Neutron Wall Loading Research Articles

Importance of helical pitch parameter in LHD-type heliotron reactor designs

One of the key issues in the design of LHD-type heliotron reactors is to secure sufficient blanket spaces. In this respect, helical pitch parameter γ is quite important because it significantly affects both the coil and plasma shapes. For a quantitative understanding of the effect of helical pitch parameter on a design window, a system design code for LHD-type heliotron reactors HELIOSCOPE (HELIOtron System design COde for reactor Performance Evaluation) is developed and parametric scans are carried out with three cases of γ = 1.15, 1.20 and 1.25. It became clear that design windows of heliotron reactors depend significantly on the engineering constraints: the stored magnetic energy of the coil system, the inboard minimum blanket space and the averaged neutron wall load. In the case of a fusion power of 3 GW, γ = 1.20 is optimum for relaxing physics requirements. But γ = 1.15 is also a possible selection if a lower fusion power or a higher neutron wall load is accepted. Since design windows are quite sensitive to the engineering constraints and physics conditions, a further detailed study on design feasibility of advanced engineering components and the effect of γ on the physics conditions is expected to optimize the value of γ.

Fluid dynamic and thermal analyses of a HCPB TBM Breeder Unit mock-up

In the frame of the WP08 EFDA Goal Oriented Training Program (GOTP) EUROBREED [1], the Working Package 1 is aimed at the design and procurement of a prototype of a Helium Cooled Pebble Bed Test Blanket Module Breeder Unit (HCPB TBM BU) and its testing in a dedicated experiment in a medium scale (<50 g/s) helium loop. This paper presents the design and analyses studies of the helium coolant fluid dynamics and of the thermal field in a HCPB TBM BU. After several design iterations, the helium coolant mass flow distribution has been improved between 3% and 7%. The analysis of the thermal field of the improved design has been performed. Considering steady state conditions with a heat flux deposed on the first wall equal to 500 kW/m 2 and a neutron wall load of 0.78 MW/m 2, the maximum temperature reached in the BU is 560.9 °C, about 2% above the maximum design temperature of the structural material (550 °C). As this maximum temperature is localized in a zone where there are not foreseen high stresses and the assumption for the analyses is conservative, the design presented for the Cooling Plates is considered as the reference for the production of the full scale BU mock-up.

Blanket Concept of Water-Cooled Lithium Lead with Beryllium for the SlimCS Fusion DEMO Reactor

As an advanced option for SlimCS blanket, conceptual design study of water-cooled lithium lead (WCLL) blanket was performed. In SlimCS, the net tritium breeding ratio (TBR) supplied from WCLL blanket was not enough because the thickness of blanket in SlimCS was limited to about 0.5 m so as to allocate the conducting shell position near the plasma for high beta access and vertical stability of plasma. Therefore, the beryllium (Be) pebble bed was adopted as additional multiplier to reach a required TBR (≥ 1.05). Considering the operating temperature of blanket materials, a double pipe structure was adopted. The nuclear and thermal analysis were carried out by a nuclear-thermal-coupled code, ANIHEAT and DOHEAT so that blanket materials were appropriately arranged to satisfy the acceptable operation temperatures. The temperatures of materials were kept in appropriate range for the neutron wall load Pn = 5 MW/m2. It was found that the local TBR of WCLL with Be blanket was comparable with that of solid breeder blanket.

A Robust Design Window for the Heliotron DEMO Reactors

According to the achievements in the Large Helical Device (LHD) experiments, a design of a DEMO reactor with a LHD-type heliotron system is foreseeable. On the other hand, a DEMO is the next step reactor and high reliability and feasibility are demanded in its design. In this study, a robust design window, i.e., the design window that is not sensitive to a change in uncertain physics parameters, was surveyed through parametric scans using a system design code. It was found that a difference in main design parameters (major radius, magnetic field strength, fusion output) gives only a small change in a dependence of physics requirements on the physics conditions. Therefore, it is important to find the design window with a lower requirement on the confinement improvement to assure the design robustness. In this respect, a reduction in the minimum inboard blanket space, one of the key parameters in a LHD-type heliotron reactor design, can effectively expand the design window and contributes to the design robustness. An acceptance of higher neutron wall load and an achievement of further high confinement improvement are also expected to make both a DEMO and a commercial reactor to be more feasible and economically attractive.

