Neutronic analysis of an inertial fusion energy breeder with SiCf/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 ThO₂ and cooled with natural lithium or (LiF)₂BeF₂ or Li₁₇Pb₈₃ 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), Li₁₇Pb₈₃ 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 B₄C shielding.     

Neutronic analysis of PROMETHEUS reactor fueled with various compounds of thorium and uranium

In this study, neutronic performance of the DT driven blanket in the PROMETHEUS-H (heavy ion) fueled with different fuels, namely, ThO₂, ThC, UO₂, UC, U₃Si₂ and UN is investigated. Helium is used as coolant, and SiC is used as cladding material to prevent fission products from contaminating coolant and direct contact fuel with coolant in the blanket. Calculations of neutronic data per DT fusion neutron are performed by using SCALE 4.3 Code. M (energy multiplication factor) changes from 1.480 to 2.097 depending on the fuel types in the blanket under resonance-effect. M reaches the highest value in the blanket fueled with UN. Therefore, the investigated reactor can produce substantial electricity in situ. UN has the highest value of ²³⁹Pu breeding capability among the uranium fuels whereas UO₂ has the lowest one. ²³⁹Pu production ratio changes from 0.119 to 0.169 according to the uranium fuel types, and ²³³U production values are 0.125 and 0.140 in the blanket fueled with ThO₂ and ThC under resonance-effect, respectively. Heat production per MW (D,T) fusion neutron load varies from 1.30 to 7.89 W/cm³ in the first row of fissile fuel breeding zone depending on the fuel types. Heat production attains the maximum value in the blanket fueled with UN. Values of TBR (tritium breeding ratio) being one of the most important parameters in a fusion reactor are greater than 1.05 for all type of fuels so that tritium self-sufficiency is maintained for DT fusion driver. Values of peak-to-average fission power density ratio, Γ, are in the range of 1.390 and ∼1.476 depending on the fuel types in the blanket. Values of neutron leakage out of the blanket for all fuels are quite low due to SiC reflector. The maximum neutron leakage is only ∼0.025. Consequently, for all cases, the investigated reactor has high neutronic performance and can produce substantial electricity in situ, fissile fuel and tritium required for (D,T) fusion reaction.     

Fission-suppressed blankets for fissile fuel breeding fusion reactors

Two blanket concepts for deuterium-tritium (DT) fusion reactors are presented which maximize fissile fuel production while at the same time suppress fission reactions. By suppressing fission reactions, the reactor will be less hazardous, and therefore easier to design, develop, and license. A fusion breeder operating a given nuclear power level can produce much more fissile fuel by suppressing fission reactions. The two blankets described use beryllium for neutron multiplication. One blanket uses two separate circulating molten salts: one salt for tritium breeding and the other salt for U-233 breeding. The other uses separate solid forms of lithium and thorium for breeding and helium for cooling.Nuclear power is the sum of fusion (D + T 14 MeV neutron+ 3.5 MeV alpha) power plus additional power from neutron-induced reactions in the blanket.     

Study of thorium–uranium based molten salt blanket in a fusion–fission hybrid reactor

Not only solid fuels, but also liquid fuels can be used for the fusion–fission symbiotic reactor blanket. The operational record of the molten salt reactor with F–Li–Be was very successful, so the F–Li–Be blanket was chosen for research. The molten salt has several features which are suited for the fusion–fission applications.The fuel material uranium and thorium were dissolved in the F–Li–Be molten salt. A combined program, COUPLE, was used for neutronics analysis of the molten salt blanket. Several cases have been calculated and compared. Not only the influence of the different fuels have been studied, but also the thickness of the molten salt, and the concentration of the ⁶Li in the molten salt.Preliminary studies indicate that when thorium–uranium–plutonium fuels were added into a F–Li–Be molten salt blanket and with a component of 71% LiF–2% BeF₂–13.5% ThF₄–8.5% UF₄–5% PuF₃, and also with the molten salt thickness of 40 cm and natural concentration of ⁶Li, the appropriate blanket energy multiplication factor and TBR can be obtained.The result shows that thorium–uranium molten salt can be used in the blanket of a fusion–fission symbiotic reactor. The research on the molten salt blanket must be valuable for the design of fusion–fission symbiotic reactor.     

