Average Neutron Wall Loading Research Articles

Integrated Liquid Metal Flowing First Wall and Open-Surface Divertor for Fusion Nuclear Science Facility: Concept, Design, and Analysis

The Fusion Energy System Studies (FESS) Fusion Nuclear Science Facility (FNSF) project team in the United States is examining the use of liquid metals (LMs) for plasma-facing components (PFCs). Our approach has been to utilize an already established fusion design, FESS-FNSF, which is a tokamak-based machine with 518 MW fusion power, a 4.8-m major radius, a 1.2-m minor radius, and a machine average neutron wall loading of ~1 MW/m2. For this design, we propose a PFC concept that integrates a flowing LM first wall (FW) and an open-surface divertor. The flowing LM first removes the surface heat flux from the FW and then proceeds to the lower section of the vacuum chamber to form a large area LM surface for absorbing high peak surface heat flux in the divertor region. In pursuing the application of large open LM surfaces in the FNSF, two new computer codes have been developed and then applied to the analysis of free-surface magnetohydrodynamic flows and heat transfer, including fast thin flowing liquid layers over the solid FW (liquid wall), a tublike divertor, and a fast flow divertor. The analysis is aimed at optimization of the liquid wall design by matching certain proposed design criteria and also at evaluation of the maximum heat fluxes, using liquid lithium (Li) as a working fluid. It was demonstrated that the flowing Li FW (at ~2 cm and ~10 m/s) can tolerate a surface heat flux of ~1 MW/m2, while the open-surface Li divertor can remove a maximum high peak heat flux of 10 MW/m2. The paper also focuses on the underlying science. One such example is the evaluation and characterization of heat transfer mechanisms and heat transfer intensification in the tublike Li divertor.

Particles Transportation and Nuclear Heating in a Tokamak by MCNPX and GEANT4

In this work as a first goal, GEANT4 Monte Carlo tool has been evaluated by means of comparison with MCNPX as alternate Monte Carlo tool in the field of tokamaks nuclear analysis. For this respect, the ITER-like geometry was studied as a candidate tokamak. The average neutron wall loading over the inner surface of the ITER first wall and neutron and gamma spectra in blanket have been calculated and compare to standards data. The results shows, GEANT4 under the HP model presents more agreement with MCNPX results. As the second achievement, the GEANT4 evaluation leads to make a confirmation to recent development in ITER nuclear heating enhancement.

Spiral Blanket: An Innovative Idea for a Solid Wall Blanket with Potential for High Neutron Wall Loading

ABSTRACTThe Advanced Power Extraction (APEX) program1 is exploring concepts for blanket designs that can enhance the potential of fusion while using both liquid and solid walls. In that context, an innovative blanket design of dry solid wall configuration, with potential for high neutron wall loading has been proposed. The blanket utilizes nanocomposited ferritic (NCF) steel structure (designated 12YWT), which has a maximum operating temperature of 800°C. The cooling/breeding material is Flibe (Li2BeF4), a low viscosity version of this molten salt, which has a melting temperature of 465°C and is compatible with ferritic steels up to 700°C2. The dimensions for this study have been taken from ARIES-AT3. The blanket module, which extends 0.3 m in the toroidal direction at mid-plane, is equipped with spiraling discs ramping from the bottom to the top. The coolant enters on the bottom at 500°C, then travels on the spiral discs, from the rear of the module to the front, then back to the rear, all the way to the top where it exits from the module at 590°C. On its way up, the coolant velocity is amplified at the first wall (FW) by centrifugal action, providing a high heat transfer coefficient for dissipating the high surface heating. The 3 mm thick FW is scalloped with semicircular projections inclined in the flow direction. This facilitates unimpeded smooth flow at the FW, while at the same time stiffening the FW against pressure, obviating the need for welded reinforcements. The discs are made of two halves assembled together with a Be pebble bed enclosed as a neutron multiplier. Preliminary neutronics analysis for the outboard blanket has shown a local tritium breeding ratio (TBR) of 1.36, and an energy multiplication (M) of 1.26 including contribution from a secondary module, using natural Li. The pressure drop of 0.56 MPa in the module produces a modified primary bending stress of 124 MPa where the allowable stress for this material at 750°C is 142 MPa. Without optimization, this blanket is capable of dissipating an average neutron wall loading of 6.4 MW/m2, with a peak value of 9.6 MW/m2 and a peak surface heating of 1.3 MW/m2. The coolant picks up energy in the secondary blanket exiting the reactor at 600°C. Assuming the use of a supercritical steam power cycle, an efficiency of 48% can be expected.

