Breeding Ratio Research Articles

Concept of a fast breeder reactor to transmute MAs and LLFPs

The long-term issues of nuclear power systems are the effective use of uranium resources and the reduction of radioactive waste. Important radioactive wastes are minor actinides (MAs: 237Np, 241Am, 243Am, etc.) and long-lived fission products (LLFPs: 129I, 99Tc, 79Se, etc.). The purpose of this study was to show a concept that can simultaneously achieve the breeding of fissile materials and the transmutation of MAs and LLFPs in one fast reactor. Transmutation was carried out by loading innovative Duplex-type MA fuel in the core region and LLFP-containing moderator in the first layer of the radial blanket. Breeding was achieved in the core and axial blanket. As a result, it was clarified that in this fast breeder reactor, a breeding ratio of approximately 1.1 was obtained, and MAs and LLFPs achieved a support ratio of 1 or more. The transmutation rate was 10.3%/y for 237Np, 14.1%/y for 241Am, 9.9%/y for 243Am, 1.6%/y for 129I, 0.75%/y for 99Tc, and 4%/y for 79Se. By simultaneously breeding fissile materials and transmuting MAs and LLFPs in one fast reactor, it will be possible to solve the long-term issues of the nuclear power reactor system, such as securing nuclear fuel resources and reducing radioactive waste.

Integrated design of breeding blanket and ancillary systems related to the use of helium or water as a coolant and impact on the overall plant design

Currently, for the EU DEMO, two Breeding Blankets (BBs) have been selected as potential candidates for the integration in the reactor. They are the Water Cooled Lithium Lead and the Helium Cooled Pebble Bed BB concepts. The two BB variants together with the associated ancillary systems drive the design of the overall plant. Therefore, a holistic investigation of integration issues derived by the BB and the installation of its ancillary systems has been performed. The issues related to the water activation due to the 16N and 17N isotopes and the impact on the primary heat transfer systems have been investigated providing guidelines and dedicated solution for the integration of safety devices as isolation valves. The tritium retention and the permeation rates through the blanket and its ancillary systems have been also assessed taking into account different operating points both for the BB and ancillaries and comparing, when possible, the releases with the operating and safety limits. Moreover, the issues related to the tritium start-up inventory as well as the uncertainties on the Tritium Breeding Ratio (TBR) due to the integration of the auxiliary systems within the Vacuum Vessel have been also studied. Finally, the impact of the BB concepts on the safety systems like the Vacuum Vessel Pressure Suppression System is described with a particular focus on the different measures that should be implemented according to the considered concept. All these aspects are then taken into account to drive future developments during the Concept Design Phase.

Modeling and comparative analysis of changeover of homogeneous and heterogeneous core of BN-1200 reactor to equilibrium mode

Equilibrium reactor core operating in closed fuel cycle with breeding ratio close to 1 is one of the key features of reactor designs developed within the framework of «Proryv» project direction (PD). Equilibrium reactor core is characterized by the stable reactivity and power density pattern, as well as isotopic composition of recycled plutonium. When putting into operation fast reactor using plutonium extracted from thermal reactor spent fuel in bicomponent nuclear power, the reactor core would operate rather long (more than 10 years) in transient mode, which requires special studies and special measures for controlling reactivity coast down during reactor run.Virtual-digital modeling of fast reactor (FR) fuel cycle closure is an efficient tool in such study. In this paper, RTM-2 code package option is outlined and used. This option includes updated fuel cycle model allowing to simulate heterogeneous core layout, as well as to use radial and axial blankets.Analytical studies have been made on sodium cooled large size reactor (1200 MW). Reactivity stabilization in such reactor is slightly more difficult as compared to lead cooled reactors, and excess reactivity exceeds βeff value attributed to BREST type reactors. Owing to heterogeneity of the core, the required excess reactivity can be decreased in transient mode down to βeff level. In equilibrium mode of such core, its breeding ratio (BR) is overrated, and this would even cause reactivity increase (although relatively small). Fuel breeding can be used for assurance of equilibrium mode with high fuel burnup.Basically, there is no need to use blankets in the core designs under consideration. However, one should not discount the demand for fuel breeding, which is also one of the potential advantages of fast reactors. Modeling has shown the possibility of adaptable variation of BR of BN-1200 M reactor within the range of values from 1.05 to about 1.4 with the use of additional blankets (axial and radial) and enriched N15.