Temperature distribution in mixed ThO2–UO2 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/m². The fuel rods containing ThO₂ and UO₂ 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 UO₂ 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 ThO₂ 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 UO₂ 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.     

Neutronics analysis of the power flattening and minor actinides burning in a thorium-based fusion–fission hybrid reactor blanket

A neutronics analysis has been performed for a thorium fusion breeder with a special task of burning minor actinide ²³⁷Np, ²⁴¹Am, ²⁴³Am, and ²⁴⁴Cm, and production of ²³³U for the future PWR application. Under a first wall fusion neutron wall loading of 0.1 MW/m² by a plant factor of 100%, preliminary neutronics calculations have been performed using the one-dimensional transport and burnup calculation code BISONC and the Monte Carlo transport code MCNP. To obtain a quasi-constant nuclear heat production density, 11 fuel rods containing the mixture of ThO₂ and minor actinides are placed in a radial direction in the fissile zone where ThO₂ is mixed with variable amounts of minor actinides. Calculation results show that the tritium breeding ratio is greater than 1.05 for both investigated Cases A and B, and the hybrid reactor is self-sufficient in the tritium required for the (DT) fusion driver in those models during the operation period. The blanket energy multiplication factor M, varies between 13.8 and 29.6 depending on the fuel types at the end of the operation period. The peak-to-average fission power density ratio (Γ) is less than 1.66 and 1.68 for both Cases A and B, respectively during the operation time. After 720 days of plant operation, the accumulated ²³³U is 1277 and 1725 kg in the blanket for the Cases A and B, respectively.     

The UWMAK-II study: A conceptual design of a helium cooled, solid breeder, Tokamak fusion reactor system

UWMAK-II is a conceptual design study of a low ß, circular Tokamak fusion power reactor. The aim of the study has been to perform a self-consistent analysis of a probable future fusion power system based on the philosophy that design decisions, wherever possible, should be conservative and should be based on present technology. As such, this system will not be the smallest, the least expensive, or the optimum Tokamak reactor. Rather, it represents a feasible system which we use to assess the technological problems uncovered and to examine possible solutions. The plasma is designed to generate 5000 MW(th) during a pulse and 1709 MW(e) continuously based upon a burn cycle with a 90 min burn and a 6.5 min rejuvenation period. The plasma carries a current of 14.9 MA and is designed with a double null poloidal divertor for impurity control and particle pumping. In addition, a low Z liner in the form of a carbon curtain is included to eliminate any source of high Z impurities. Plasma heating to ignition involves the use of neutral beam heating for a 10 sec period during which 200 MW of 500 keV deuterium atoms are injected into the plasma.The blanket design employs helium cooling and the solid lithium-bearing compound, lithium aluminate (Li₂Al₂O₄) for breeding tritium. The structural material is 316 stainless steel and beryllium is used as a neutron multiplier. The neutron radiation environment produces radiation damage that considerably influences blanket and system performance. The most significant effect is the loss of ductility which appears to limit the usable lifetime of the blanket first wall to about 2 yr at a 14 MeV neutron wall loading of 1.16 MW/m². The solid breeder blanket minimizes the tritium inventory but because of the low fractional burnup in the plasma and the need for roughly a one day reserve of fuel, the inventory is 17.7 kg. Induced radioactivity levels in the structure are of the order of 1 Ci/W(th) at shutdown after two years of operation. The main contributors to the activity are ) and ). Afterheat levels are slightly above 1% of thermal power but the afterheat power density is low, less than 0.1 w/g. The power cycle involves a He---Na intermediate heat exchanger followed by a sodium—steam system. The sodium intermediary is used to minimize tritium leakage through the power cycle and to provide a working fluid for thermal energy storage such that continuous electrical output is produced despite a pulse plasma cycle. A materials resource study has been completed for a UWMAK-II type system and beryllium appears to present a particular problem with regard to adequate resources. Other materials that could present problems of procurement include chromium and nickel. A preliminary economic analysis has been carried out to identify major cost areas and this is described.     

