Breeding Ratio Research Articles

Development of high thermal conductivity UO2–Th heterogeneous fuel

Abstract A novel design of heterogeneous UO2–Th nuclear fuel for light water reactors (LWRs) is proposed. The main advantages of the design are higher thermal conductivity, reduced maximal temperature in the fuel, lower heat stored in the fuel system, or higher thermal breeding ratio. Thermal, neutronics and simple economics calculations were performed to determine the benefits of the new fuel concept. Moreover, an experimental setup utilizing flashing method for measuring of thermal characteristics was designed. It will serve to confirm theoretical assumptions and to precisely determine thermal conductivity and heat capacity of the proposed fuel design and to assess thermal resistance between UO 2 pellets and Th plates. The many substantial benefits of the proposed fuel design are reached with very competitive cost, which is also estimated.

Integration of the Neutral Beam Injector System Into the DCLL Breeding Blanket for the EU DEMO

The integration of plant systems involving penetrations into the in-vessel components, such as heating and current drive, fuel cycle, and diagnostics, is a complex task constrained by top-level requirements of remote maintainability and high reliability. Within the EUROfusion Power Plant Physics and Technology Program, some activities are ongoing to assess the integration of different systems into the breeding blanket (BB), specifically neutral beam injector (NBI), electron cyclotron resonance heating launchers, diagnostics sightlines, fueling lines, and specific protections for the first wall (like startup limiters). This paper describes the integration of the NBI system into the Dual-Coolant Lithium-Lead (DCLL) BB for the EU DEMO. After identifying the major issues impacting the mechanical, thermal-hydraulic and neutronic behavior of the blanket, the integration efforts have been focused on minimizing the invasiveness of the NBI system and exploring different NBI options for the best compromise between plasma heating and BB performance. This paper describes the adaptation of the DCLL BB design to allocate the neutral beam duct. A particular attention is devoted to the redistribution of breeding and shielding functions, the new path of fluid circuits, and the additional cooling needs. The consequences of design modifications on key neutronic aspects like tritium breeding ratio and shielding capability are addressed. Besides, the thermal loads transferred to the BB walls from the neutral beam and the plasma are discussed, and a preliminary thermal assessment of the proposed solution is presented.

Neutronic investigation of waste transmutation option without partitioning and transmutation in a fusion-fission hybrid system

Abstract A feasibility of reusing option of spent nuclear fuel in a fusion-fission hybrid system without partitioning was checked as an alternative option of pyro-processing with critical reactor system. Neutronic study was performed with MCNP 6.1 for this option, direct reuse of spent PWR fuel (DRUP). Various options with DRUP fuel were compared with the reference design concept; transmutation purpose blanket with (U-TRU)Zr fuel loading connected with pyro-processing. Performance parameters to be compared are transmutation performance of transuranic (TRU) nuclides, required fusion power and tritium breeding ratio (TBR). When blanket part is loaded only with DRUP, initial keff level becomes too low to maintain a practical subcritical system, increasing the required fusion power. In this case, production rate of TRU nuclides exceeds the incineration rate. Design optimization is done for combining DRUP fuel with (U-TRU)Zr fuel. Reactivity swing is reduced to about 2447 pcm through fissile breeding compared to (U-TRU)Zr fuel option. Therefore, a required fusion power is reduced and tritium breeding performance is improved. However, transmutation performance with TRU nuclides especially 241Am is degraded because of softening effect of spectrum. It is known that partitioning and transmutation should be accompanied with fusion-fission hybrid system for the effective transmutation of TRU.

Neutronic Analysis of a Thorium-Fueled Reduced Moderation Boiling Water Reactor

<p class='IJASEITAbtract'>A study about Resource Renewable Boiling Water Reactor (RBWR) core, a reduced moderation boiling water reactor that features the breeding ratio larger than 1 was conducted. This study focuses on the neutronic performances of the core and aims to investigate the core sustainability when using thorium as the main fertile fuel. A fuel-self-sustaining core with high burnup set as the design target. ²³³U+Th were used as the initial fuel, and the impact of initial fissile (²³³U) content in the core fissile zone on the core neutronic performances was evaluated. Parameters related to the neutronic performances such as the core burnup, fissile breeding, and fissile inventory ratio (FIR) are considered in this study. From these results, it was confirmed that it is feasible to create a self-sustaining fuel cycle system using thorium fueled RBWR. However, there was a trade-off between the core burnup and fissile breeding that can be a significant challenge in the development of this system. Evaluating the other design variables may be considered to address this challenge. The further study to analyze the safety performances of the core is required to arrive at a safe and reliable reactor system.