Damage production and accumulation in SiC structures in inertial and magnetic fusion systems

Radiation damage parameters in SiC/SiC composite structures are determined in both magnetic (MFE) and inertial (IFE) confinement fusion systems. Variations in the geometry, neutron energy spectrum, and pulsed nature of neutron production result in significant differences in damage parameters between the two systems. With the same neutron wall loading, the displacement damage rate in the first wall in an IFE system is ∼10% lower than in an MFE system, while gas production and burnup rates are a factor of 2 lower. Self-cooled LiPb and Flibe blankets were analyzed. While using LiPb results in higher displacement damage, Flibe yields higher gas production and burnup rates. The effects of displacement damage and helium production on defect accumulation in SiC/SiC composites are also discussed.

Neutronics analysis of the conceptual design of a component test facility based on the spherical tokamak

One of the crucial aspects of fusion research is the optimisation and qualification of suitable materials and components. To enable the design and construction of DEMO in the future, ITER is taken to demonstrate the scientific and technological feasibility and IFMIF will provide rigorous testing of small material samples. Meanwhile, a dedicated, small-scale components testing facility (CTF) is proposed to complement and extend the functions of ITER and IFMIF and operate in association with DEMO so as to reduce the risk of delays during this phase of fusion power development. The design of a spherical tokamak (ST)-based CTF is being developed which offers many advantages over conventional machines, including lower tritium consumption, easier maintenance, and a compact assembly. The neutronics analysis of this system is presented here. Based on a three-dimensional neutronics model generated by the interface programme MCAM from CAD models, a series of nuclear and radiation protection analyses were carried out using the MCNP code and FENDL2.1 nuclear data library to assess the current design and guide its development if needed. The nuclear analyses addresses key neutronics issues such as the neutron wall loading (NWL) profile, nuclear heat loads, and radiation damage to the coil insulation and to structural components, particularly the stainless steel vessel wall close to the NBI ports where shielding is limited. The shielding of the divertor coil and the internal Poloidal Field (PF) coil, which is introduced in the expanded divertor design, are optimised to reduce their radiation damage. The preliminary results show that the peak radiation damage to the structure of martensitic/ferritic steel is about 29 dpa at the mid-plane assuming a life of 12 years at a duty factor 33%, which is much lower than its ∼150 dpa limit. In addition, TBMs installed in 8 mid-plane ports and 6 lower ports, and 60% 6Li enrichment in the Li 4SiO 4 breeder, the total tritium generation is ∼3.7 kg in the whole lifetime of CTF.

Neutronic calculations of a thorium-based fusion–fission hybrid reactor blanket

Fusion–fission hybrid reactor has advantages of production of nuclear fuel, transmutation of long-life nuclear waste, and having inherent safety. Considering the fast development of nuclear power industry and the abundant thorium resources in China, the concept of a thorium-based breeding blanket for fuel production is proposed. The blanket using ThN or ThO 2 dispersed in graphite or BeO is investigated under a first neutron wall loading of 0.57 MW/cm 2 as ITER. The design helium cooled pebble bed design is employed according to the experience in the fusion reactor field. Preliminary neutronic calculations are performed using the one-dimensional transport and burnup calculation code BISONC and the Monte Carlo transport code MCNP. The behavior of the neutronic potential is observed for 960 days. The cumulative fissile fuel enrichment values varied between 6.88% and 8.56% depending on the fuel types. The tritium breeding ratio is greater than 1.05 for all investigated fuel types and the hybrid reactor is self-sufficient in the tritium required for the (DT) fusion driver in those modes during the operation period. The blanket energy multiplication factor M, varies between 13.66 and 15.85 depending on the fuel types at the end of the operation period. In addition, the effect of 233Pa on the 233U production and k eff are also discussed.

Thermo-hydraulic and thermo-mechanical analysis of Korean helium cooled solid breeder TBM

Korea has been developing a helium cooled solid breeder (HCSB) TBM of which unique feature is that the amount of beryllium multiplier is reduced by replacing some of it with graphite reflector. Its components have thermo-hydraulic and thermo-mechanical design requirements for the associated thermal loads since the TBM should handle severe surface heat flux from the plasma as well as volumetric heat generation by neutron wall loading. In this paper, results of thermo-hydraulic and thermo-mechanical analyses are presented for the Korean HCSB TBM. To investigate that the TBM components are effectively cooled below the allowable temperature, thermo-hydraulic calculation was performed using the three-dimensional TBM model. Thermo-mechanical computations were then carried out using the surface temperature data obtained from the thermo-hydraulic simulation. The computational results show that the current TBM design is reasonable under the ITER normal condition.