Utilization of Ceramic Uranium Fuels in ARIES-RS Fusion Reactor

This study presents the neutronic performance of the ARIES-RS fusion reactor design using different natural ceramic uranium fuels, namely UO₂, UN or U₃Si₂, dispersed in graphite matrix. These fissionable fuels inserted as micro spheres into the first range quadratic channels at the immediate neighborhood of the first wall in the inboard blanket to amplify fusion power and breed fissile fuel. Neutron transport calculations were performed with the help of the SCALE4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S₈–P₃ approximation. Among the investigated fuels, UN showed the best neutronic performance while UO₂ and U₃Si₂ had similar performances. Numerical results pointed out that inserting fissionable fuel zone even with a small thickness (10 cm) in a pure fusion reactor increased fusion power from 2170 MW to 4500, 5250 and 4150 MW depending on the fuel type. Furthermore significant amount of fissile fuel was produced to be charged to light water reactors.     

Requirements for helium cooled pebble bed blanket and R&D activities

This work aims to give an outline of the design requirements of the helium cooled pebble bed (HCPB) blanket and its associated R&D activities. In DEMO fusion reactor the plasma facing components have to fulfill several requirements dictated by safety and process sustainability criteria. In particular the blanket of a fusion reactor shall transfer the heat load coming from the plasma to the cooling system and also provide tritium breeding for the fuel cycle of the machine. KIT has been investigating and developed a helium-cooled blanket for more than three decades: the concept is based on the adoption of separated small lithium orthosilicate (tritium breeder) and beryllium (neutron multiplier) pebble beds, i.e. the HCPB blanket. One of the test blanket modules of ITER will be a HCPB type, aiming to demonstrate the soundness of the concept for the exploitation in future fusion power plants. A discussion is reported also on the development of the design criteria for the blanket to meet the requirements, such as tritium environmental release, also with reference to the TBM.The selection of materials and components to be used in a unique environment as the Tokamak of a fusion reactor requires dedicated several R&D activities. For instance, the performance of the coolant and the tritium self-sufficiency are key elements for the realization of the HCPB concept. Experimental campaigns have been conducted to select the materials to be used inside the solid breeder blanket and R&D activities have been carried out to support the design. The paper discusses also the program of future developments for the realization of the HCPB concept, also focusing to the specific campaigns necessary to qualify the TBM for its implementation in the ITER machine.     

功率展平的压水堆乏燃料发电包层中子学初步研究

聚变裂变混合堆在增殖核燃料、嬗变长寿命核废料及固有安全性等方面具有较大优势,同时,它比纯聚变堆在工程及技术方面要求低,因此较聚变堆更易实现。本工作基于目前国际聚变实验堆（ITER）所能达到的技术水平,提出一种直接利用乏燃料进行发电的聚变裂变混合堆包层概念,利用在不同位置放置不同乏燃料体积分数的方法对燃料增殖区实现了功率展平。计算结果表明：功率展平后的包层功率不均匀系数更小,且包层中燃料区的能量输出要比不展平情况下的能量输出高约21.7%。燃料富集度到运行末期最大可达5.23%。从中子学角度初步论证了该包层的可行性。     

Reaction balance and efficiency analysis of a D-D fusion/fission hybrid with satellite D-3He reactors

Selected reactor physics and isotope balance characteristics of a fusion hybrid supported D-³He satellite nuclear energy system are formulated and investigated. The system consists of two types of reactors: a parent D-fueled fusion device and a number of smaller reactors optimized for D-³He fusion. The parent hybrid station breeds the helium-3 for the satellites and also breeds fissile fuel for an existing fission reactor economy. Various hybrid operational regimes are examined in order to determine favorable reactorQ values and effective fusion and fission efficiencies. A number of analytical correlations between power output, plasma energetics, blanket neutronics, breeding capacity, and energy conversion cycles are established and evaluated. Numerical examples of performance parameters such as fission-to-fusion power, overall conversion efficiency, and the ratio of satellite to parent fusion power are presented. The range of reactor efficiencies is elucidated as affected by the internal plasma power balances. As an upper bound based on optimistic injection and direct conversion efficiencies, we find the D-³He satellite system power output attaining at best 1/3 of the parent fusion power.     

