Blanket Energy Multiplication Research Articles

Neutronics analysis of thorium-uranium based molten salt blanket in stellarator-type FFHR

• Detailed CAD model has been designed for thorium-uranium based molten salt blanket. • Complex geometry structure has been analyzed in coupling with Monte Carlo calculation. • Neutronics features have been simulated in consideration of the burnup effect. • The thickness and the 6 Li enrichment have been optimized in the simulation process. The stellarator-type Fusion-Fission Hybrid Reactor (FFHR) is an alternative conceptual design for the tokamak-based devices in order to realize the stable power output. It can bring the viable energy source for the fusion technology in a foreseeable future, and has been widely studied with a lot of advanced concepts being proposed. However, there are some challenges in the neutronics design of the FFHR blanket for the purpose of obtaining the appropriate Tritium Breeding Ratio (TBR) and blanket energy multiplication factor, since the blanket system has a complicated three-dimensional structure in helical shape, which can not be easily translated into native geometry representation syntax of most Monte Carlo neutron transport codes. In this paper, a liquid fuel, F-Li-Be molten salt, with uranium and thorium dissolved has been proposed for the preliminary design of stellarator-type FFHR blanket. Besides, the CAD-PSFO code [ 19 ] has been employed for the simulation of complex geometry structure in the helical blanket, and the OMCB package [ 22 ] has been used in the burnup analysis of the neutronics features in the molten salt blanket. Moreover, detailed geometry design of the blanket arrangement and thickness have been studied in order to meet the requirement of energy multiplication factor in the stellarator-type FFHR. Also, the composition and concentration of the fuel material and 6 Li in molten salt have been discussed for the purpose of tritium self-sufficiency with steady energy supply in the whole lifetime.

Minor actinides transmutation in a molten salt blanket in the fusion-fission hybrid reactor core

Abstract Fusion-fission hybrid reactor (FFHR), a representative kind of subcritical nuclear reactor, has the advantage of better safety margin. There are less constraints for the loading of active fuel when burning spent nuclear fuel. A simplified three-dimensional model of tokamak blanket is constructed by the Serpent Monte Carlo Method Code to investigate the MA transmutation characteristics in the Flibe molten salt blanket. Based on the model with one homogenous fuel region, the effect of key design parameters: MA content, plutonium content, Li-6 enrichment in the initial fuel were determined, during which performance characteristics such as MA transmutation amount, keff, tritium breeding ratio (TBR), blanket energy multiplication factor M and conversion ratio were obtained. The results show that it is hard to improve the tritium breeding and meanwhile enhance the MA reducing in one homogenous fuel region. To balance these two functions, a double molten salt fuel regions model with separative MA transmutation zone and tritium breeding zone was proposed.

Monte Carlo analysis of LWR spent fuel transmutation in a fusion-fission hybrid reactor system

Abstract The aim of this paper is to determine neutronic performances of the light water reactor (LWR) spent fuel mixed with fertile thorium fuel in a FFHR. Time dependent three dimensional calculations for major technical data, such as blanket energy multiplication, tritium breeding ratio, cumulative fissile fuel enrichment and burnup have been performed by using Monte Carlo Neutron-Particle Transport code MCNP5 1.4, coupled with a novel interface code MCNPAS, which is developed by our research group. A self-sustaining tritium breeding ratio (TBR>1.05) has been kept throughout the calculations. The study has shown that the fissile fuel quality will be improved in the course of the transmutation of the LWR spent in the FFHR. The latter has gained the reusable fuel enrichment level conventional LWRs between one and two years. Furthermore, LWR spent fuel - thorium mixture provides higher burn-up values than in light water reactors.

