficiency). Several types of reactors, research methods have been carried out such as the type of output power produced, fuel variations, fuel composition variations, ring geometry variations, and core configuration variations have been carried out. Research has been carried out related to the thermal hydraulic analysis of GFR using various numerical methods such as Computation Fluid Dynamics, Runge Kutta method up to Order 5, a genetic algorithm, and MATLAB code to find some parameters i.e. d. This article discusses the results of several neutronic and thermal hydraulic analyzes from research that has been carried out and discusses things that can be developed from previous GFR research.
Journal Article
Neutronic Study on Ac-225 Production for Cancer Therapy by (n,2n) Reaction of Ra-226 or Th-230 Using Fast Reactor Joyo
Ac-225 has lately drawn considerable attention as a radioisotope for targeted alpha therapy treatment for certain types of prostate, blood-derived, and disseminated cancers, but its supply is limited. Therefore, we investigated the production method of Ac-225 by nuclear transmutation in a fast neutron reactor. The authors investigated irradiation of Ra-226 or Th-230 as a target nuclide in the experimental fast reactor Joyo, owned and operated by Japan Atomic Energy Agency, which has abundant fast neutrons and a large loading region with high heat removal capacity. Ra-226 is in increasing demand as a target nuclide to produce Ac-225. Therefore, as another option, we selected Th-230, which is 50 times more abundant than Ra-226 in natural uranium, as an alternative nuclide. Irradiation of Ra-226 and Th-230 with high energy neutrons above the threshold causes an (n,2n) reaction, producing Ra-225 and Th-229, respectively, which are the parent nuclides of Ac-225. The analyses showed that 47 GBq of Ac-225 can be generated annually by irradiating 1 g of Ra-226, and 6.5 GBq of Ac-225 can be semi-permanently generated every year by one-time irradiation of 50 g of Th-230 for 10 years (5 EFPY). It can be concluded that 100 MWt Joyo has potential to produce more than 70% of the current global supply of Ac-225 and/or to generate the parent nuclide Th-229, which keeps producing Ac-225 for thousands of years.
Journal Article
Use of Eutectic Na–Tl Coolant in a Modular Fast Reactor
One of the main problems of fast reactors with sodium coolant is its high chemical activity in interaction with water and air. This requires complex safety systems, fire extinguishing, and diagnostics and complicates the design of steam generators and pipelines. The use of Na–Tl eutectic as a coolant for the first and intermediate circuits of a modular fast neutron reactor can alleviate this problem. Sodium–thallium eutectic (92.9%Na–7.1%Tl) has a much lower chemical activity than pure sodium and prevents or extinguishes its ignition by forming an inert surface layer. Also, this coolant has a higher boiling point and a lower melting point in comparison with sodium. The introduction of thallium into the coolant can change the neutron spectrum in the core and the neutronic characteristics of the reactor. In this paper, the effect of replacement of Na coolant with the Na–Tl eutectic on the neutronic characteristics of a modular fast reactor with metal fuel is studied. In addition, the influence of possible changes in the isotopic composition of thallium is studied. The activation of thallium isotopes in the coolant is simulated and their contribution to the gamma-radiation source is investigated.
Journal Article
Burnup Performance of modified CANDLE shuffling in axial direction on gas cooled fast reactors with UN-Th fuels
The modified CANDLE (Constant Axial shape of Neutron flux, nuclide densities, and power shape During Life of Energy production) strategy shuffling in an axial direction has been successfully applied to gas cooled fast reactor. This paper investigated the utilization of natural uranium, enriched nitride and Thorium ( 238 U 15 N and 232 Th) as fuel on 3000MWt reactor power and refuelling every 10 years of burnup. The reactor core is partitioned into ten regions with the same volume in the axial direction. Initially, a fuel input ( 238 U 15 N and 232 Th) is placed in region 1. After ten years of burnup, a fuel in region 1 was moved to region 2, then to region 3, and so on until the fuel in region 9 was moved to region 10. The fuel in 10 th region was removed in the core. The neutronic computations were performed in two different ways by using SRAC 2006 code and JENDL4.0 as a nuclear data library. Firstly, PIJ was used for fuel cell calculations and secondly CITATION for a reactor core calculation. The results show that the effective multiplication factor is greater than one, at the beginning of life, its value is approximately 1.0028 and at the end of life is approximately 1.0279. This indicates that the reactor is capable to operate through a burn-up period by employing 238 U 15 N and 232 Th (10%) as fuel cycle input. At the beginning of life, the integral conversion ratio is approximately 9.9455 and at the end of life is approximately 1.2906. The burnup level at the end life is about 30.29% HM.
