Blanket Coolant Research Articles

Power flattening of an inertial fusion energy breeder with mixed ThO 2–UO 2 fuel

Neutronic performance of a blanket-driven ICF (inertial confinement fusion) neutron and based on SiC f/SiC composite material has been investigated for fissile fuel breeding and a flat fission power density. The blanket is fueled with ThO 2 and UO 2 mixed by various mixing methods for achieving a flat fission power density, and cooled with different coolants, natural lithium, (LiF) 2BeF 2, Li 17Pb 83 and 4He for the nuclear heat transfer. MCNP4B code is used for calculations of neutronic data per DT fusion neutron. M (fusion energy multiplication rate) increases to ∼2.6 in the 4He coolant blanket. On the other hand, this value is ∼2.00 in the Li 17Pb 83 coolant blanket because of more capture reactions than fission reactions in this blanket. Therefore, the investigated reactor can produce substantial electricity in situ. Total fissile fuel breading ratio (FFBR) ( 232Th γ+ 238U γ) is increased from ∼0.16 to ∼0.37 by using coolants having high neutron multiplier cross-section. Values of TBR (tritium breeding ratio) being one of the main neutronic parameters for a fusion reactor are greater than 1.05 in all case of mixed fuels and type of coolants. The highest and the lowest TBR values are ∼1.34 and ∼1.07 in the natural Li and the 4He coolant blanket, respectively. Consequently, the breeder hybrid reactor is self-sufficient in the tritium required for the DT fusion driver. Γ (peek-to-average fission power density ratio) of the blanket is reduced to ∼1.1. Furthermore, quasi-constant fission power density profile is obtained in FFB (fissile fuel breeding) zone for case of PMF (parabolic mixed fuel) and all type of coolants. Values of neutron leakage out of the blanket in all case of mixed fuels and type of coolants are quite low due to SiC reflector and B 4C shielding. The maximum neutron leakage is only ∼0.025.

The EU advanced dual coolant blanket concept

The advanced dual coolant (A-DC) blanket is one of the EU advanced concepts to be investigated in the frame of the long-term power plant conceptual study (PPCS). Its basic concept—following the ARIES-ST concept—is based on the use of helium-cooled ferritic steel structure, the self-cooled Pb–17Li breeding zone, and SiC/SiC flow channel inserts. The latter serves as electrical and thermal insulators and therefore minimize the pressure losses and enable a relatively high Pb–17Li exit temperature leading to a high thermal efficiency. The present work on PPCS is drawn extensively on the preparatory study on plant availability (PPA) carried out in 1999 where a maximum neutron wall load of 5 MW/m 2 (corresponding maximum surface heat load of 0.9 MW/m 2) was given in the reference case of the A-DC blanket. In the following stage of PPCS the A-DC blanket is normalized and adapted to a typical size of commercial reactors (e.g. 1500 MWe) which requires iterative calculations between the blanket layout and the system code analysis. The status of the work with some idea improvements is reported.

Temperature distribution in mixed ThO 2–UO 2 fuel rods located in blanket of an inertial fusion energy breeder

Investigations of neutronic analysis and temperature distribution in fuel rods located in a blanket driven ICF (Inertial Confinement Fusion) have been performed for various mixed fuels and coolants under a first wall load of 5 MW/m 2. The fuel rods containing ThO 2 and UO 2 mixed by various mixing methods for achieving a flat fission power density are replaced in the blanket and cooled with different coolants; natural lithium, flibe, eutectic lithium and helium for the nuclear heat transfer. It is assumed that surface temperature of the fuel rod increases linearly from 500 °C (at top) to 700 °C (at bottom) during cooling fuel zone. Neutronic and temperature distribution calculations have been performed by MCNP4B Code and HEATING7, respectively. In the blanket fueled with pure UO 2 and cooled with helium, M (fusion energy multiplication ratio) increases to ∼3.9 due to uranium having higher fission cross-section than thorium. The high fission energy released in this blanket, therefore, causes proportionally increasing of temperature in the fuel rods to 823 °C. However, the M is ∼2.00 in the blanket fueled with pure ThO 2 and cooled with eutectic lithium because of more capture reaction than fission reaction. Maximum and minumum values of TBR (tritium breeding ratio) being one of main neutronic paremeters for a fusion reactor are ∼1.07 and ∼1.45 in the helium and the natural lithium coolant blanket, respectively. These consequences bring out that the investigated reactor can produce substantial electricity in situ during breeding fissile fuel and can be self-sufficient in the tritium required for the DT fusion driver in all cases of mixed fuels and coolant types. Quasi-constant fission power density profiles in FFB (fissile fuel breeding) zone are obtained by parabolically increasing mixture fraction of UO 2 in radial and axial directions for all coolant types. Such as, in the helium coolant blanket and the case of PMF (parabolically mixed fuel), Γ (peek-to-average fission power density ratio) of the blanket is reduced to ∼1.1, and the maximum temperatures of the fuel rods in radial direction of the FFB zone are also quasi-constant. At the same time, in the case of PMF, for all coolant types, the temperature profiles in the radial direction of the fuel rods rise proportionally with surface temperature from the top to the bottom of fuel rods in the axial direction. In other words, for each radial temperature profile in the axial direction, temperature differences between centerline and surface of the fuel rods are quasi-constant. According to the coolant types, these temperature diffences vary between 30 and 45 °C.