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

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

Neutron transport-burnup code MCORGS and its application in fusion fission hybrid blanket conceptual research

Fusion science and technology has made progress in the last decades. However, commercialization of fusion reactors still faces challenges relating to higher fusion energy gain, irradiation-resistant material, and tritium self-sufficiency. Fusion Fission Hybrid Reactors (FFHR) can be introduced to accelerate the early application of fusion energy. Traditionally, FFHRs have been classified as either breeders or transmuters. Both need partition of plutonium from spent fuel, which will pose nuclear proliferation risks. A conceptual design of a Fusion Fission Hybrid Reactor for Energy (FFHR–E), which can make full use of natural uranium with lower nuclear proliferation risk, is presented. The fusion core parameters are similar to those of the International Thermonuclear Experimental Reactor. An alloy of natural uranium and zirconium is adopted in the fission blanket, which is cooled by light water. In order to model blanket burnup problems, a linkage code MCORGS, which couples MCNP4B and ORIGEN-S, is developed and validated through several typical benchmarks. The average blanket energy Multiplication and Tritium Breeding Ratio can be maintained at 10 and 1.15 respectively over tens of years of continuous irradiation. If simple reprocessing without separation of plutonium from uranium is adopted every few years, FFHR–E can achieve better neutronic performance. MCORGS has also been used to analyze the ultra-deep burnup model of Laser Inertial Confinement Fusion Fission Energy (LIFE) from LLNL, and a new blanket design that uses Pb instead of Be as the neutron multiplier is proposed. In addition, MCORGS has been used to simulate the fluid transmuter model of the In-Zinerater from Sandia. A brief comparison of LIFE, In-Zinerater, and FFHR–E will be given.

Neutronic investigations of a laser fusion driven lithium cooled thorium breeder

The paper investigates the main parameters of a Laser Inertial Confinement Fusion Fission Energy (LIFE) driven thorium breeder. A similar blanket to the (LIFE) engine design in Lawrence Livermore National Laboratory is chosen in order to allow mutual feedback between two geographically separated teams towards a more advanced and improved design under consideration of totally independent views. In the basic design, frozen (D,T) fusion fuel ice is shot to the center of 5 m diameter spherical fusion reactor chamber cavity in pulsed mode (10–30 Hz). Fusion fuel burns through direct or indirect laser beam irradiation. The first wall surrounds the fusion chamber and is made of S-304 steel (2 cm). The fusion reactor cavity is kept in high vacuum. It is followed by a natural lithium coolant zone. A 2nd S-304 layer (2 cm) separates the lithium zone on the right side from the graphite reflector (30 cm). The outer boundary of the graphite reflector is also covered with a 3rd S-304 layer (2 cm).The calculations have been performed for a fusion driver power of 500 MWth with the last available version of MCNP, namely with MCNPX-2.7.0. In the first calculation phase, the thickness of the natural lithium coolant-tritium breeder zone (ΔRLi) has been varied as 50, 60, 70, 80, 90 and 100 cm to select the coolant thickness ΔRLi to have a satisfactory tritium breeding ratio (TBR) for continuous fusion reactor operation. For a pure fusion blanket without any fissionable elements in the coolant, TBR values are calculated as 1.237, 1.312, 1.370, 1.415, 1.449 and 1.476, respectively, for corresponding coolant thicknesses. A ΔRLi value of 50 cm would keep TBR > 1.05 for self-sustaining tritium supply. These ΔRLi values lead to blanket energy multiplication values of M = 1.209, 1.216, 1.219, 1.222, 1.223 and 1.224, respectively, and have been calculated, as a result of exoenergetic neutron absorption in 6Li. For coolant thickness values >50 cm, the increase of “M” would remain minor.In the second phase, ThO2 has been suspended in the form of micro-size tristructural-isotropic (TRISO) particles in the lithium coolant for 233U breeding. TRISO fuel has the great advantage of high mechanical stability. Furthermore, fission products will be separated from the coolant. TRISO particles have been dispersed homogenously in the lithium coolant with volume fractions Vtr = 1, 2, 3, 4, 5 and 10 vol-%. Calculations with ΔRLi = 50 cm and by variable Vtr result with TBR = 1.229, 1.222, 1.214, 1.206, 1.1997 and 1.1622, respectively. Parasitic neutron absorption in Thorium decreases the TBR values. For Vtr < 5 vol-% TRISO in the coolant, the increase of the neutron absorption in thorium will be compensated to a great degree through neutron multiplications via 232Th(n,f) and 232Th(n,2n) reactions so that the sacrifice on TBR remains acceptable. However, for Vtr > 5 TRISO vol-%, neutron absorption in thorium reduces TBR drastically. On the other hand, the blanket energy multiplication M increases with thorium volume fraction, namely as M = 1.2206, 1.2322, 1.2426, 1.2536, 1.2636, 1.3112 for respective TRISO volume fractions due to the contribution of fission energy. Fissile fuel productions in the blanket are calculated as 17.23, 33.09, 48.66, 64.21, 79.77 and 159.71 233U (kg/year), respectively.