Comparative study of a stationary two-component nuclear energy system with light water and fast breeder reactors versus one-component one with self-sufficient non-breeding fast neutron reactors

Russia is currently in its initial phase of transitioning to a two-component nuclear power structure, consisting of nuclear power plants with thermal and fast reactors operating in a closed nuclear fuel cycle. Alternative approaches to closing the nuclear fuel cycle often have different assumptions regarding the role of fast reactors in this new structure. They can either imply that fast reactors are self-sufficient and are economically viable in such a way, that they can completely replace existing thermal reactor technology, or that they are used for breeding fuel for the whole nuclear energy system and are essentially complementary to VVER reactors in a permanent two component symbiotic structure. The results of this study show that nuclear power scenarios favoring the development of a large-scale fast reactor fleet with a breeding ratio ∼1 without any additional VVER type component in the nuclear power structure is optimal from an economic standpoint.

Novel features of the helical volumetric neutron source FFHR-b2

The compact helical fusion reactor FFHR-b2 is a multipurpose HElical Volumetric Neutron Source available as the component test facility to test the divertor and blanket systems. In the latest design, three new technologies of the high-temperature superconducting conductor, the ceramic pebble divertor, and the cartridge-type blanket are adopted. Tin-based functional liquid metals (FLMs) including lithium are considered as the working fluid for the blanket. The FLMs have multi functions of low vapor pressure, low density, low melting point, low corrosiveness, high tritium breeding ratio, and so on. The target construction cost of the FFHR-b2 has been set to 200 billion JPY. The main target of FFHR-b2 is to demonstrate the fusion gain larger than one. To achieve this in a relatively small device, the tangential neutral beam injection (NBI) heating with a moderate energy of 80 keV is adopted for the FFHR-b2. The beam-plasma fusion with 5 MW of NBI heating can produce a fusion power of 5 MW that is larger than the absorbed power of 3 MW.

Investigation on the Tritium Self-Sufficiency and Neutron Shielding Based on a Water-Cooled Blanket for CFETR at 1.5 GW

China Fusion Engineering Test Reactor (CFETR) is a new test tokamak device in China, which is to bridge the scientific and technical gaps between ITER and DEMO. One of the main missions of CFETR is to achieve tritium self-sufficiency with a tritium breeding ratio (TBR) of no less than 1.05. In addition, the neutron shielding evaluation is also a significant issue, which could provide a significant reference for the water-cooled ceramic breeder (WCCB) blanket modification and improvements. In this work, a 3-D simplified neutronics model containing the WCCB blanket is developed and the neutronics performance is calculated based on the 1.5 GW fusion power. The nuclear analyses are carried out by the Monte Carlo N-particle (MCNP) transport code, including the TBR, neutron wall loading (NWL), fast neutron flux on inboard side, and nuclear heating deposition on main in-vessel components. It is found that the net TBR reaches 1.23, which meets the condition of tritium self-sufficiency. Moreover, it is found that the peak NWL of inboard and outboard occurs in the equatorial plane. The inboard peak NWL is 1.89 MW/m <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> and the outboard peak NWL is 1.97 MW/m <sup xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sup> . In addition, the fast neutron flux and nuclear heating deposition at the inboard midplane in toroidal field coil (TFC) are obtained. According to the radiation limits of the CFETR TFC referring to ITER, the results show the nuclear responses of the TFC meet the limit criteria.

The effects of side interactions on the neutronic analysis of HCCR breeding blanket

Helium Cooled Ceramic Reflector (HCCR) is a candidate as Multi-Module Segmentation Breeding Blanket used the Korean DEMO Reactor (K-DEMO). Neutronic analyses of the breeding blanket are substantial due to the prominent role of neutrons in tritium breeding, energy deposition, and radiation damages. This work aims to consider the effects of side reactions with low cross-sections on the neutronic study to make the simulation results more accurate. The neutronic analysis of the HCCR blanket has been performed using the MCNPX.2.6.0 3D Monte Carlo code and ENDF/B-VII nuclear data library. The effects of side reactions have been considered on Tritium Breeding Ratio (TBR) and Neutron Multiplication Factor (NMF) computations. The 9Be(n,T)7Li is a side reaction to tritium breeding, and the 7Li(n,2n) and W(n,2n) reactions with low cross-section cause neutron multiplication. Furthermore, the deposited energy by alpha particles has been investigated. Finally, the reaction rate of the photonuclear response has been examined using MCNPX.