Investigation of the performance parameters and temperature distribution in fuel rod dependent on operation periods and first wall loads in fusion–fission reactor system fueled with ThO 2

The neutronic performance parameters, fissile breeding and temperature distribution in the fuel rod are investigated for different coolants (‘He, CO 2’, ‘Li 2BeF 4’, ‘Li’, and ‘Li 17Pb 83’) in the fissile fuel breeding zone with volume ratios of V coolant/ V fuel, ε, (0.5,1,2) under various first wall loads ( P w=2–10 MW m −2) in a fusion–fission reactor fueled with ThO 2. Depending on the type of coolant in the fission zone, first wall loads and volume ratios, fusion power plant operation periods between 1 and 4 years are evaluated to achieve a fissile fuel enrichment quality between 1.2274% and 13.7305% in the above mentioned situation for intervals of half month and by plant factor of 75%. A fusion reactor with (D,T) reaction acts as an external high energetic neutron source. The fissile fuel zone, containing 10 fuel rod rows in the radial direction, covers the cylindrical fusion plasma chamber with 300 cm chamber dimension. At the end of four years, the cumulative fissile fuel enrichment (CFFE) values, indicating rejuvenation performance, increased to 10.184%, 12.218%, 9.650% and 11.089% from 0% in gas, flibe, natural lithium and eutectic lithium coolant blankets with ε=0.5 for 10 MW m −2, respectively, without reaching the melting point of the fuel material. However, for ε=1, the CFFE values increased to 10.59% (the best CFFE value for gas coolant), 11.372%, 10.414% (the best CFFE value for natural lithium coolant) and 10.963% from 0% in the above mentioned coolants for 10 MW m −2, respectively. In the same way, for ε=2, the CFFE increased to 10.181%, 13.7305% (the best CFFE value for all coolants), 9.1369% and 11.4809% (the best CFFE value for eutectic lithium coolant) for 10 MW m −2, respectively. At the beginning of the operation period, for ε=0.5, the tritium breeding ratio (TBR) values, being about 0.9092, 0.7075, 0.7921 and 0.9512 for the above mentioned coolants, respectively, at the end of four years increased to 1.2924, 1.1475, 1.0724 and 1.441 (the highest TBR value for all blankets) for 10 MW m −2. For ε=1, these increments are 1.3067, 1.2303, 1.1653 and 1.4033 for 10 MW m −2. However, for ε=2, these values are 1.2256, 0.9993, 0.9341 and 1.4069 for 10 MW m −2 without reaching the melting point of the fuel material. For ε=0.5, the blanket energy multiplication ( M) increases to 2.7349, 2.8045, 2.5685 and 2.8183 (the highest M value for all blankets) for 10 MW m −2 from 2.1534, 2.0251, 2.0918 and 2.0793 in the blankets cooled with gas, flibe, natural lithium and eutectic lithium coolants, respectively, at the end of four years. These increments become 2.6707, 2.7301, 2.4332 and 2.7812 for 10 MW m −2 from 2.1037, 1.8985, 2.004 and 1.9874, respectively, for ε=1. However, the blanket energy multiplication ( M) increases to 2.4122, 2.4573, 2.1413 and 2.3800 for 10 MW m −2 from 2.0196, 1.7323, 1.8658 and 1.8303, respectively, for ε=2. The maximum temperatures in the centerline of the fuel rods have not exceeded the melting point of the fuel material for all coolants and ε under changing first wall loads between 2 and 10 MW m −2 during the operation periods. While the maximum CFFE values have been obtained in fuel rod row#10 in the gas, natural lithium and eutectic lithium coolant blankets, it has been obtained in fuel rod row#1 in the flibe coolant blanket for all ε and P w. Therefore, the investigated hybrid blankets are self-sufficient for all coolant and volume fractions and P w=10 MW m −2. The best neutron economy has been shown by flibe.