Journal Article
INVESTIGATION OF A SELF-ACTUATED, GRAVITY-DRIVEN SHUTDOWN SYSTEM IN A SMALL LEAD-COOLED REACTOR
Passive safety systems in a nuclear reactor allow to simplify the overall plant design, beside improving economics and reliability, which are considered to be among the salient goals of advanced Generation IV reactors. This work focuses on investigating the application of a self-actuated, gravity-driven shutdown system in a small lead-cooled fast reactor and its dynamic response to an initiating event. The reactor thermal-hydraulics and neutronics assessment were performed in advance. According to a first-order approximation approach, the passive insertion of shutdown assembly was assumed to be influenced primarily by three forces: gravitational, buoyancy and fluid drag. A system of kinematic equations were formulated a priori and a MATLAB program was developed to determine the dynamics of the assembly. Identifying the delicate nature of the balance of forces, sensitivity analysis for coolant channel velocities and assembly foot densities yielded an optimal system model that resulted in successful passive shutdown. Transient safety studies, using the multi-point dynamics code BELLA, showed that the gravity-driven system acts remarkably well, even when accounting for a brief delay in self-actuation. Ultimately the reactor is brought to a sub-critical state while respecting technological constraints.
Journal Article
Neutronics Analysis of Heterogeneous Core of Small Modular Reactor Type GFR Thorium Nitride Fueled using OpenMC
The use of mixed fuel in the form of thorium-uranium nitride (ThN-UN) offers the potential to improve thermal efficiency and optimize criticality, especially in fast reactors like Gas-cooled Fast Reactors (GFRs). To maintain its criticality, Th-232 absorbs neutrons and transforms into a new fissile material U-233. The inclusion of fertile material Th-232 necessitates a more precise reactor geometry design to ensure the reactor remains critical state until the end of the burn-up period. This study aims to compare variations in the percentage of U-233 enrichment in heterogeneous core geometries with five fuel variations (F1:F2:F3:F4:F5) using the OpenMC program at a power of 100 MWth. Benchmarking is performed by measuring the effective multiplication factor (k-eff) over 5 years of burn-up using reference data from previous studies. If the error value is less than 2%, further calculations will be conducted for both homogeneous and heterogeneous core configurations. The homogeneous calculations indicate that the U-233 enrichment percentage of 8.5% yields the most optimal results. This homogeneous data then is used for calculations in the heterogeneous core with 5 fuel variations. The heterogeneous core configuration is designed using five types different fuel percentages cases, each with variations in ring geometry. The comparison results show that the case 5 heterogeneous core geometry design, with a percentage distribution of 7%:7.5%:8.5%:9%:10.5%, achieves good optimization in terms of k-eff value, excess reactivity, neutron flux, and extended burn-up over a 10-years period.
Journal Article
Advancing Gas-cooled Fast Reactor Technology: Outcomes of the Euratom SafeG Project on ALLEGRO Research and Development
The paper describes R&D activities suported by Euratom funded projects (SafeG, TREASURE) focused on advancing the development of the Gas-cooled Fast Reactor (GFR) demonstrator ALLEGRO, a key technology in the Generation IV nuclear reactor systems. With the backing of the Generation IV International Forum (GIF) and European Sustainable Nuclear Industrial Initiative (ESNII) of the Sustainable Nuclear Energy Technology Platform (SNETP), GFR technology promises high efficiency in both electricity and industrial heat production, owing to its ability to achieve high core outlet temperatures and close the fuel cycle. The presented activities revolve around addressing crucial technical challenges linked to the ALLEGRO demonstrator. The described projects have brought together leading European and international experts in GFR and High-Temperature Reactor (HTR) technologies. The key outcomes of the SafeG Project include advancements in core safety (WP1), innovative materials (WP2), passive safety systems (WP3), standardization and codes (WP4), and education and training activities (WP5). The future activities are outlined for the follow-up Project TREASURE.
Journal Article