Octalithium plumbate as breeding blanket ceramic: Neutronic performances,synthesis and partial characterization

A neutronic assessment of the performances of a helium-cooled Li₈PbO₆ breeding blanket (BB) for the conceptual design of a DEMO fusion reactor is given. Different BB configurations have been considered in order to minimize the amount of beryllium required for neutron multiplication, including the use of graphite as reflector material. The calculated neutronic responses: tritium breeding ratio (TBR), power deposition in TF coils and power amplification factor, indicate the feasibility of Li₈PbO₆ as breeding material. Furthermore, the synthesis and characterization of Li₈PbO₆ by X-ray phase analysis are also discussed.     

The fusion breeder

The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the U.S. fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the U.S. fusion program and the U.S. nuclear energy program. There is wide agreement that many approaches will work and will produce fuel for five equal-sized LWRs, and some approach as many as 20 LWRs at electricity costs within 20% of those at today's price of uranium ($30/lb of U₃O₈). The blankets designed to suppress fissioning, called symbiotes, fusion fuel factories, or just fusion breeders, will have safety characteristics more like pure fusion reactors and will support as many as 15 equal power LWRs. The blankets designed to maximize fast fission of fertile material will have safety characteristics more like fission reactors and will support 5 LWRs. This author strongly recommends development of the fission suppressed blanket type, a point of view not agreed upon by everyone. There is, however, wide agreement that, to meet the market price for uranium which would result in LWR electricity within 20% of today's cost with either blanket type, fusion components can cost severalfold more than would be allowed for pure fusion to meet the goal of making electricity alone at 20% over today's fission costs. Also widely agreed is that the critical-path-item for the fusion breeder is fusion development itself; however, development of fusion breeder specific items (blankets, fuel cycle) should be started now in order to have the fusion breeder by the time the rise in uranium prices forces other more costly choices.     

NAPTH: Neutronics analysis code system for the fusion–fission hybrid reactor with pressure tube type blanket

The fusion–fission hybrid reactor is considered as a potential path to the early application of fusion energy. A new concept with pressure tube type blanket has recently been proposed for a feasible hybrid reactor. In this paper, a code system for the neutronics analysis of the pressure tube type hybrid reactor is developed based on the two-step calculation scheme: the few-group homogeneous constant calculation and the full blanket calculation. The few-group homogeneous constants are calculated using the lattice code DRAGON4. The blanket transport calculation is performed by the multigroup Monte Carlo code. A link procedure for fitting the cross sections is developed between these two steps. An additional procedure is developed to calculate the burnup, power distribution, energy multiplication factor, tritium breeding ratio and neutron multiplication factor. From some numerical results, it is found that the code system NAPTH is reliable and exhibits good calculation efficiency. It can be used for the conceptual design of the pressure tube type hybrid reactor with precise geometry.     

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/m² is performed recently. Three sets parameters of core plasma for different DEMO design objectives are proposed. A helium-cooled blanket system with ceramic breeder (Li₄SiO₄), 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.     

Neutronics Calculations in Pellets and Blankets of Laser Fusion Reactor Concept SENRI-I

The neutronic properties of SENRI-I, a reference design of laser fusion reactor proposed by Institute of Engineering, Osaka University, are discussed on the basis of the one-dimensional neutron transport calculations in burning DT plasmas and blankets. The softening of the fusion neutron energy spectrum, the neutron heating and the neutron multiplication are studied and discussed for the compressed DT pellets with various thickness of fuel plasmas and lead or lead-polyethylene tampers.