Neutronic analysis of an inertial fusion energy breeder with SiC f/SiC

Neutronic performance of a blanket driven ICF (Inertial confinement fusion) neutron based on SiC f/SiC composite material is investigated for fissile fuel breeding. The investigated blanket is fueled with ThO 2 and cooled with natural lithium or (LiF) 2BeF 2 or Li 17Pb 83 or 4He coolant. MCNP4B Code is used for calculations of neutronic data per DT neutron. Calculations have show that values of TBR (tritium breeding ratio) being one of the main neutronic paremeters of fusion reactors are greater than 1.05 in all type of coolant, and the breeder hybrid reactor is self-sufficient in the tritium required for the DT fusion driver. Calculations show that natural lithium coolant blanket has the highest TBR (1.298) and M (fusion energy multiplication) (2.235), Li 17Pb 83 coolant blanket has the highest FFBR (fissile fuel breeding ratio) (0.3489) and NNM (net neutron multiplication) (1.6337). 4He coolant blanket has also the best Γ (peek-to-average fission power density ratio) (1.711). Values of neutron leakage out of the blanket in all type of coolants are quite low due to SiC reflector and B 4C shielding.

The EU advanced lead lithium blanket concept using SiC f/SiC flow channel inserts as electrical and thermal insulators

Preparatory work on the EU advanced dual coolant (A-DC) blanket concept using SiC f/SiC flow channel inserts as electrical and thermal insulators has been carried out at the Forschungszentrum Karlsruhe in co-operation with CEA (SiC f/SiC composite-related issues) as a conceptual design proposal to the EU fusion power plant study planned to be launched in 2001 within the framework of the EU fusion programme having the main objective of specifying the characteristics of an attractive and viable commercial D–T fusion power plant. The basic principles and the study method for the A-DC blanket concept are presented in this report. The results of this study show that the A-DC blanket concept has a high potential for further development due to its high thermal efficiency and its simple concept solution.

Temperature distribution in nuclear fuel rod and variation of the neutronic performance parameters in (D–T) driven hybrid reactor system

Temperature distribution in nuclear fuel rod and variation of the neutronic performance parameters are investigated for different coolants under various first wall loads (Pw=2, 5, 7, 8, 9, and 10 MW m−2) in (D, T) (deuterium and tritium) driven and fueled with UO2 hybrid reactors. Plasma chamber dimension, DR, with a line fusion neutron source is 300 cm. The fissile fuel zone is considered to be cooled with four different coolants with various volume fractions, the volumetric ratio of coolant-to-fuel [ε(Vm/Vf) = 1:2, 1:1, and 2:1], gas (He, CO2), flibe (Li2BeF4), natural lithium (Li), and eutectic lithium (Li17Pb83). Calculation in the fuel rods and the behavior of the fissile fuel have been observed during 4 years for discrete time intervals of Δt=15 days and by a plant factor (PF) of 75%. As a result of the calculation, cumulative fissile fuel enrichment (CFFE) value indicating rejuvenation performance has increased by increasing Pw for all coolants and ε. Although CFFE and neutronic performance parameter values increase to the higher values by increasing Pw, the maximum temperature in the centerline of the fuel roads has exceeded the melting point (Tm>2830°C) of the fuel material during the operation periods. However, the best CFFE (11.154%) is obtained in gas coolant blanket for ε=1:2 (29.462% coolant, 58.924% fuel, 11.614% clad), under 10 MW m−2 first wall load, followed by flibe with CFFE=11.081% for ε=2:1 (62.557% coolant, 31.278% fuel, 6.165% clad), under 7 MW m−2, and flibe with CFFE=9.995% for ε=1:1 (45.515% coolant, 45.515% fuel, 8.971% clad), under 7 MW m−2 during operation period without reaching the melting point of the fuel material. While maximum CFFE value has been obtained in fuel rod row#10 in gas, natural lithium, and eutectic lithium coolant blankets, it has been obtained in fuel rod row#1 in flibe coolant blanket for all ε and Pw. At the same condition, the best neutronic performance parameter values, tritium breeding ratio (TBR)= 1.4454, energy multiplication factor (M)= 9.2018, and neutron leakage (L)= 0.0872, have been obtained in eutectic lithium coolant blankets for the ε=1:2, followed by gas, natural lithium, and flibe coolant blankets. The isotopic percentage of 240Pu is higher than 5% in all blankets for Pw⩾7 MW m−2, so that plutonium component in all blankets can be never reach a nuclear weapon grade quality during the operation period.