Utilization of TRISO Fuel with LWR Spent Fuel in Fusion-Fission Hybrid Reactor System

HTRs use a high performance particulate TRISO fuel with ceramic multi-layer coatings due to the high burn up capability and very neutronic performance. TRISO fuel because of capable of high burn up and very neutronic performance is conducted in a D-T fusion driven hybrid reactor. In this study, TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The neutronic effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on the fuel performance has been investigated for Flibe, Flinabe and Li20Sn80 coolants. The reactor operation time with the different first neutron wall loads is 24 months. Neutron transport calculations are evaluated by using XSDRNPM/SCALE 5 codes with 238 group cross section library. The effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on tritium breeding (TBR), energy multiplication (M), fissile fuel breeding, average burn up values are comparatively investigated. It is shown that the high burn up can be achieved with TRISO fuel in the hybrid reactor.

Search for reality of solid breeder blanket for DEMO

For a tokamak fusion DEMO reactor with the fusion output of 2.95 GW, neutronic and thermal design of blanket is under way to find a feasible blanket concept. For the continuity with the Japanese ITER-TBM options, this study considered water-cooled blanket with solid breeding materials of Li ceramics (Li 4SiO 4, Li 2TiO 3 and Li 2ZrO 3) and Be multipliers (Be and Be 12Ti). Based on a neutronics-heat coupled analysis, the tritium breeding ratio was evaluated so as to satisfy constraints of the operating temperature of ≤900 °C for Li ceramics and Be 12Ti, and ≤600 °C for Be. Cooling water condition was assumed to be 23 MPa and 290–360 °C. The result indicates that surplus tritium production in lower neutron wall load ( P n ) blanket compensates a shortfall in higher P n blanket and thus the overall tritium production can marginally satisfy fuel self-sufficiency.

The potential presence and minimisation of plutonium within the irradiated beryllium in fusion power plants

Beryllium is expected to be one of the materials used in DEMO-type fusion power plants as blanket neutron multiplier for solid breeders and also in ITER as protective layer of plasma facing components. Among the aspects to be considered when dealing with Be a relevant one is its impurities content. One of the impurity reported by manufacturers is uranium, and that could represent a main source of concern, because its irradiation under fusion neutrons will lead to the formation of actinides, such as plutonium. The paper presents a parametric assessment utilizing for beryllium material composition a range of possible U impurity content between 0 and 100 ppm wt. First Wall neutron wall load ranging from 0.5 to 2.2 MW/m 2 and neutron fluence from 0.5 to 10.5 MW year/m 2 are considered (for ITER and DEMO-type reactor, respectively) to estimate the Pu production. In all those analyses the irradiation scenarios considered are based on a deuterium–tritium plasma. The evaluation have been performed using the EASY-2007 activation code package, with the EAF2007 activation library. An estimate of the Pu production at the end of the ITER fusion machine preliminary operation with deuterium–deuterium plasma is also presented. A preliminary analysis of the regulatory constraints on the maximum Pu production and indications of possible way for its minimisation is presented and discussed in the paper.

Physics Basis of a Fusion Development Facility Utilizing the Tokamak Approach

The objective of the Fusion Development Facility (FDF) under consideration is to carry forward advanced tokamak physics for optimization of fusion reactors and enable development of fusion’s energy applications. A concept of FDF based on the tokamak approach with conservative expressions of advanced physics and nonsuperconducting magnet technology is presented. It is envisioned to nominally provide 2 MW/m2 of neutron wall loading and operate continuously for up to 2 weeks as required for fusion nuclear component research and development. FDF will have tritium breeding capability with a goal of addressing the tritium self-sufficiency issue for fusion energy. A zero-dimensional system study using extrapolations of current physics and technology is used to optimize FDF for reasonable power consumption and moderate size. It projects a device that is between the DIII-D tokamak (major radius 1.8 m) [J. L. Luxon, Nucl. Fusion, Vol. 42, p. 614 (2002)] and the Joint European Torus (major radius 3 m) [P. H. Rebut, R. J. Bickerton, and B. E. Keen, Nucl. Fusion, Vol. 25, p. 1011 (1985)] in size, with an aspect ratio A of 3.5 and a fusion gain Q of 2 to 5. Theory-based stability and transport modeling is used to complement the system study and to address physics issues related to specific design points. It is demonstrated that the FDF magnetohydrodynamic stability limits can be readily met with conservative stabilizing conducting wall placement. Transport analysis using a drift-wave-based model with an edge boundary condition consistent with the pedestal stability limit indicates that the FDF confinement requirement can also be readily satisfied. A surprising finding is that the toroidal Alfvén eigenmodes are stabilized by strong ion Landau damping. Analysis of vertical stability control indicates that the basis configuration with an elongation κx [approximately] 2.35 can be controlled using a power supply technology similar to that used in DIII-D. Peak heat fluxes to the divertor are somewhat lower than those of ITER [R. Aymar, P. Barabaschi, and Y. Shimomura, Plasma Phys. Control. Fusion, Vol. 44, p. 519 (2002)], but FDF will operate with a higher duty factor.