The neutronic and thermal features in the blanket of the SENRI-I design are also examined. The tritium breeding ratio is high enough (~1.6), depending on the neutron energy spectrum from a pellet. The maximum temperature increase per 1,000 MJ DT fusion reactions is ~3°C in the inner liquid Li layer and ~1.5°C in the stainless steel first wall. A parametric study is also presented on the effect of varying the thickness of the inner Li blanket ΔRⁱ to examine the thickness required for the enough tritium breeding ratio and energy deposition.     

Analysis of Time-Dependent Tritium Breeding Capability of Water Cooled Ceramic Breeder Blanket for CFETR

Attaining tritium self-sufficiency is an important mission for the Chinese Fusion Engineering Testing Reactor(CFETR) operating on a Deuterium-Tritium(D-T) fuel cycle. It is necessary to study the tritium breeding ratio(TBR) and breeding tritium inventory variation with operation time so as to provide an accurate data for dynamic modeling and analysis of the tritium fuel cycle. A water cooled ceramic breeder(WCCB) blanket is one candidate of blanket concepts for the CFETR. Based on the detailed 3D neutronics model of CFETR with the WCCB blanket,the time-dependent TBR and tritium surplus were evaluated by a coupling calculation of the Monte Carlo N-Particle Transport Code(MCNP) and the fusion activation code FISPACT-2007.The results indicated that the TBR and tritium surplus of the WCCB blanket were a function of operation time and fusion power due to the Li consumption in breeder and material activation.In addition, by comparison with the results calculated by using the 3D neutronics model and employing the transfer factor constant from 1D to 3D, it is noted that 1D analysis leads to an over-estimation for the time-dependent tritium breeding capability when fusion power is larger than 1000 MW.     

Physical investigation for neutron consumption and multiplication in fusion–fission hybrid test blanket module

Inelastic scattering of high energy fusion neutrons does affect the performance of fusion blanket based on the choice of different materials. It will also affect the behavior of source neutrons in a subcritical fusion fission hybrid blanket and consequently the transmutation and tritium breeding performance. A fusion fission hybrid test blanket module (HTBM) is designed which is presumed to be tested in a large sized tokamak and plasma neutron source is similar to ITER. In this preliminary design of HTBM the neutron source and loss factors are computed for the detailed neutronic performance analysis. The neutronic analysis of hybrid blanket module is performed for five different TRU fuel types: TRU-Zr, TRU-Mo, TRU-Oxide, TRU-Carbide and TRU-Nitride. In this module design, it is aimed to burn and transmute the TRU nuclides from high-level radioactive waste of PWR spent fuel. The effect of TiC reflector on transmutation and tritium breeding performance of HTBM is also quantified. MCNPX is used for neutronic computations. Neutron spectrum, capture to fission ratio and waste transmutation ratio of each fuel type are compared to evaluate their waste transmutation performance. Tritium breeding ratio is also compared for two coolant options: Li and LiPb eutectic.     

A fusion–fission hybrid reactor with water-cooled pressure tube blanket for energy production

A fusion–fission hybrid reactor is proposed to achieve the energy gain of 3000 MW thermal power with self-sustaining tritium. The hybrid reactor is designed based on the plasma conditions and configurations of ITER, as well as the well-developed pressurized light water cooling technologies. For the sake of safety, the pressure tube bundles are employed to protect the first wall from the high pressure of coolant. The spent nuclear fuel discharged from 33GWD/tU Light Water Reactors (LWRs) and natural uranium oxide are taken as driver fuel for energy multiplication. According to thermo-mechanics calculation results, the first wall of 20 mm is safe. The radiation damage analysis indicates that the first wall has a lifetime of more than five years. Neutronics calculations show that the proposed hybrid reactor has high energy multiplication factor, tritium breeding ratio and power density; the fuel cannot reach the level of plutonium required for a nuclear weapon. Thermal-hydraulic analysis indicates that the temperatures of the fuel zone are well below the limited values and a large safety margin is provided.