Recent progress of SiC-fibers and SiC/SiC-composites for fusion applications

Recent progress in R&D of SiC fibers and reinforced SiC matrix (SiC/SiC) composites in Japan, especially focusing on the activities of CREST-ACE program, is presented. Firstly, the present status of high performance SiC fiber development, such as Hi-Nicalon Type-S and Tyrano-SA, is provided. The high performance SiC matrix production by reaction sintering (RS) method improved in both strength and thermal conductivity are accomplished. The efforts to make appropriate fiber-matrix interfacial microstructure by CVI and PIP methods have been successful, resulting in the production of high strength and high fracture toughness SiC/SiC composites. Several joining processes using PIP, RS and mechanical fastener for composites are introduced. Dimensional stability under radiation damage has been studied by neutron and charged particle irradiation. The SiC/SiC composites prepared with Type-S SiC fiber with a stoichiometric composition did not exhibit mechanical property degradation. Based on the development of SiC composites, test module concepts to verify the advanced fluid systems including SiC/SiC/Be/He coolant blanket are presented.

Design and analysis of the vacuum vessel for RTO/RC-ITER

Recent progress in design and analysis of the vacuum vessel (VV) for the reduced technical objectives/reduced cost International Thermonuclear Experimental Reactor (RTO/RC-ITER) is presented. The basic VV design is similar to the previous ITER VV. However, because the back plate for the blanket modules could be eliminated, its previous functions could be transferred to the VV. For this option, the blanket modules are supported directly by the VV and the blanket coolant channels are structurally part of the VV double wall structure. In addition, a ‘tight fitting’ configuration is required to correctly position the modules’ first wall. Although such modifications of the VV complicate its structure and increase its fabrication cost, the design of the VV is considered to be still feasible. The structural analyses of the VV have been conducted using several FE models of the VV, including global and local models. Although further assessment is required, based on the analyses performed to date, the structural aspects of the VV for the case without the back plate appear feasible.

Coupling of Electromagnetics and Structural/Fluid Dynamics—Application to the Dual Coolant Blanket Subjected to Plasma Disruptions

Some aspects concerning the coupling of quasi-stationary electromagnetics and the dynamics of structure and fluid are investigated. The necessary equations are given in a dimensionless form. The dimensionless parameters in these equations are used to evaluate the importance of the different coupling effects. A finite element formulation of the eddy-current damping in solid structures is developed. With this formulation, an existing finite element method (FEM) structural dynamics code is extended and coupled to an FEM eddy-current code. With this program system, the influence of the eddy-current damping on the dynamic loading of the dual coolant blanket during a centered plasma disruption is determined. The analysis proves that only in loosely fixed or soft structures will eddy-current damping considerably reduce the resulting stresses. Additionally, the dynamic behavior of the liquid metal in the blankets` poloidal channels is described with a simple two-dimensional magnetohydrodynamic approach. The analysis of the dimensionless parameters shows that for small-scale experiments, which are designed to model the coupled electromagnetic and structural/fluid dynamic effects in such a blanket, the same magnetic fields must be applied as in the real fusion device. This will be the easiest way to design experiments that produce transferable results. 10 refs., 7 figs.