The breeding potential of ‘flinabe’ and comparison to ‘flibe’ in ‘CLiFF’ high power density concept

Because of its low melting temperature (240 °C) and low vapor pressure, the molten salt LiF–NaF–BeF 2 (Flinabe) with ratio 1:1:1 has been suggested as the front flowing liquid layer, FFLL (2-cm thick) in the Convective Liquid Flow First Wall (CLiFF) high power density concept (7 MW/m 2 average neutron wall load, 10 MW/m 2 max.). However, because of its lower Li concentration relative to LiF–BeF 2 (Flibe, with ratio 2:1), its deployment as a breeder has been a concern. In this paper, we compare the local tritium breeding ratio (TBR) of Flinabe to Flibe in two arrangements: (1) as a FFLL and breeder in the blanket; and (2) as only the FFLL with LiPb/SiC conventional blanket following the FW. The impact of the choice of ferritic steel (FS) and SiC as the structural material on TBR was also examined. In addition, the paper discusses what constitutes a reasonable value for local TBR if all sources of uncertainties in the achievable and the required TBR are considered (e.g. modeling, nuclear data, burn up rate in the plasma, etc.)

Toroidal reactor designs as a function of aspect ratio and elongation

A `common basis' systems study of superconducting (SC) and normal conducting (NC) DT burning fusion power and materials testing reactor designs is presented. The figures of merit for power and materials testing reactors are, respectively, projected cost of electricity (COE) and direct cost. A common 0-D plasma modelling basis is used and the plasma geometry and engineering aspects of the SC and NC designs are treated in an equivalent manner that is consistent with the limitations of their respective magnet technologies. Aspect ratios A in the range 1.2⩽A⩽6 and plasma elongations κ in the range 1.5⩽κ⩽3 are explored and an MHD stability (beta limit) physics basis that accurately describes the increase of normalized beta βN and toroidal beta βT with decreasing A and/or increasing κ is incorporated. With this MHD basis taken into account and with the usual reactor geometry, physics and engineering constraints and costing bases applied, the results of the study show that for SC power reactor designs with κ = 2 the COE has a minimum for 2⩽A⩽3 and increases with a further increase in A(A>3). For NC power reactors the COE has minima at A≈2. For both SC and NC power reactors, the minima are more apparent with lower κ. While SC options appear to offer lower COE for power plants, the direct cost for NC test reactors with similar fusion power output is significantly lower. Within the NC category, test designs that combine modest A and maximum elongation show promise for achieving ITER-like testing capabilities at low direct cost. For example, an NC coil design with A = 2, κ = 3 could produce fusion power of 200 MW at 1.23 MW/m2 average neutron wall loading at a total direct cost of about 650 million US dollars. This NC design with a fissile blanket could also convert about 1270 kg of fission reactor waste per full power year. A possible cost effective development scenario for fusion power is identified for NC A = 2 toroidal devices for physics and material testing studies for use in the near future. The selection of the SC or NC coil option could then be made for the construction of the demonstration power reactor.

Neutron wall loading of Tokamak reactors

Neutron wall loading (Γn) is a key parameter for the selection of fusion power core component materials. It also impacts the economic, performance, design, safety and environmental aspect of the fusion power plant. This paper reports the determination of the range of Γn for economically competitive fusion power plants based on the analysis that couples the MHD stability physics results to a system design code. Cost of electricity (COE) was selected as the parameter to be minimized. For both normal conducting and superconducting coil options, at thermal efficiency of 46% and at the power output range of 1–2 GW(e) the average neutron wall loading is 4–7 MW/m2. For a given power output, higher thermal efficiency will allow lower Γn. At the above range of Γn, in order to have economical fusion power reactors, for the solid first wall design option, high thermal efficiency of 46% to 57.5% requires the use of alloys like V and W-alloy, respectively. The corresponding COE can be projected to be in the economically competitive range of 62–54.6 mill/kWh.