Simulation of Plasma Parameters for HCSB-DEMO by a 1.5D Plasma Transport Code

Simulation of the core plasma parameters of HCSB-DEMO (helium-cooled solid breeder, HCSB), by using a 1.5D plasma transport code, was carried out. The study includes investigations of operational scenarios, temperature and density profiles of both ions and electrons, fusion and radiated power, distribution of the safety factor, sensitivity analyses for some input parameters as well as a primary estimate of the divertor heating load. The results indicate that the following fusion reactor parameters can be properly set for HCSB-DEMO, namely major radius of 7.2 m, aspect ratio of 3.4, elongation of 1.85, triangularity of 0.45, plasma current of 14.8 MA, normalized beta of 4.4, toroidal field (TF) of 6.86 T, average electron density of 1.5×1020 m−3, average electron temperature of 14.5 keV, fusion power of 2.55 GW, neutron wall loading of 2.3 MW·m−2 and fusion multiplication factor of 35.

Structural materials for fusion reactors

Fusion Reactors will require specially engineered structural materials, which will simultaneously satisfy the harsh conditions such as high thermo mechanical stresses, high heat loads and severe radiation damage without compromising on safety considerations. The fundamental differences between fusion and other nuclear reactors arise due to the 14MeV neutronics of structural materials. There exists considerable uncertainty in the nuclear data at such energies because there aren’t any strong enough sources for such neutrons except fusion reactors themselves! We thus encounter a problem of iterative nature in which we must try several experiments with the available materials in the near term. The development of such structural materials is thus going to require the experimental data of the kind that may be generated on reactors like ITER, high-performance modeling and a penetrating metallurgical insight to overcome technological challenges in terms of achieving required properties such as low activation by controlling the impurities, good thermo-mechanical properties by microstructure engineering, good chemical compatibility and high radiation resistance. These materials need to withstand a neutron wall load of the order of 2–3 MW/m2, which can lead up to 30 dpa of radiation damage and 300 appm helium production per full power year in DEMO like reactors. Such conditions lead to unprecedented events related to the failure of materials due to irradiation creep, Ductile-Brittle Transition Temperature (DBTT) shift and helium embrittlement.

LOCA analysis for Korean helium cooled solid breeder TBM

One of the major ITER goals is test blanket module (TBM) program which is for the demonstration of the breeding capability that would lead to tritium self-sufficiency in a reactor and the extraction of high-grade heat suitable for electricity generation under the ITER fusion environment. While the engineering design of Korean helium cooled solid breeder (HCSB) TBM and its ancillary systems has been performed, a safety assessment on different possible accident scenarios should be carried out for the purpose of licensing. In this paper, accident analyses for several loss of coolant accident (LOCA) cases were performed in order to assess safety aspects of the TBM design using RELAP5/MOD3.2. Since the TBM forms a loop with helium cooling system (HCS) which is one of ancillary systems required for removing heat deposited in the TBM by neutron wall loading and surface heat flux from plasma, it is necessary to model the complete loop for accident analysis. In this study, the helium passage including the TBM and HCS was nodalized for each accident scenario. The TBM and HCS components were modeled as the associated heat structures provided by RELAP5 to include heat transfer across solid boundaries. Based on computational results it was found that current design of the TBM is robust from the safety point of view.

Conceptual design study of fusion DEMO plant at SWIP

The basic definition and development strategy of the DEMO plant based on the Chinese fusion power plant (FPP) program are presented briefly. A conceptual design study of fusion HCSB-DEMO reactor with a fusion power of 2550 MW and a neutron wall loading of 2.3 MW/m 2 is performed recently. Three sets parameters of core plasma for different DEMO design objectives are proposed. A helium-cooled blanket system with ceramic breeder (Li 4SiO 4), the structure material of low-activation ferritic steel (LAF/M) and Be neutron multiplier based on Chinese ITER HCSB-TBM design foundation are considered. The design parameters, preliminary analyses and the basic structure as well as development strategy of HCSB-DEMO reactor are introduced.