Thermal-hydraulic and structural design of the TITAN-I reversed-field-pinch fusion power core

Thermal-hydraulic and structural design of the first wall, blanket, and shield of the deuterium-tritium fueled TITAN-I reversed-field-pinch (RFP) fusion reactor is presented. Taking advantage of the characteristic low toroidal magnetic field of an RFP reactor, liquid lithium is used as the primary coolant to remove the thermal energy at an elevated temperature, thereby realizing a high power conversion efficiency of 44%. The use of liquid lithium has also led to a self-cooled design of the fusion power core in which the primary coolant is also the tritium breeder. The structural material is the vanadium alloy, V3Ti1Si. Tubular coolant channels are used in the first wall/blanket and rectangular channels in the hot shield. These are laid along the much larger poloidal field to minimize magnetohydrodynamic (MHD) pressure drop. Although the neutron wall loading of 18.1 MW/m2 is high, resulting in a radiation heat flux on the first wall of 4.6 MW/m2, three aspects of the design have made the removal of the reactor power at high temperature possible. These are: (1) the use of small-diameter circular tubes as coolant channels in the first wall, (2) the use of high-velocity MHD turbulent flow in the first-wall coolant tubes, and (3) thermal separation of the first-wall and blanket/shield coolant circuits, thereby allowing different exit temperatures. The thermal-hydraulic design was optimized by a design code developed for this purpose. Detailed structural design was performed by the finite element code ANSYS. The coolant inlet temperature is 320°C, and the coolant exit temperatures for the first-wall and blanket/shield coolant circuits are 442°C and 700°C, respectively. Lithium flow velocity in the first-wall coolant tubes is 21.6 m/s, and is ≤ 50 cm/s in the blanket/shield coolant channels. The total pressure drop in the first-wall coolant circuit is 10 MPa and in the blanket coolant circuit it is 3 MPa. The pumping power for coolant circulation is less than 5% of the net electric output. The material stresses are well within the design limits. The TITAN-I design suggests the feasibility and advantage of liquid-metal cooling of high wall loading RFP fusion reactors.

Safety and environmental aspects of organic coolants for fusion facilities

Organic coolants, such as OS-84, offer unique advantages for fusion reactor applications. These advantages are with respect to both reactor operation and safety. The key operational advantage is a coolant that can provide high temperature (350–400°C) at modest pressure (2–4 MPa). These temperatures are needed for conditioning the plasma-facing components and, in reactors, for achieving high thermodynamic conversion efficiencies (>40%). The key safety advantage of organic coolants is the low vapor pressure, which significantly reduces the containment pressurization transient (relative to water) following a loss of coolant event. Also, from an occupational dose viewpoint, organic coolants significantly reduce corrosion and erosion inside the cooling system and consequently reduce the quantity of activation products deposited in cooling system equipment. On the negative side, organic coolants undergo both pyrolytic and radiolytic decomposition, and are flammable. While the decomposition rate can be minimized by coolant system design (by reducing coolant inventories exposed to neutron flux and to high temperatures), decomposition products are formed and these degrade the coolant properties. Both heavy compounds and light gases are produced from the decomposition process, and both must be removed to maintain adequate coolant properties. As these hydrocarbons may become tritiated by permeation, or activated through impurities, their disposal could create an environmental concern. Because of this potential waste disposal problem, consideration has been given to the recycling of both the light and heavy products, thereby reducing the quantity of waste to be disposed. Preliminary assessments made for various fusion reactor designs, including ITER, suggest that it is feasible to use organic coolants for several applications. These applications range from first wall and blanket coolant (the most demanding with respect to decomposition), to shield and vacuum vessel cooling, to an intermediate cooling loop removing heat from a liquid metal loop and transferring it to a steam generator or heat exchanger.

Passive safety in a helium-cooled tritium breeding blanket with steel structure

Helium is attractive for use as a fusion blanket coolant for a number of reasons. It is neutronically and chemically inert, nonmagnetic, and will not change phase during any off-normal or accident condition. A significant disadvantage of helium, however, is its low density and volumetric heat capacity. This disadvantage manifests itself most clearly during undercooling accidents such as a loss of coolant accident (LOCA) or a loss of flow accident (LOFA). This paper proposes a new helium-cooled, tritium breeding blanket concept which uses a metallic structure, and which performs significantly better during such accidents than related designs. The proposed blanket uses modified, reduced-activation HT-9 steel as a structural material and is designed for neutron wall loads exceeding 4 MW/m2. This concept uses novel features such as: (1) a “beryllium-joint” design which allows beryllium to be used to conduct heat away from the first wall, while accommodating swelling of the beryllium, and (2) a shield cooled by naturally circulating water. These features help the blanket passively withstand a worst-case undercooling accident scenario.