A helium-cooled blanket design of the low aspect ratio reactor

An aggressive low aspect ratio scoping fusion reactor design (C.P.C. Wong, R. Cerbone, E.T. Cheng, R.L. Miller, R.D. Stambaugh, Proc. of 17th IEEE/NPSS Symp. on Fusion Engineering, pp. 1053) indicated that a 2 GW(e) reactor can have a major radius as small as 2.9 m resulting in a device with competitive cost of electricity at 49 mill/kWh. One of the technology requirements of this design is a high performance high power density first wall and blanket system. A 15 MPa helium-cooled, V-alloy and stagnant LiPb breeder first wall and blanket design was utilized. Due to the low solubility of tritium in LiPb, there is the concern of tritium migration and the formation of V-hydride. To address these issues, a lithium breeder system with high solubility of tritium has been evaluated. Due to the reduction of blanket energy multiplication to 1.2, to maintain a plant Q of >4, the major radius of the reactor has to be increased to 3.05 m. The inlet helium coolant temperature is raised to 430°C in order to meet the minimum V-alloy temperature limit everywhere in the first wall and blanket system. To enhance the first wall heat transfer, a swirl tape coolant channel design is used. The corresponding increase in friction factor is also taken into consideration. To reduce the coolant system pressure drop, the helium pressure is increased from 15 to 18 MPa. Thermal structural analysis is performed for a simple tube design. With an inside tube diameter of 1 cm and a wall thickness of 1.5 mm, the lithium breeder can remove an average heat flux and neutron wall loading of 2 and 8 MW/m 2, respectively. This reference design can meet all the temperature and material structural design limits, as well as the coolant velocity limits. Maintaining an outlet coolant temperature of 650°C, one can expect a gross closed cycle gas turbine thermal efficiency of 45%. This study further supports the use of helium coolant for high power density reactor design. When used with the low aspect ratio reactor concept a competitive fusion reactor can be projected at 51.9 mill/kWh.

Power plant design study of a high aspect ratio Tokamak using a SiC composite structure

The DREAM (DRastically EAsy Maintenance) Tokamak is a fusion power plant which is designed from the viewpoint of maintenance feasibility. For this purpose, the DREAM reactor uses a plasma with a very high aspect ratio ( A) and adopts SiC as a structural material. The choice of SiC affects the design of the core plasma, i.e. large inboard shield thickness, low synchrotron radiation reflectivity, and small plasma elongation for positional stability. The objectives of this study are to explore the feasibility of a high- A device, such as a power plant, and to clarify the technological impact of SiC material on the plasma design. Plasma size is optimized by the physics guidelines similar to ITER. The plasma major and minor radii of DREAM are 16 m and 2 m, respectively, and the average neutron wall load is 2.5 MW m −2, the maximum toroidal field is 20 T, and the fusion power is 5.5 GW. Steady-state operation is obtained with 50 MW of external current-drive power and 90% bootstrap current. The divertor heat load is estimated to be about 10 MW m −2. A radiative divertor concept is adopted to achieve a low divertor plasma temperature. The DREAM Tokamak concept is found to be a possible candidate for a future power plant with more than 5 GW of fusion power and an acceptable divertor condition.

A cost minimized tokamak reactor using conservative coils/cooling technology

One attractive concept for future commercial reactors is one of low cost with minimum advanced technology. Cost-minimized reactors with different operating modes are compared to reveal the most cost-minimized reactor with impact of deign parameters on the cost of electricity (COE). Three kinds of operating modes are considered; i.e. first stability region (FS), second stability region (SS) and reversed shear (RS). Deuterium–tritium fueled, 1000 MW(e) reactors with conventional aspect ratio are assumed. Several criteria for the parameter survey are used; for example, (1) thermal-to-electricity conversion efficiency is assumed to be 34.5% with water as a coolant, (2) average neutron wall load must not exceed 5 MW m −2 with a plasma major radius >5 m, (3) 2 MeV neutral beam injector is applied. It is concluded that the RS operating mode is the most cost-minimized. The results indicated that attaining high β N (the Troyon beta coefficient), low q 95 (safety-factor on the 95% flux surface), and high f bs (fraction of bootstrap current) is the best way to obtain a cost-minimized reactor. This cost-minimized RS reactor is advantageous because it can be designed without using advanced coil technology nor an advanced cooling system.