Investigation on the Effect of Tritium Breeding Ratio Using Temperature Control Mechanism for DEMO Blanket

The tritium breeding ratio (TBR) is a key parameter to assess the tritium self-sufficiency of the tritium fuel cycle system for a demonstrate power plant reactor (DEMO reactor). In this paper, a new approach on the TBR optimization of breeding blanket is developed based on the temperature control mechanism theory considering the variation of the blanket structural parameters. The local TBR issues are explored based on a pressurized water-cooled reactor (PWR) blanket module. The blanket materials volume ratio issues are studied to support the blanket structure optimization. The TBR variation is presented with the change of blanket internal structural parameters, including the pipe bore, the pipe pitch, and so on, via the blanket temperature field control approach. It is found that the pipe bore and the pipe pitch should be considered simultaneously in the blanket design to ensure a reasonable temperature field. It indicates that any of the blanket structure changes, including pipe bore, pipe pitch, and so on, will lead to the material volume ratio rise or reduction, which can cause the local TBR variation accordingly. And the results confirm that the pipe bore with 9 mm will achieve the maximum TBR for the blanket module. It is concluded that TBR is affected by the blanket temperature field dynamical variation through the temperature regulation in considering the tritium release and the recovery.

Status of the EU DEMO HCLL breeding blanket design development

Abstract In the framework of the European “HORIZON 2020” innovation and research programme, the EUROfusion Consortium develops a design of a fusion power demonstrator (DEMO). One of the key components in the fusion reactor is the Breeding Blanket (BB) surrounding the plasma, ensuring tritium self-sufficiency, heat removal for conversion into electricity, and neutron shielding. CEA-Saclay, with the support of Wigner-RCP and Centrum výzkumu Řež, is in charge of the development of one of the four BB concepts investigated in Europe for DEMO: the Helium Cooled Lithium Lead (HCLL) BB. The rationales of the HCLL are the use of Eurofer as structural material, eutectic liquid lithium-lead (PbLi) as tritium breeder and neutron multiplier, and helium gas as coolant. This paper shows the basic description of the DEMO HCLL BB concept and its design evolution during the past years, from a design based on the ITER Test Blanket Module (TBM) concept to a more advanced design called “Advanced-Plus” concept. This new HCLL BB concept that has been designed in order to improve Tritium Breeding Ratio (TBR) and shielding performances is presented. This new reference HCLL BB design has been analyzed and show very promising nuclear performances. Nevertheless, the “Optimized Conservative” concept, based on ITER TBM, is still considered as a robust back-up solution since structural improvements are still necessary on the “Advanced-Plus” concept. Moreover, a new Back Supporting Structure (BSS) is presented in this paper, designed to support the BB modules, with the aim to reduce pressure drops and thermal stresses.

Preliminary analysis of tritium fuel cycle in Z-pinch-driven fusion–fission hybrid reactor

Attaining tritium self-sufficiency is indispensable in a Z-pinch-driven fusion–fission hybrid reactor (Z-FFR). In this paper, a conceptual design is presented in which the Z-FFR tritium cycle system was divided into eight subsystems. A theoretical analysis of tritium inventory based on the mean residence time was performed to quantitatively obtain the tritium distribution in each subsystem. Tritium self-sufficiency judgment criteria were established using a tritium mass flow analysis method. The dependency relationships between the burning rate, tritium breeding ratio, extraction efficiency, and tritium self-sufficiency were also specified for the steady state.

Monte Carlo calculations of the incineration of plutonium and minor actinides of laser fusion inertial confinement fusion fission energy (LIFE) engine

In this paper, neutronic analysis in a laser fusion inertial confinement fusion fission energy (LIFE) engine fuelled plutonium and minor actinides using a MCNP codes was investigated. LIFE engine fuel zone contained 10 vol% TRISO particles and 90 vol% natural lithium coolant mixture. TRISO fuel compositions have Mod①: reactor grade plutonium (RG-Pu), Mod②: weapon grade plutonium (WG-Pu) and Mod③: minor actinides (MAs). Tritium breeding ratios (TBR) were computed as 1.52, 1.62 and 1.46 for Mod①, Mod② and Mod③, respectively. The operation period was computed as ∼21 years when the reference TBR > 1.05 for a self-sustained reactor for all investigated cases. Blanket energy multiplication values (M) were calculated as 4.18, 4.95 and 3.75 for Mod①, Mod② and Mod③, respectively. The burnup (BU) values were obtained as ∼1230, ∼1550 and ∼1060 GWd tM–1, respectively. As a result, the higher BU were provided with using TRISO particles for all cases in LIFE engine.