Blanket and divertor design for the Steady State Tokamak Reactor (SSTR)

The tritium breeding blanket and plasma facing components (first wall and divertor) have been designed for the Steady State Tokamak Reactor (SSTR). Low activation ferritic steel (F82H) was chosen for the structural material of the first wall and the blanket. The ceramic breeder pebble bed concept with beryllium neutron multiplier was adopted. The blanket is divided into two parts; a replaceable blanket and a permanent blanket. Electrical insulation made of functionally gradient material (FGM) was introduced in the blanket box structure to drastically reduce electromagnetic loads during a plasma disruption and to simplify the attaching mechanism of the blanket. Low-temperature and high-density divertor plasma conditions and a radiative cooling concept lower the heat flux to the divertor plate to below 3 MW/m 2 and enable us to adopt the same cooling conditions as the blanket coolant. Puffing mechanisms of gas (deuterium) and iron are installed for the radiative cooling of the divertor plasma.

Transient flow behavior for a helium-cooled ceramic breeder blanket

This paper presents a thermal-hydraulic analysis of a helium-cooled ceramic breeder blanket for the ITER. It is found that the blanket can operate at moderate pressure and temperatures, which coupled to the use of an inert gas, makes it particularly attractive from a safety standpoint. Transient analysis of the manifolding flow distribution indicates that the introduction of a diffusive plate at the inlet of the blanket coolant flow path produces a more uniform coolant exit temperature. Control of the pump pressure head ensures that the coolant temperature is kept below its maximum limit, set by the maximum allowable operating temperature of the breeder.

An optimum rankine power cycle for the lithium-cooled TITAN-I reversed-field-pinch reactor

A thermal power cycle for the TITAN-I fusion reactor is presented. TITAN is a compact, high-power-density (neutron Wall loading of 18 MW/m 2 ) reactor based on the reversed-field-pinch (RFP) confinement concept. TITAN-I incorporates a fusion power core design with liquid lithium as the coolant and breeder, and vanadium alloy (V-3Ti-1Si) as the structural material. The total thermal power of TITAN-I is 2918 MW t . Of this total thermal power, 736 MW t is removed by the first-wall coolant, 29 MW t by the divertor coolant and the remaining 2153 MW t by the blanket coolant. The inlet temperature of the primary coolant is 320 °C. The exit temperatures are 440 °C, 540 °C and 700 °C for the first-wall, divertor and blanket coolants, respectively. Coolants from the first wall and divertor are mixed upon exit which gives the first-wall/divertor mixed exit temperature of 442 °C. The blanket coolant is kept separate. The coolant flow rates in the first-wall, divertor and blanket circuits are 1464, 31 and 1352 kg/ s, respectively. Liquid lithium is also chosen as the secondary coolant in the intermediate heat exchangers. Several power cycle modifications are considered. The selected power cycle has two components — one for the conversion of the thermal power in the first-wall and divertor coolants, and the other for conversion of the thermal power in the blanket coolant which is at much higher thermal potential. The power cycle for the first wall and divertor is a non-reheat, superheat Rankine cycle with 4 stages of regenerative feed-water heating. The pressure and temperature of the throttle steam are 10.7 MPa and 396 °C, respectively. This results in a gross thermal efficiency of 37.3%. The power cycle for blanket is a Rankine cycle with 2-stage reheat and 7 stages of regenerative feed-water heating. The throttle steam conditions are 565.6 °C and 21.4 MPa. The gross thermal efficiency for this cycle is 46.5%. The overall gross thermal efficiency of the lithium-cooled TITAN reactor is 44%. All the power cycle analyses are done by the code PRESTO developed at Oak Ridge National Laboratory in the United States.