Evaluation of thorium-based nuclear fuel breeding performance of a fast neutron irradiator based on a low-aspect ratio tokamak

Fast neutron irradiation has the capability of transforming fertile material, such as 238U or 232Th, into fissile material that can be used to manufacture fuel for conventional thermal light-water fission reactors. This would help both expand and extend the lifetime of nuclear power, which is currently constrained to natural uranium resources, which could only last between 50 and 150 years depending on the growth of the nuclear reactor fleet. In this paper, a design for a fast neutron source based on a low-aspect ratio tokamak is evaluated, based on two figures of merit: the tritium breeding ratio (TBR), a measure of tritium self-sufficiency, and the reactor support ratio (RSR), related to the speed of fissile material breeding. The system has neutron multiplying regions to ensure there are enough neutrons to breed both tritium and fissile material; it also has tritium breeding regions that contain 6Li which, when bombarded with neutrons, produce tritium. The MCNP code is used to perform neutronics simulations and obtain values for energy-resolved neutron fluxes for the relevant regions of the system, and the ORIGEN code is used to perform kinetic calculations for nuclear reactions between the neutrons and the materials within the regions. Simulations show that for the case of a spherical tokamak with aspect ratio of 1.8 at 120 MW of neutron power, the tritium breeding ratio can have values ranging 0.85 and 1.25, and with that power level the system can bring 336 assemblies to 5% 233U enrichment in irradiation times between 24 and 65 months. The values depend on both the selection of materials and key geometry parameters, such as the thickness of container walls, as well as the volumes for both neutron multiplying and tritium breeding regions. FLiBe was found to give the best performance as neutron multiplier, since it contributes to the tritium self-sufficiency as well; from the tritium breeding perspective, lithium oxide was found to be the best choice.

Validation of the neutron lead transport for fusion applications

Lead is an important material, both for fusion or fission reactors. The cross sections of natural lead should be validated because lead is a main component of lithium-lead modules suggested for fusion power plants and it directly affects the crucial variable, tritium breeding ratio. The presented study discusses a validation of the lead transport libraries by dint of the activation of carefully selected activation samples. The high emission standard 252Cf neutron source was used as a neutron source for the presented validation experiment. In the irradiation setup, the samples were placed behind 5 and 10 cm of the lead material. Samples were measured using a gamma spectrometry to infer the reaction rate and compared with MCNP6 calculations using ENDF/B-VIII.0 lead cross sections. The experiment used validated IRDFF-II dosimetric reactions to validate lead cross sections, namely 197Au(n, 2n)196Au, 58Ni(n,p)58Co, 93Nb(n, 2n)92mNb, 115In(n,n')115mIn, 115In(n,γ)116mIn, 197Au(n,γ)198Au and 63Cu(n,γ)64Cu reactions. The threshold reactions agree reasonably with calculations; however, the experimental data suggests a higher thermal neutron flux behind lead bricks. The paper also suggests 252Cf isotropic source as a valuable tool for validation of some cross-sections important for fusion applications, i.e. reactions on structural materials, e.g. Cu, Pb, etc.

Study on Multiphysics Coupling and Automatic Neutronic Optimization for Solid Tritium Breeding Blanket of Fusion Reactor

A tritium breeding blanket (TBB) is an essential component in a fusion reactor, which has functions of tritium breeding, energy generation and neutron shielding. Tritium breeding ratio (TBR) is a key parameter to evaluate whether the TBB could produce enough tritium to achieve tritium self-sufficiency (TBR &gt; 1) for a fusion reactor. Current codes or software struggle to meet the requirements of high efficiency and high automation for neutronic optimization of the TBB. In this paper, the multiphysics coupling and automatic neutronic optimization method study for a solid breeder TBB is performed, and a corresponding code is developed. A typical module of China fusion engineering test reactor (CFETR) helium cooled ceramic breeder (HCCB) TBB was selected, and a 3D neutronics model of an initial scheme is developed. The automatic neutronic optimization was performed by using the developed code for verification. Results indicate that the TBR could increase from 1.219 to 1.282 (~5.17% improvement), and that the maximum temperature of each type of material in the optimized scheme is below the allowable temperature. It is of great scientific significance and engineering value to explore and study the algorithm for automatic neutronic optimization and the code development of the TBB.