Optimization of Th-U fuel breeding based on a single-fluid double-zone thorium molten salt reactor

Abstract Molten salt reactor (MSR) shows great promise of high thermal-electric conversion efficiency, inherent safety, and on-line reprocessing. Furthermore, fuel breeding can be realized in both thermal and fast spectrum reactors with Th-U fuel cycle, which can make use of the abundant resource of thorium. In order to combine the advantages of high breeding ratio (BR) in fast spectrum and low fissile inventory for criticality in thermal spectrum, a single-fluid double-zone thorium-based molten salt reactor (SD-TMSR) core configuration is proposed. First, assemblies in both the inner and the outer zones are optimized by adjusting the ratios of molten salt and graphite with consideration of BR, 233U inventory, double time (DT), and temperature coefficient of reactivity (TCR) at the startup time, aiming to have a minimal DT for fuel breeding and a negative TCR for security. The results show that DT has the optimum value when the ratios of molten salt and graphite in the inner zone and the outer zone are 0.357 and 1.162, respectively. And the TCR can be improved to −2 pcm/K when the side length of the graphite hexagonal prism is 7.5 cm. Then, based on the optimized geometry, the burn-up calculations with a series of on-line reprocessing rates are carried out with a self-developed MSR reprocessing sequence (MSR-RS). The results show that on-line reprocessing is beneficial to the Th-U fuel breeding. When the reprocessing rate is 200 l/d, iso-breeding can be achieved. When the reprocessing rate is 5 m3/d, the performance of Th-U breeding is improved significantly, and only 16 years is needed for the 233U doubling. The results also show that TCR remains negative and inherent safety is satisfied during all the operation time.

Fusion DEMO reactor design based on nuclear analysis

Abstract The neutronics of a fusion reactor is crucial in the conceptual design of all in-vessel components and the overall reactor design. A successful DEMO reactor must demonstrate self-sufficient production of tritium to allow sufficient plant availability and electrical generation. The neutronics shows that the design constraint of the neutron wall load (NWL) is related to the tritium breeding ratio and nuclear heating, which determine the fusion power and plasma configuration and are essential for the blanket design in a fusion reactor. To align the outlet temperatures of the blanket modules, the coolant piping system must be arranged to match the distribution of the NWL. 3D neutronics analysis ensures that the blanket design allows a self-sufficient supply of tritium. The maintenance scheme for the remote handling equipment to operate Japan’s DEMO reactor should consider the shutdown dose rate and residual heat in the vacuum vessel. Base on neutronics analysis, it is important to classify radwaste disposal and to develop radwaste management scenarios. The analyses in this paper show that the radiation exposure caused by nuclides is as low as 3.4 μSv/y when the radwaste is disposed in a shallow concrete pit. This radiation exposure level is below the shallow-land burial limit in Japan (

Nuclear analysis of the HCLL “Advanced-Plus” breeding blanket

Abstract In the frame of the European R&D effort to develop a fusion demonstrator (DEMO) a new concept of Helium Cooled Lithium Lead breeding blanket, called “Advanced-Plus” design, is studied to improve the tritium breeding. The aim of this paper is to present the nuclear analysis of this new HCLL design carried out with the TRIPOLI-4® Monte Carlo code and the Joint European Fission and Fusion nuclear data version 3.2 (JEFF-3.2). Nuclear quantities such as: neutron flux, tritium breeding ratio, nuclear heating, displacement damage and helium production are reported. Sensitivity analysis is also carried out to investigate the impact, in tritium breeding, of the next DEMO baseline with reduced outboard thickness and the advantage, in tritium production point of view, of the single module segment against the current multi-module segment option.