Transmutation of fission products by high-field tokamak reactor with force-balanced coils

A DT sub-ignited tokamak reactor is investigated as a transmuter of fission products with an average neutron wall loading of 1 MW m −2 basing on a novel coil concept to generate strong magnetic fields. By the use of the force-balanced coils (FBC) which reduces the net centering force by a factor of ten from that exerted on toroidal magnetic coils of the same dimension, the steady-state tokamak reactor can be designed to achieve the fusion reaction of Q DT≃3.7. To transmute the fission product 129I effectively, a blanket is designed which consists of the neutron multiplier Be, the tritium breeder Li 2O and a neutron reflector C. Using the optimal blanket, the amount of 129I is effectively halved in 2 years with quasi-self-consistent condition, Q NET≃5.3, local TBR>1.7 and nuclear heating rate in superconducting magnets <1 mW cm −3.

Study of design parameters for minimizing the cost of electricity of tokamak fusion power reactors

The impact of the design parameters on the cost of electricity (COE) is studied through a parameter survey in order to minimize the COE. Three kinds of operating modes are considered; first stability (FS), second stability (SS) and reversed shear (RS). The COE is calculated by a coupled physics-engineering-cost computer system code. Deuterium-tritium type, 1000 MW(e) at electric bus bar, steady state tokamak reactors with aspect ratios A from 3 to 4.5 are assumed. Several criteria are used for the parameter survey; for example, (a) the thermal to electrical conversion efficiency is assumed to be 34.5% using water as a coolant; (b) the average neutron wall load must not exceed 5 MW/m2 for plasma major radius Rp > 5 m; (c) a 2 MeV neutral beam injector (NBI) is applied. It is found that the RS operating mode most minimizes the COE among the three operating modes by reducing the cost of the current drive and the coils and structures. The cost-minimized RS reactor can attain high fbs, high βN and low q95 at the same time, which results in a short Rp of 5.1 m, a low Bmax of the maximum magnetic toroidal field (TF) of the TF coils of 13 T and a low A of 3.0. It can be concluded that this cost-minimized RS reactor is the most cost-minimized within the frameworks of this study. This cost-minimized RS reactor has two advantages: one is that a Bmax = 13 T TF coil can be made by use of ITER coil technology and the other is that the same cooling technology as that of ITER (water cooling) can be used.

Engineering Design Issues of a Low Aspect Ratio Tokamak Volumetric Neutron Source

Engineering design issues of a volumetric neutron source (VNS) based on a steady state low aspect ratio DT tokamak are presented. At the present the major radius is 0.8 m, the minor radius 0.6 m for an aspect ratio of 1.33, the plasma current is 10.1 MA, the toroidal field at the major radius is 1.8 T, and the fusion power is 39 MW giving an average neutron wall loading of 1.0 MW/m{sup 2} on the outboard side with an available testing area of 10 m{sup 2}. Two neutral beams delivering more than 20 MW are used to drive the steady state fusion plasma. A single turn unshielded water cooled dispersion strengthened (DS) Cu centerpost is used in conjunction with a conducting Cu bell jar which acts as a vacuum boundary and the return legs for the toroidal field (TF) coils. The centerpost is 9 m long, carries 7.2 MA and is specially shaped to minimize ohmic heating, which is calculated using temperature dependent DS Cu properties and increases in resistivity due to nuclear transmutations are accounted for. A naturally diverted plasma scrapeoff layer dominated by pressure-driven instabilities is assumed giving a peak heat flux of 5.2 MW/m{sup 2} onmore » the diverter plates. Fabrication approaches for the centerpost and its replacement time lines have been estimated to be feasible and reasonable. 5 refs., 6 figs., 2 tabs.« less

Exploration and assesment of design windows for a Tokamak-based volumetric neutron source

Abstract Design options for a Tokamak volumetric neutron source (VNS) for fusion nuclear technology testing are explored using well-defined and consistent sets of requirements and constraints. Optimum designs with a peak value of the neutron wall load and minimum cost are obtained with intermediate-to-high aspect ratios of about 3.4–3.6. Sensitivity analysis shows that the nuclear technology testing requirements can be achieved in a realistic Tokamak VNS design within the expected uncertainty ranges of the physics and engineering assumptions database. A typical optimum VNS design has an average neutron wall load of 1.5 MW m−2, peak wall load of about 2 MW m−2, major radius of about 1.6 m. minor radius of about 0.5 m. aspect ratio of 3.6, fusion power of about 100 MW, plasma amplification factor Q ≈ 2 and on-axis magnetic field of about 7 T. Normal conducting magnets are used with an overall inboard shield thickness of 0.3 m.