Neutronics Calculations to Support the Fusion Evaluated Nuclear Data Library (FENDL)

In the design of fusion reactors, determining radiation levels due to neutrons and photons (gammas) throughout the reactor and its surroundings is important. Radiation transport codes need to have accurate cross-section libraries in order to produce accurate results. The Fusion Evaluated Nuclear Data Library (FENDL) is an international effort coordinated by the International Atomic Energy Agency, Nuclear Data Section, that assembles a collection of the best nuclear data for fusion applications. In the current FENDL-3.1d data library, neutron cross sections for 65 of the 180 isotopes present in the library come from ENDF/B-VII.1. Monte Carlo–based neutronics calculations using cross-section libraries from FENDL (versions 2.1 and 3.1d), ENDF/B (versions VII.1 and VIII.0), and candidate new evaluations for key structural elements/isotopes such as iron and chromium were performed. The calculations were performed in reactor-relevant models including a one-dimensional (1-D) cylindrical model of ITER, a three-dimensional (3-D) computer-aided design (CAD)–based model of ITER, and a 3-D CAD-based model of the U.S. Fusion Energy System Studies Fusion Nuclear Science Facility (FNSF). The results show that neutron fluxes calculated with different cross-section libraries can be as much as 12% higher and as much as 8% lower than those calculated with the reference cross-section library (FENDL-2.1). Nuclear heating calculated with different cross-section libraries can be as much as 14% higher and as much as 8% lower than those calculated with the reference cross-section library. Iron displacements per atom calculated with different cross-section libraries can be as much as 9% higher and as much as 9% lower than those calculated with the reference cross-section library. Helium production calculated with different cross-section libraries can be as much as 19% higher and as much as 2% lower than those calculated with the reference cross-section library. Tritium production in the ITER 1-D model’s nonbreeding regions calculated with different cross-section libraries can be as much as 246% higher and as much as 5% lower than those calculated with the reference cross-section library. The tritium breeding ratio in the FNSF 3-D model calculated with different cross-section libraries averaged 1% higher at the inboard and 1.4% higher at the outboard than those calculated with the reference cross-section library.

Conceptual design of the supercritical CO2 cooled lithium lead blanket for CFETR

A conceptual design of the supercritical CO2 cOoled Lithium-Lead (COOL) blanket has been proposed as one advanced blanket candidate for the Chinese Fusion Engineering Testing Reactor (CFETR). At present, the COOL blanket is designed to fulfill the requirement of operating under the fusion power of 1.5 GW of CFETR while realizing the tritium self-sufficiency. Structural designs are carried out for outboard and inboard blanket segments and analyses with regard to neutronics and thermo-hydraulics are conducted to evaluate the blanket performance. Results indicate that a comparatively high Tritium Breeding Ratio (TBR) of 1.183 and a PbLi outlet temperature of 600 – 700°C can be achieved for this blanket concept. Besides, a Power Conversion System (PCS) based on the supercritical CO2 recompressing cycle is preliminarily designed for the blanket and the gross thermal efficiency is estimated to be 39% – 46% when assuming the turbine inlet temperature ranges between 550°C and 650°C.