Thermal-hydraulic and structural design for the lithium-cooled TITAN-I reversed-feeld-pinch reactor

A thermal-hydraulic and structural design of the first wall and blanket for the 18 MW/m 2 neutron wall-loading TITAN reversed-field-pinch fusion reactor is presented. The primary coolant is liquid lithium and the structural material is the vanadium alloy, V-3Ti-1Si. Various design limits, which take the operational environment into consideration, for this vanadium alloy for the first wall and blanket are discussed. The first wall is made of small-diameter tubes. The blanket coolant channels are a combination of tubular, square, and rectangular channels. The first-wall and blanket coolant circuits are separate, allowing different coolant exit temperatures. The coolant channels are aligned with the larger poloidal magnetic field to reduce magnetohydrodynamic (MHD) pressure drops. At the design heat flux of 4.6 MW/m 2 on the first wall, MHD turbulent-flow heat transfer is used. Both the separation of the coolant circuits and the use of MHD turbulent-flow heat transfer substantially increase the heat flux limit on the first wall. These are favorable for high-power-density fusion reactors. The inlet temperature of lithium is 320 °C and the exit temperatures are 440 °C and 700 °C for the first wall and blanket, respectively. The pressure drop in the first wall circuit is 10 MPa and is less than 3 MPa for the blanket circuit. The pumping power is less than 5% of electric output. The maximum structure temperature is less than 750 °C and the material stresses are shown to be within the allowable limits. One-dimensional thermal and stress analyses are adequate for the thermal-hydraulic and structural design for TITAN-I. This is verified by two-dimensional, finite element analysis. The use of liquid lithium as the coolant and vanadium alloy as the structural material enables the removal of the reactor thermal energy at high temperature, which has resulted in a gross thermal efficiency of 44%.

A conceptual design of RFP fusion power core — REPUTER-I

REPUTER-I is a conceptual design study on a compact reversed field pinch fusion reactor with a solid breeder blanket. The first wall is structurally independent from the blanket. The blanket is composed of SiC reinforced blocks which form a stable arch, and the stresses in SiC are basically compressive. Pressurized water is chosen for the first wall and as a limiter coolant in order to remove a high heat flux, while helium gas is used for the blanket coolant. A pile of thin Li 2 O tiles and thick beryllium tile provides a satisfactory tritium breeding ratio. A pumped limiter is shown to meet the demand for heat and particle exhaust removal in a compact RFP core. Magnet system design and plasma performance analysis are made as well on the first wall, blanket and limiter design.

Nitrate-Salt-Cooled Blanket Concepts

Molten nitrate and nitrate/nitrite salt-cooled blankets have the unique property of low-pressure operation when compared to other potential blanket coolants. Tokamak and mirror blanket designs have been developed with draw salt coolant, a lithium aluminate breeder, and a beryllium multiplier in an HT-9 ferritic steel structure. These blankets have outstanding economic performance and good engineering feasibility. The most evident problem is the high activation of this coolant, which affects system safety. There are also several critical data needs that must be resolved by experiment. This class of coolants merits further consideration for commercial fusion reactors.

Effect of first wall flaws on reactor performance

The influence of primary blanket coolant leaks through flaws in the first wall on Tokamak fusion plasma performance is investigated. A one-dimensional, three region radial diffusion model of impurity transport in the plasma is developed. The model includes plasma removal and recycling in the boundary layer and is used to correlate plasma performance with coolant leakage rates. In turn, coolant leak rates are estimated via molecular or viscous flow, as appropriate through both elastically loaded cracks and cracks opened by creep. Fatigue and environmentally induced subcritical crack growth and unstable crack growth flaw sizes are estimated for austenitic and ferritic stainless steels and compared to the coolant leak rate flaw sizes. The materials and plasma analyses are combined to yield estimates of the critical leak rate for three coolant candidates: helium, water, and lithium. The results indicate that the maximum-leak-rate (MLR) flaw sizes for helium, water and lithium are 5 to 8 mm, 6 to 8 mm and 1.5 to 3 × 10 4 mm, respectively, for elastically loaded cracks and 0.1 to 0.3 mm, 0.2 to 0.3 mm, and 700 to 1500 mm, respectively, for cracks opened by creep. The threshold flaw sizes for fatigue and corrosion fatigue subcritical crack growth of Type 316 stainless steel and HT-9 range from 0.2 to 2 mm and the flaw sizes for unstable crack growth of irradiated 316 stainless steel and HT-9 are 4 mm and 50 mm, respectively. These results suggest that the MLR sizes for helium and water are among the smallest flaws that may affect reactor performance and that the threshold for subcritical crack growth in fatigue is a critical material property. Also, creep processes are shown to have a significant effect on the MLR flaw size.