Burnup Analysis of Thorium-Uranium Based Molten Salt Blanket in a Fusion-Fission Hybrid Reactor

Progress on the fusion-fission hybrid reactor (FFHR) brings fusion a viable energy source in foreseeable future. Energy multiplication in a FFHR makes a much easier prerequisite for the fusion reaction than a fusion reactor. The molten salt reactor has advantages on heat transfer and post-processing of the spent fuels. A fission blanket made of molten salt was studied for the FFHR. The molten salt consists of F-Li-Be, with nuclear fuels dissolved in it. When thorium-uranium-plutonium fuels were added into a F-Li-Be molten salt zone with a component of 71% LiF -2% BeF2 -13.5% ThF4 -8.5% UF4 -5% PuF3, the appropriate blanket energy multiplication factor and TBR can be obtained. Two different molten salt models (Single molten salt zone model and multi molten salt zone model) were designed and compared in this study. The changes in blanket multiplication factor, M, and the tritium breeding ratio, TBR, during burnup life are investigated. The burnup analysis of the molten salt blanket was carried out by the COUPLE2 code. Through the burnup analysis, the breeding of the fissile fuel 233U and the transmutation of the minor actinides were also studied.

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 0.1MW/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 1725kg in the blanket for the Cases A and B, respectively.

LIFE hybrid reactor as reactor grade plutonium burner

The early version of the conceptual modified design of the Laser Inertial Confinement Fusion Fission Energy (LIFE) engine consists of a spherical fusion chamber of 5m diameter, surrounded by a multi-layered blanket. The first wall is made of 2cm thick ODS and followed by a Li17Pb83 zone (2cm), acting as neutron multiplier, tritium breeding and front coolant zone. It is separated by an ODS layer (2cm) from the FLIBE molten salt zone (50cm), containing fissionable fuel. A 3rd ODS layer (2cm) separates the molten salt zone on the right side from the graphite reflector (30cm).Calculations have been conducted for a constant fusion driver power of 500MWth in S8-P3 approximation using 238-neutron groups. Reactor grade (RG) plutonium carbide fuel in form of TRISO particles with volume fractions of 2%, 3%, 4%, 5% and 6% have been dispersed homogenously in the FLIBE coolant.Tritium breeding ratio (TBR) values per incident fusion neutron for the above cited cases start with TBR=1.35, 1.52, 1.73, 2.02 and 2.47, respectively. With the depletion of fissionable RG-Pu isotopes, TBR decreases gradually. At startup, higher fissionable fuel content in the molten salt leads to higher blanket energy multiplication, namely M0=3.8, 5.5, 7.7, 10.8 and 15.4 with 2%, 3%, 4%, 5% and 6% TRISO volume fraction, respectively. Calculations have led to very high burn up values (>400,000 MD.D/MT). TRISO particles can withstand such high burn ups. Such high burn ups would lead to drastic reduction of final nuclear waste per unit energy production.

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 40cm 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.