Status of design of the first wall for ITER

The First Wall (FW) is one of the key components of the International Thermonuclear Experimental Reactor (ITER). It must be designed to remove the surface heat flux from the plasma, while, in combination with a structurally integral shield, reduce the nuclear effects in the vacuum vessel and protect the superconducting coils from excessive nuclear heating and radiation damage. As the principal plasma facing element it will be exposed to the 1500 MW fusion power at an average neutron wall loading of about 1 MW/m2 and must be designed for a life of 0.3 MWa/m2 during the Basic Performance Phase. In addition the First Wall must successfully withstand peaked thermal and electromagnetically loads resulting from plasma disruptions.A conceptual design of the ITER First Wall has evolved as a result of the Engineering Design Activity. This design utilizes a beryllium plasma facing layer, bonded onto a copper layer which conducts the heat to the contained active cooling tubes. The copper is bonded to stainless steel for structural support and attachment to the water cooled stainless steel shield section. As discussed in this paper, the basic FW configuration developed for use at most locations, referred to as the primary wall, must be modified for different power requirements at two specific locations, the plasma limiter and the baffle. Carbon Fiber Composite (CFC) is being considered as an alternative material for the baffle and limiter FW because of its high thermomechanical performance and its high resistance to thermal shock.

Shielding analysis for the ITER divertor and vacuum-pumping ducts

Abstract Three-dimensional shielding analysis has been performed for the ITER outline design presented to TAC-4 in January 1994. The average neutron wall loading in ITER is 0.913 MW m−2 for the nominal 1500 MW fusion power. The peak outboard and inboard neutron wall loadings are 1.193 and 0.923 MW m−2 respectively. The divertor modules along with the vacuum vessel provide adequate shielding for the toroidal field (TF) coils in the divertor region. The walls of the horizontal divertor ducts are completely out of the direct line of sight of source neutrons. The contribution to total nuclear heating from streaming into the 24 divertor ducts is only 0.21 kW without additional duct shielding. No separate divertor duct shielding is needed and the duct wall 20 cm thick along with the mechanical structure 11 cm thick will provide adequate protection for the TF coils against streaming radiation. The values of total nuclear heating in the TF coils are only 4.14 and 4.79 kW for the stainless steel-water blanket-shield and LiV breeding-blanket designs respectively.

Convertible liquid metal blankets for ITER with Pb-17Li as breeding material

A convertible blanket concept is proposed for ITER, where, without replacement of the blanket structure, a non-breeding Pb alloy is used during the basic performance phase and the eutectic Pb-17Li during the enhanced performance phase. The concept is based on austenitic steel as structural material, an average neutron wall load of 1 MW m −2 and either helium or water as coolant. The same design concept was used for both coolant options with respect to a stiff blanket segment box, direct cooling of the first wall using toroidal ducts, poloidal hairpin tubes to cool the quasi-stagnant liquid metal and tritium removal outside the vacuum vessel. Various design options were considered for the first-wall and pool cooling and corresponding headers. Owing to the different coolant properties, different combinations were selected for the two versions. The performance of the two versions was assessed among other things with respect to tritium breeding and control, reliability and R&D needs.

A European proposal for an ITER water-cooled solid breeder blanket

Abstract The water-cooled solid breeder blanket concept proposed here aims to replace the shielding blanket for the enhanced performance phase of the international thermonuclear experimental reactor (ITER). The nominal performances are as follows: an average neutron wall load of 1 MW m−2 which corresponds to a fusion power of about 1.5 GW, and an average neutron fluence of 1 MWy m−2. The proposed blanket concept has been designed to accept a power increase of about 30% and power transients up to 3–5 GW for a short time. This blanket concept is based on a breeder inside tube (BIT)-type blanket with poloidal breeding elements made of 316 L-type stainless steel and filled with lithium metazirconate and beryllium pebbles. Inlet and outlet water temperatures of 160 and 200°C have been considered with a medium-pressure cooling system during plasma burn. The diameters of the breeding elements are compatible with the space available in test fission reactor core channels, making in-pile testing, required for blanket development and qualification, easier. A conservative approach using qualified materials, a blanket concept easily testable in fission reactors and on-going mock-up testing, which can be qualified using blanket test modules during the basic performance phase of ITER, will allow the blanket reliability required for the enhanced performance phase to be achieved.