DEMO design using the SYCOMORE system code: Influence of technological constraints on the reactor performances

The next step for fusion energy after the ITER tokamak is the demonstration power plant DEMO. In this framework, system codes are used to address high-level key design issues for the DEMO pre-conceptual phase. They aim at capturing the interactions between the subsystems of a fusion reactor. SYCOMORE is a modular system code which includes physics and technology models coupled to an optimizer in order to explore a large design parameter space. In the present paper, trade-off studies focused on technology modules are reported including the influence of some design-driving assumptions on the reactor performances and size, starting from a European DEMO1-like design (more than 500 MW net electric power and 2 h burn duration). The increase of the mechanical stress limits in TF and CS magnets can help reducing the reactor size, slightly more when high temperature superconductors are used in the TF coil. The tritium breeding ratio can be improved to more than 1.10 by a moderate increase of the size, but the tritium burn-up ratio needs one additional meter of major radius for every percent increase. Divertor coolant options are also compared, showing some differences between helium, hot and cold water scenarios at various incident divertor heat fluxes.

Cylindrical breeding blankets for fusion reactors

Abstract Blankets with a cylindrical structure have been developed for a water-cooled ceramic breeder concept considering a pipe break inside the blanket container. Most existing studies have examined a box structure for the blanket container. Structural change from a box to a cylinder can decrease in first wall thickness while maintaining the pressure resistance of the container. Two blankets with radial and toroidal/poloidal cylindrical axes have been developed. Their first walls are hemispherical and half cylindrical. The tritium breeding capabilities of the developed blankets were studied using Monte Carlo code MCNP5.14. Different combinations of breeder and neutron multiplier materials were considered. With the combination of lithium oxide and beryllium, the tritium breeding ratio of the blankets including surrounding gaps was evaluated as higher than 1.20, leading to a net tritium breeding ratio higher than 1.05.

Investigation on the “advanced-plus” Helium Cooled Lithium Lead Breeding Blanket design concept for TBR enhancement regarding thermal and mechanical behavior

Abstract In the framework of the European “HORIZON 2020” research program, the EUROfusion Consortium develops a design of a fusion demonstrator (DEMO). The Breeding Blanket (BB) directly surrounding the plasma is a major component ensuring tritium self-sufficiency, shielding against neutrons from D-T plasmas and heat extraction for electricity conversion. CEA-Saclay, with the support of Wigner-RCP and IPP-CR is responsible of developing the Helium Cooled Lithium Lead (HCLL) BB concept. In order to enhance Tritium Breeding Ratio (TBR), an alternative HCLL BB design, called “Advanced-Plus” concept, has been investigated, with the aim to reduce the amount of steel, while ensuring good thermal and mechanical behavior. The thermal behaviors of the horizontal Stiffening Plates (hSPs) with various channels designs in terms of dimensions and layout have been parsed on simplified FE models for verifying and optimizing the HCLL Breeding Zone (BZ) before analyzing a HCLL BB ¼ model. Moreover, sensitive analyses have been carried out for testing various hSPs pitches, alternative Helium hydraulic schemes inside the hSPs as well as the detachable FW option in applying a zero Heat Flux (HF) on the FW. Furthermore, mechanical calculations have also been performed to assess the behavior of the module in faulted condition. The results obtained with the Cast3M-qualified-FEM code are presented and discussed.

Progress of Concept Design for CFETR Diagnostic System

Chinese fusion engineering test reactor (CFETR) is under conceptual design to bridge gaps between ITER and DEMO. CFETR is envisioned to produce a 200 MW (200 MW in phase I and will be increased to 1-GW level) fusion power with tritium breeding ratio >1 and a duty cycle time of approximately 0.3–0.5. This paper is to present the status in the conceptual design of CFETR diagnostic system. Based on the experience obtained in the development of ITER diagnostics and combined with CFETR special requirements, progress of CFETR diagnostic system has been described. They mainly includes: 1) the conditions and constraints in CFETR environment and 2) one set of proposed measurements and candidate diagnostic techniques for CFETR phase I. The possible needed R&D activities for future work are under review. Two R&D examples under progress will be introduced.