Fissile fuel breeding and minor actinide transmutation in the life engine

Progress on The National Ignition Facility (NIF) brings fusion a viable energy source in foreseeable future. Energy multiplication in a fusion–fission (hybrid) reactor could lead earlier market penetration of fusion energy for commercial utilization. Originally, scientists at the Lawrence Livermore National Laboratory (LLNL) have worked out a hybrid reactor design concept; the so-called Laser Inertial Confinement Fusion–Fission Energy (LIFE) engine, which has consisted of a spherical fusion chamber of ∼5 m diameter, surrounded by a multi-layered blanket with a beryllium multiplier zone after the first wall. However, earlier work had indicated extreme power peaks at immediate vicinity of the first wall of a hybrid assembly, if a beryllium multiplier is used. Hence, in the current work, the beryllium multiplier zone has been removed in order to mitigate fission power peaks at the vicinity of the first wall as a result of neutron moderation on beryllium. Furthermore, minor actinides (MA) will cause significant neutron multiplication under fusion neutron irradiation so that an extra beryllium multiplier will not be needed. Present work has made following modifications on the LLNL design of the original (LIFE) engine: • Omission of beryllium multiplier. • TRISO fuel has been suspended as micro-size particles in Flibe coolant in lieu of being dissolved in uranium salt or imbedded carbon matrix in macro-size pebbles. • Carbide fuel is used. • Fissionable fuel charge is kept lower than in the LLNL (LIFE) engine. The modified (LIFE) engine is kept similar to the LLNL design to a great degree in order to allow mutual feedback between two geographically separated teams towards a more advanced and improved design under consideration of totally independent views. The first wall is made of ODS (2 cm) and followed by a Li 17Pb 83 zone (2 cm), acting as neutron multiplier, tritium breeding and front coolant zone. It is separated by an ODS layer (2 cm) from the Flibe molten salt zone (50 cm), containing MA as fissionable fuel. A 3rd ODS layer (2 cm) separates the molten salt zone on the right side from the graphite reflector (30 cm). Calculations have been conducted for a fusion driver power of 500 MW th in S 8-P 3 approximation using 238-neutron groups. Minor actinides (MA) out of the nuclear waste of LWRs are used as fissile carbide fuel in TRISO particles with volume fractions of 0, 2, 3, 4 and 5% have been dispersed homogenously in the Flibe coolant. For these cases, tritium breeding at startup is calculated as TBR = 1.134, 1.286, 1.387, 1.52 and 1.67, respectively. In the course of plant operation, TBR and fissile neutron multiplication factor decrease gradually. For a self-sustained reactor, TBR > 1.05 can be kept for all cases over 8 years. Higher fissionable fuel content in the molten salt leads also to higher blanket energy multiplication, namely M = 3.3, 4.6, 6.15 and 8.1 with 2, 3, 4 and 5% TRISO volume fraction at start up, respectively. For all investigated cases, fissile burn up exceeds 400 000 MW D/MT. Major damage mechanisms have been calculated as DPA = 50 and He = 176 appm per year. This implies a replacement of the first wall every 3 years.

Neutronics performance evaluation of fast-fission fuel breeding blankets for a fusion–fission hybrid reactor

An investigation was conducted to evaluate the range of performance of several kinds of fast-fission natural uranium blankets adopted different fuel forms: oxide form, carbide form, nitride form, and metallic form based on plasma parameters of the FDS-FB (Fusion Driven Sub-critical System for Fuel Breeding). The objective parameters of the performance were in terms of four figures of merit (FOMs), which consist of fissile fuel breeding mass (FFBM), the blanket energy multiplication ( M), tritium breeding ratio (TBR) and the waste produced mass (WPM). The results showed that the best characteristic parameters of FFBM and M were found in the blanket loaded with metallic uranium fuel, followed by carbide fuel, nitride fuel and finally the oxide fuel in two cases of with and without considering tritium self-sustaining. Metallic uranium was found to deposit with the smallest waste inventory, and the fast-fission blanket loaded with carbide uranium or nitride uranium showed significantly higher waste produced mass.

Neutronics analysis of water-cooled energy production blanket for a fusion–fission hybrid reactor

Neutronics calculations were performed to analyse the parameters of blanket energy multiplication factor ( M) and tritium breeding ratio (TBR) in a fusion–fission hybrid reactor for energy production named FDS (Fusion-Driven hybrid System)-EM (Energy Multiplier) blanket. The most significant and main goal of the FDS-EM blanket is to achieve the energy gain of about 1 GWe with self-sustaining tritium, i.e. the M factor is expected to be ∼90. Four different fission materials were taken into account to evaluate M in subcritical blanket: (i) depleted uranium, (ii) natural uranium, (iii) enriched uranium, and (iv) Nuclear Waste (transuranic from 33 000 MWD/MTU PWR (Pressurized Water Reactor) and depleted uranium) oxide. These calculations and analyses were performed using nuclear data library HENDL (Hybrid Evaluated Nuclear Data Library) and a home-developed code VisualBUS. The results showed that the performance of the blanket loaded with Nuclear Waste was most attractive and it could be promising to effectively obtain tritium self-sufficiency and a high-energy multiplication.

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.