Neutronic analysis of the ALLEGRO fast reactor core with deterministic ERANOS code and Monte Carlo Serpent code

The Gas-cooled Fast Reactor is an advanced concept selected to be part of the fourth generation of nuclear reactors. Before its development, the construction of the experimental demonstration reactor ALLEGRO is planned. Given the current interest in this innovative system, a neutronic study of the ALLEGRO reactor core is performed using the Monte Carlo Serpent calculation code and the deterministic ERANOS code. A detailed description of the core design and ceramic fuel composition is provided. To reduce the computational cost of the simulations for future applications, different cases are analyzed using different computational options in the TGV/VARIANT module and a 7-group energy structure. To validate the ERANOS models, the results are compared with the three-dimensional heterogeneous reference model developed in Serpent. The main parameters of the core, at the beginning of life, are calculated, such as the k-eff value, flux and power distributions, Doppler constant, effect of helium density on reactivity and the β-eff value. The main discrepancies in the results correspond to the diffusion and simplified transport calculations, due to the low density of helium. By using the 7-group structure, it was possible to reduce the calculation time with a reduced penalty in the precision of the results. Furthermore, the best agreement was obtained, with respect to the Serpent model, for the transport calculation (P3) with the 7-energy group structure. The k-eff and fuel isotope mass evolution during burnup is analyzed assuming an operating time of 365 days. The relative fuel fraction at the beginning and end of cycle, the breeding ratio and the average burnup are also given.

Corrosion of pure Cr and W in molten FLiNaK with hydrogen fluoride solution

Corrosion of pure Cr and pure W in molten FLiNaK at 773 K was investigated to deeply understand the effect of the alloying elements on the corrosion mechanisms of reduced activation ferritic/martensitic (RAFM) steels. The molten salt blanket environment was simulated by hydrogen fluoride gas bubbling. Electrochemical polarization and weight change measurements were conducted to estimate corrosion rates. The oxidation reaction rate of pure Cr was extremely high (instantly reached 2.9 × 104 mm/year). The corrosion rate of W was lower than that of the RAFM steel, JLF-1 steel, and pure Fe by one order of magnitude. W was nobler than Ni, which is known as a corrosion-resistant metal to the molten fluoride salts and hydrogen fluoride. The compatibility of W coating with the first wall structure of blankets was evaluated with the local tritium breeding ratio (TBR). When a Pb neutron multiplier was used, thin W coating almost did not affect the local TBR.

Conceptual design of a blanket module integrated with divertor for CFETR

Achieving tritium self-sufficiency (tritium breeding ratio (TBR)≥1.1) is one of the most important missions of the Chinese Fusion Engineering Test Reactor (CFETR). The neutronics analysis shows that the global TBR of the CFETR water-cooled ceramic breeder (WCCB) blanket is 1.147 when the plasma is surrounded by blankets (without any windows), which can meet the tritium self-sufficiency requirement. In the engineering design phase, however, a number of blanket modules are replaced by auxiliary systems. It would lead to decline of TBR and failure of tritium self-sufficiency. Therefore, a new blanket module concept extended to the divertor region and integrated with the divertor is proposed to increase TBR. The blanket module is located at the back of the divertor target plates. This paper presents the design of the proposed blanket module. The feasibility of the blanket design is evaluated considering neutronics, thermal and mechanical performances.

Neutronics optimization study on the first wall design of CFETR for TBR enhancement

Tritium breeding capability is one of the most important objectives for fusion reactor design, and the tritium self-sufficiency has been set as a crucial target for the China Fusion Engineering Test Reactor (CFETR). The first wall (FW) is the first component of the blanket facing plasma, which bears high heat flux and high neutron current. Due to the neutrons entering into the blanket pass through the FW and then absorbed by the breeder materials, the FW design directly influences the blanket neutronics and tritium breeding performance. In this study, the helium-cooled solid breeder (HCSB) blanket of CFETR is selected as the reference design, and the impacts of the FW design parameters on tritium breeding performance have been studied comprehensively. A new conceptual mixed armor design and FW neutronics optimization guidelines have been proposed to enhance the tritium breeding ratio (TBR). Based on the mixed armor design and the optimization guidelines, the preliminarily optimized helium-cooled FW design shows about 9% TBR enhancement.

Neutronic experiment and analyses of a hybrid tritium breeding blanket mockup for CFETR

A neutronic experiment has been performed using DT neutrons on a blanket mockup of the hybrid tritium breeding blanket (HTBB) at CAEP. As a backup blanket configuration for CFETR, HTBB design which uses uranium alloy, water coolant and ceramic breeder pebble bed has been proposed to obtain high tritium breeding ratio. The experiment has provided validation not only to the preliminary design of HTBB, but also to calculation code and nuclear data used in the simulations. Various neutron responses such as fission reaction rate, plutonium production rate, activation reaction rate and tritium production rate (TPR) have been measured. Numerical analyses were carried out with MCNP-4C code using ENDF/B-VII.1, TENDL-2017, IRDFF-v1.05 and FENDL3.1 nuclear data libraries. The comparisons between the calculation and experimental results show an agreement within about 10% for most of the measured quantities. The ratios of calculated to measured results (C/E) of tritium production rate are in 0.98–1.04 range which shows good agreement.