Neutronics Analysis of Helium Cooled Ceramic Breeder Blanket with S-shaped Lithium Zone and Cooling Plate for CFETR

China fusion engineering test reactor (CFETR) is a tokamak fusion experimental device under design to bridge the research and development gaps between international thermonuclear experimental reactor and demonstration. Helium cooled ceramic breeder (HCCB) blanket is the candidate blanket concepts for CFETR. A blanket with S-shaped lithium zone and cooling pipes reduces the space of helium manifold and is beneficial to tritium breeding. The neutronics analyses of the HCCB blanket have been performed with the Monte Carlo code monte carlo N particle transport code and nuclear cross section data from the fusion evaluated nuclear data library-3.1b data library. The tritium breeding ratio without port is 1.22 and meets the requirements of tritium self-sufficiency, as well as, the nuclear heating power of 214 MW is achieved and the global energy multiplication factor is 1.33. The shielding performance was shown to meet the design limit. The aim of the analysis is to provide the reference for the design and optimization of CFETR blanket system.

Current Progress of Tritium Fuel Cycle Technology for CFETR

Tritium fuel cycle technology is crucial for the successful development of China fusion engineering test reactor (CFETR) which aims to bridge the gap between a fusion experimental reactor (ITER) and a demonstration reactor. One of the major missions of CFETR is to realize tritium fuel self-sufficiency with a tritium breeding ratio over 1.0. In order to attain tritium self-sufficiency as well as ensure safe handling of tritium, it is important to establish a closed tritium fuel cycle. Therefore, a tritium plant which consists of various tritium processing systems and tritium safety systems is indispensable for the successful operation of CFETR. Since the mid-2000s, when China joined the ITER project, a series of R&D activities on tritium fuel cycle technology have been carried out. The current progress in terms of conceptual design of tritium plant, development of sub-system technologies and tritium related materials is presented in this paper.

Li vaporization property of Li8ZrO6 and Li5AlO4 as tritium breeders

Abstract New tritium breeding materials that maintain large tritium breeding ratio for thermonuclear fusion reactors should be developed. In this study, we investigate the lithium-vaporization behavior from the sintered bodies of Li8ZrO6 and Li5AlO4 having high lithium atom densities. From the viewpoint of lithium density, we investigate the possibility of using the sintered compacts of Li8ZrO6 and Li5AlO4 as a tritium breeder. Lithium vaporization rates from the sintered bodies of Li8ZrO6 and Li5AlO4 are high. However, for both materials, the lithium vaporization rate gradually decreases with the annealing time by “self-stabilization” in which the surfaces of each sintered body are covered by the stable layer of the second phase. Considering the temperature dependency of equivalent vapor pressure and the volume fraction of tritium breeding material loaded at high-temperature region, it is suggested that the sintered compacts of Li8ZrO6 and Li5AlO4 have high potentials for the tritium breeder.

Divertor options impact on DEMO nuclear performances

Abstract The present paper addresses the impact of the divertor option on the nuclear performances of the Demonstration fusion reactor (DEMO). As the effect of the number and size of the divertor has been already evaluated, in this work the focus has been posed on the composition in terms of amount of cooling inside the divertor cassette. Transport responses, as the Tritium breeding ratio (TBR), neutron and gamma fluxes and spectra inside the plasma chamber, as well as activation responses such as shutdown dose rate, decay gamma fluxes and heating have been evaluated for two different blanket concepts of the future European DEMO reactor: DCLL and WCLL. Three different divertor compositions have been tested demonstrating the importance of this component not only locally but in the global radiation field. The transport analysis has been performed with the Monte Carlo code MCNP5 and the JEFF3.1.1 and JEFF3.2 data libraries. The activation responses calculated using Advanced D1S method have been recently assessed and summarized in the present paper.

Tritium breeding performance of a DEMO based on the Double Null divertor configuration

Abstract Different design modifications of a Double Null (DN) DEMO were analysed by means of 3D Monte Carlo calculations using the MCNP code to assess the effect of a DN divertor configuration and various in-vessel components (IVC) on the tritium breeding performance. A simplified DN DEMO model was set up to this end. A Tritium Breeding Ratio (TBR) of 1.14 was obtained with this model for the HCPB breeder blanket concept without taking into account any auxiliary equipment such as limiters or port plugs. The inclusion of such design modifications results in a significant reduction of the TBR for different breeder blanket concepts. To compensate the loss of breeder space in the reactor either the integration of breeder materials in the divertor body or the enlargement of the available breeder zone could be considered. Both of these options can provide an increase of the TBR and ensure a sufficient margin for a DN DEMO with a HCPB breeder blanket. Meeting the TBR requirement with a Pb-Li based breeder blanket concept would be more challenging and requires significant changes to the overall IVC configuration.