Neutronics Assessment of Molten Salt Breeding Blanket Design Options

Neutronics assessment has been performed for molten salt breeding blanket design options that can be utilized in fusion power plants. The concepts evaluated are a self-cooled Flinabe blanket with Be multiplier and dual-coolant blankets with He-cooled FW and structure. Three different molten salts were considered including the high melting point Flibe, a low melting point Flibe, and Flinabe. The same TBR can be achieved with a thinner self-cooled blanket compared to the dual-coolant blanket. A thicker Be zone is required in designs with Flinabe. The overall TBR will be ~1.07 based on 3-D calculations without breeding in the divertor region. Using Be yields higher blanket energy multiplication than obtainable with Pb. A modest amount of tritium is produced in the Be (~3 kg) over the blanket lifetime of ~3 FPY. Using He gas in the dual-coolant blanket results in about a factor of 2 lower blanket shielding effectiveness. We show that it is possible to ensure that the shield is a lifetime component, the vacuum vessel is reweldable, and the magnets are adequately shielded. We conclude that molten salt blankets can be designed for fusion power plants with neutronics requirements such as adequate tritium breeding and shielding being satisfied.

05/00106 Remote experiment participation on Tore-Supra: Theis, J.-M. and Larsen, J.-M. Fusion Engineering and Design, 2004, 71, (1–4), 257–261

The possibility of LWR spent fuel rejuvenation in (D, T) (Deuterium and Tritium) driven hybrid reactors having 12.5 cm of the fissile zone thickness is investigated for various plasma chamber dimensions (DR=18.7, 118.7, 218.7 and 418.7 cm) with a linear fusion neutron source (Plasma dimension is assumed as DR/2) under different first wall loads (Pw=2, 4, 6, 8 and 10 MW m−2). The behavior of the spent fuel is observed over 36 months for discrete time intervals of Δt=15 days and by a plant factor of 75%. The fissile fuel zone is considered to be cooled with three different coolants, gas (helium or CO2), Flibe (Li2BeF4) and Natural lithium. As a result of the calculation, in the case of the first wall load and the plasma chamber, dimensions are selected to be high, although the first wall material was damaged considerably by the high neutron flux (DPA>100 and He>500 ppm for Pw>2 MW m−2 over 3 years of operation) and maximum temperature in centerline of the fuel rod (Tm) has reached the melting point (Tm>2600°C for Pw>6 MW m−2 and DR>1 m), it was observed that the neutronic performance of hybrid reactor improved unnegligibly. For DR=18.7 cm, at the beginning of rejuvenation, TBR (Tritium Breeding Ratio) values being about 1.35, 1.11 and 1.39 for gas, Flibe and Natural lithium, respectively, the end of the rejuvenation increased to 1.39, 1.17 and 1.43 for 2 MW m−2 and to 1.55, 1.39 and 1.57 for 10 MW m−2. In addition, the blanket energy multiplication (M) increase 5.10, 4.87 and 4.82 for 2 MW m−2 and to 6.55, 7.04 and 6.04 for 10 MW m−2 from 4.67, 4.24 and 4.46, respectively. For only Flibe, when the preferred DR value is about 1 m, M values increased to 5.70 and to 9.5 from 4.43 for 2 MW m−2 10 MW m−2, respectively. At the same conditions, average CFFE values indicating the rejuvenation performance increased to 3.27%, 3.80% and 3.21% for 2 MW m−2, to 6.71%, 8.74% and 6.52% for 10 MW m−2 from 0.418%, respectively. For Flibe coolant, while the same CFFE value is 11.5% about DR=1 m, it is 11.9% about DR=4 m. The contribution of a large plasma chamber (DR>1 m) to neutronic performance can be neglected. The best rejuvenation performance and neutron economy has been shown by Flibe. For all cases, the denatured character of the initial fuel charge remains denatured for all investigated cases over the whole plant operation period in a hybrid reactor, although the plutonium quality increases continuously during the rejuvenation process. In addition, our calculations have proved that the effects of the important fission products (135Xe, 149Sm) and plasma densities up to 1021 (D+T)/cm3 can be neglected on the neutronic performance of hybrid reactor which is rejuvenating LWR spent fuel.