Three-dimensional neutron source models for toroidal fusion energy systems

Developments in computer architecture and neutronics code capabilities have enabled high-resolution analysis of complex 3D geometries. Thus, accurately modeling 3D source distributions has become important for nuclear analyses. In this work two methods are described which generate and sample such 3D sources based directly on the plasma parameters of a fusion device and which facilitate the ability to update the neutron source following changes to the plasma physics configuration. The cylindrical mesh method is for toroidally symmetric machines and utilizes data in a standard file format which represents the poloidal magnetic flux on an R–Z grid. The conformal hexahedral mesh method takes plasma physics data generated in an idealized toroidal coordinate system and uses a Jacobian transformation and a functional expansion to generate the source. This work describes each methodology and associated test cases. The cylindrical mesh method was applied to ARIES-RS and the conformal hexahedral mesh method was applied to a uniform torus and ARIES-CS. The results of the test cases indicate that these improved source definitions can have important effects on pertinent engineering parameters, such as neutron wall loading, and should therefore be used for high-resolution nuclear analyses of all toroidal devices.

Numerical and Statistical Analysis of FR Spent Fuel Transmutation in a Thorium Fusion Breeder

In this study, a numerical analysis and an analysis of variance (ANOVA) are applied to find the best suitable neutronic parameters for the performance analysis in a thorium fusion rector. The numerical and ANOVA approach are employed to investigate the neutronic characteristics of a fusion reactor using ThO2 90% + FR spent fuel 10% fuel mixtures. Three different neutronic parameters for the ANOVA and numerical approach, namely, moderator/fuel volume fractions (Vm/Vf), plasma chamber dimensions (PCD) and neutron wall loading (NWLs) as time dependent are selected for neutronic performance characteristics including tritium breeding ratio (TBR), multiplication factor (M), total fission rate (Σf), 232Th(n,γ) reaction, burn up and/or transmutation (B/T) and fissile fuel breeding (FFBR). Moreover, effects of the NWLs, Vm/Vf fractions and PCD in the B/T of FR spent fuel mixed thorium are investigated. Numerical and statistics approach results are evaluated for TBR, M, Σf fission rate, 232Th(n,γ) reaction, B/T and FFBR.

Optimization activities on design studies of LHD-type reactor FFHR

Recent activities on optimizing the base design of the large helical device (LHD)-type helical reactor FFHR (force free helical reactor) are presented. Three candidates to secure the blanket space are proposed with the aim of reactor size optimization without deteriorating α-heating efficiency and by taking cost analyses into account. In this way the key engineering aspects are investigated; from 3D blanket designs, it is demonstrated that the peaking factor of the neutron wall loading is 1.2–1.3 and a blanket covering ratio of over 90% is possible by proposing discrete pumping with a semi-closed shield (DPSS) concept. Helical blanket shaping along the divertor field lines is the next big issue. For large superconducting magnet systems under the maximum nuclear heating of 200 W/m 3, cable-in-conduit conductor (CICC) and alternative conductor designs are proposed with a robust design of cryogenic support posts. For access to ignited plasmas, new methods are proposed, in which a long rise-up time over 300 s reduces the heating power to 30 MW and a new proportional-integration-derivative (PID) control of the fueling can handle the thermally unstable plasma at high-density operation. This paper focuses on FFHR2m1, which is a modified version of FFHR.

Comparison analysis of 1D/2D/3D neutronics modeling for a fusion reactor

Fusion reactors have the characteristics of a strong anisotropic neutron flux distribution, wide range of neutron energy and spatial non-uniformity of the power density distribution caused by the external fusion neutron source. The static state neutronics analyses of a fusion power reactor (named FDS-II) have been carried out to find out the effects of the 1D/2D/3D geometric descriptions for computational models. The neutron wall loading distribution, tritium-breeding ratio and power density distribution are calculated to evaluate the nuclear performance using the Monte Carlo transport code MCNP4C and nuclear data library FENDL-2.1. Comparison of the results suggests that the value of TBR calculated by the 1D approach overestimates the tritium-breeding ratio especially in the inner blanket. The power density distribution based on the 2D and 3D models were similar since they all consider the effects of the axial components. For obtaining global scalar quantities for general use, simplified models of the fusion reactor are usually sufficient. However, more calculations are needed to test the validity of each model in different regions of phase space.