Molten salt breeding blanket: Investigations and proposals of pre-conceptual design options for testing in DEMO

In Europe, the DEMO test facility for Breeding Blankets (BB) will be the next step for fusion energy after ITER. The BB is a key component facing the plasma and ensuring tritium self-sufficiency, shielding against neutrons and heat extraction for electricity production. Within the EU research program, the two concepts candidates as driver BB in DEMO consider high-pressure coolants (water or helium) and Lithium-Lead or Beryllium pebble beds as breeder. In this article, an advanced BB concept is propounded as a candidate to be tested in DEMO taking into account Molten Salt (MS) as breeder and coolant and, consequently, named Breeder and Coolant Molten Salt (BCMS) BB. BCMS BB have several advantages, which lead to their investigations. As the MS is not pressurized during operation, high requirements associated to Nuclear Pressure Equipment (ESPN / PED) could be relaxed. Furthermore, the MS magnetohydrodynamic (MHD) flow is low due to its weak electrical conductivity. Besides, MS BB have identified drawbacks such as the structural material compatibility and the MS chemistry control needed during irradiation. A bibliographic review of MS and compatible materials is conducted. Promising BCMS BB are analysed regarding the Tritium Breeding Ratio (TBR) to test the capability of such MS breeders to reach the targeted TBR. The main nuclear quantities are evaluated with various MS on a dedicated model via the TRIPOLI-4® Monte-Carlo code: TBR, nuclear heating, neutron flux, displacement damage and helium production are reported and discussed. Moreover, the thermal and thermo-hydraulic responses of the MS BB are evaluated regarding the pressure drops (ΔP), the Pumping Power (PP) and the maximum temperatures in the structures. Nonetheless, preliminary design options are investigated. Finally, the major purpose is to evaluate the feasibility of a BCMS BB and identify their advantages and drawbacks.

Statistical Analysis of Tritium Breeding Ratio Deviations in the DEMO Due to Nuclear Data Uncertainties

For the stable and self-sufficient functioning of the DEMO fusion reactor, one of the most important parameters that must be demonstrated is the Tritium Breeding Ratio (TBR). The reliable assessment of the TBR with safety margins is a matter of fusion reactor viability. The uncertainty of the TBR in the neutronic simulations includes many different aspects such as the uncertainty due to the simplification of the geometry models used, the uncertainty of the reactor layout and the uncertainty introduced due to neutronic calculations. The last one can be reduced by applying high fidelity Monte Carlo simulations for TBR estimations. Nevertheless, these calculations have inherent statistical errors controlled by the number of neutron histories, straightforward for a quantity such as that of TBR underlying errors due to nuclear data uncertainties. In fact, every evaluated nuclear data file involved in the MCNP calculations can be replaced with the set of the random data files representing the particular deviation of the nuclear model parameters, each of them being correct and valid for applications. To account for the uncertainty of the nuclear model parameters introduced in the evaluated data file, a total Monte Carlo (TMC) method can be used to analyze the uncertainty of TBR owing to the nuclear data used for calculations. To this end, two 3D fully heterogeneous geometry models of the helium cooled pebble bed (HCPB) and water cooled lithium lead (WCLL) European DEMOs were utilized for the calculations of the TBR. The TMC calculations were performed, making use of the TENDL-2017 nuclear data library random files with high enough statistics providing a well-resolved Gaussian distribution of the TBR value. The assessment was done for the estimation of the TBR uncertainty due to the nuclear data for entire material compositions and for separate materials: structural, breeder and neutron multipliers. The overall TBR uncertainty for the nuclear data was estimated to be 3~4% for the HCPB and WCLL DEMOs, respectively.

The helium bubble: Prospects for 3He-fuelled nuclear fusion

The helium bubble: Prospects for 3He-fuelled nuclear fusion