CFFTP Fusion Pilot Plant Scoping Study

The CFFTP Pilot plant concept is a driven, steady-state H-mode tokamak with ion cyclotron current drive. The fuel cycle uses low-tritium-inventory technologies, including compact toroid fuelling. The mechanical design is based on helium cooling, radial blanket maintenance, ceramic pebble breeder blanket, and demountable copper magnets. It is expected to operate for 1 full-power-year with 0.25 MW/m{sup 2} average (0.4 MW/m{sup 2} peak) neutron wall load. The machine would produce 20 MW of fusion power with 40 MW of auxiliary power. The 2.7 MA plasma current is ramped up inductively, and then sustained by the bootstrap current (32%) and fast wave current drive (68%). The plasma would be roughly the size of the TFTR plasma, but elongated with double-null divertors and an aspect ratio of 5. The total electric power consumption would be around 450 MWe. The tritium supply requirements, given partial breeding, would be only 0.8 kg. The on-site tritium inventory would be about 0.3 kg. The direct construction cost is estimated at 1.1 B$Cdn, with a total project cost of 2.3 B$Cdn (1992). CFFTP Pilot would provide steady-state integrated nuclear testing at a fluence and neutron wall load of about one-quarter ITER CDA, for a cost of about one-thirdmore » ITER CDA. The blanket test area would be similar to the ITER CDA blanket test port area.« less

Activation analyses for different fusion structural alloys

The leading candidate structural materials, viz., the vanadium alloys, the nickel or the manganese stabilized austenitic steels, and the ferritic steels, are analysed in terms of their induced activation in the TPSS fusion power reactor. The TPSS reactor has 1950 MW fusion power and inboard and outboard average neutron wall loading of 3.75 and 5.35 MW/m 2, respectively. The results show that, after one year of continuous operation, the vanadium alloys have the least radioactivity at reactor shutdown. The maximum difference between the induced radioactivity in the vanadium alloys and in the other iron-based alloys occurs at about ten years after reactor shutdown. At this time, the FW-vanadium activation is three orders of magnitude less than other alloys' FW activation.

ARIES-I fusion-power-core engineering

The ARIES research program is a multi-institutional project, the goal of which is to determine the economic, safety, and environmental potential of tokamak fusion reactors. The ARIES-I steady-state tokamak reactor is a conceptual, DT-burning, 1000 MWe reactor with a major radius of 6.75 m, a minor radius of 1.5 m, and an average neutron wall loading of 2.5 MW/m 2. The ARIES-I plasma operates in the first MHD stability regime with a toroidal beta of 1.9%. The choice to operate in the first stability regime, with a high aspect ratio and with a low plasma current, leads to the need for high magnetic field to achieve adequate fusion power density ( β 2 B 4). The toroidal field at the plasma center is 11 T and the maximum field at the coil is 21 T. Nonetheless, it is found that the maximum stress in the structural material of these magnets is ∼ 700 MPa and industrially available alloys can be used. The impurity-control and particle-exhaust system is based on a high recycling double-null divertor system. The low-activation silicon-carbide (SiC) composite is used as structural material. The breeder material, Li 2ZrO 3, and the multiplier material, Be, are both sphere-packed between poloidally nested SiC-composite shells. The divertor plates consist of SiC-composite tube shells protected with 2 mm-thick tungsten armor. The first wall, blanket, shield, and divertor are all helium cooled with an inlet coolant temperature of 350°C at a pressure of 10 MPa. The high helium-outlet temperature of 650°C ensures a relatively high gross thermal efficiency of 49%. The ARIES-I design has demonstrated that tokamak reactors have the potential to achieve a high level of safety coupled with a Class-C waste-disposal rating.