Considerations for Fusion-Based Actinide Destruction

A fusion based actinide destruction system is advantageous because of higher actinide destruction efficiency and higher energy efficiency when compared to other destruction technologies. The unique neutron multiplication capability due to the n,2n reactions in blanket materials with 14 MeV D-T neutrons enhances further the performance efficiency.Investigation of a high performance fusion based actinide destruction system was conducted. A self-cooled, actinide-carrying molten salt blanket can be designed to operate at a high sub-criticality factor of 0.95-0.96, with less than 0.4 wt% actinide concentration dissolved in the molten salt. The corresponding blanket energy multiplication is 160. Lithium-6, which is required for tritium breeding, can be used as a variable to shape the neutron spectrum and control the criticality factor, and thus to maintain a constant fission thermal power output from the actinide destruction plant.Sub-criticality can be maintained in all cases of the actinide destruction plant, during normal operation and abnormal conditions.A fusion device projected from a tokamak experiment can produce 30 MW fusion power, with a plasma amplification factor of 2. It is considered adequate to drive the sub-critical molten salt blanket. The total thermal fission power is about 4000 MW, which is able to destroy 1.6 metric tons of actinides annually when operating at full power.

Nuclear aspects of molten salt blankets

Nuclear aspects of candidate molten salts, namely a mixture of LiF and BeF 2 (FLIBE) and a mixture of LiF, NaF, and BeF 2 (FLINABE), were investigated for application as blanket coolants in tokamak fusion power plants. Tritium breeding, blanket energy multiplication, and neutron transmutation of these salts were assessed. Neutron activation of FLIBE and FLINABE was evaluated and site-boundary dose due to a worst-case loss of vacuum accident was estimated. Formation of F 2, TF and T 2 during power plant operation was analyzed and issues relevant to corrosion of structural materials due to the fluorine and fluoride species was assessed. Mechanism to control the corrosion of structural materials due to TF has been identified. Depletion of LiF, BeF 2, and NaF in the salts was calculated and quantities of the make-up fluorides to be added into the salts were estimated.

Study of fissile fuel breeding concept for the force-free helical reactor

The force-free helical reactor (FFHR) is a demo relevant helical-type D–T fusion reactor based on the large helical device (LHD). In this study, the possibilities of fissile fuel breeding (FFB) and incineration of curium oxide discharged from spent fuel of pressured water reactor (PWR) have been investigated in the FFHR. For this purpose, FFB zone containing mixed fuel has been located in the blanket of the FFHR by using two different design shapes (models 1 and 2). Four different mixtures, namely, 244CmO 2–UC, 244CmO 2–UO 2, 244CmO 2–UN and 244CmO 2–U 3Si 2, have been used as fuel. The mixed fuels have been spherically prepared, and cladded with SiC to prevent fission products from contaminating coolant and fuel–coolant contact. The prepared fuel spheres have been replaced in the FFB zone as ten rows in radial direction, and two different coolants, helium (for models 1a and 2a) and flibe (for models 1b and 2b), have been selected for the nuclear heat transfer in the FFB zone. Calculations of neutronic data per DT fusion neutron have been performed individually for each of 16 different cases created by using four different mixed fuels in each of four different models. In the models 1 and 2, the case of helium coolant has higher M (the blanket energy multiplication ratio), which is one of main parameters in a fusion–fission hybrid reactor than the case of flibe coolant. M is quite high, and varies in the ranges of 4.747–5.895 and 2.453–2.784 in the models 1 and 2, respectively, depending on the fuel and coolant types. The fissile fuel breeding ratio (FFBRs) are also quite high in the model 1a with respect to model 1b, although coolant type does not change the FFBRs substantially in the model 2. Total FFBR ( 239Pu+ 245Cm breeding ratio) varies from 0.551 to 0.650 and from 0.170 to 0.184, in the models 1 and 2, respectively, depending on the fuel and coolant types. Values of tritium breeding ratio (TBR), which is another important parameter in a fusion–fission hybrid reactor, are about 1.2 for all investigated cases so that tritium self-sufficiency is maintained for (D,T) fusion driver. Values of peak-to-average fission power density ratio, Γ are in the range of 1.108–1.249 in all investigated cases, depending on the fuel and coolant types. Furthermore, values of neutron leakage out of the blanket for all models are quite low due to B 4C shielding. 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.

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.