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Hastelloy N®, a nickel-based alloy, and 316 stainless steel are among the candidate structural materials being considered for the construction of the molten salt reactor (MSR). Most of the proposed MSR concepts use molten fluoride salts as coolant which can be quite corrosive to structural alloys. The results of studies on the corrosion behavior of the two alloys in molten Li2BeF4 (FLiBe) salt at 700°C are discussed. This salt is being considered as the primary coolant for MSR designs featuring solid fuel particles, but the reported results also provide insights into the corrosion in MSR designs where the uranium fuel is dissolved in the molten fluoride salt. Corrosion was observed to occur predominantly by de-alloying of Cr from the alloy surface and into the molten salt, with more pronounced attack occurring along the grain boundaries than in the bulk grains. The magnitude and the mechanisms of corrosion were different for corrosion tests performed in graphite and metallic capsules, a result warranting recognition given the coexistence of structural alloys and graphite in the molten salt medium in the MSR.

With the aim of developing neutron moderator and neutron reflector materials for fluoride-salt-cooled high-temperature reactors (FHRs), static infiltration tests were performed on graphite materials in molten 2LiF–BeF2 (FLiBe) salt, which is a potential primary coolant, at 700 °C and various pressures. The weight gain ratios of four grades (NBG-18, IG-110, NG-CT-10, and NG-CT-50) of graphite after infiltration were measured to determine their infiltration curves. The threshold pressure for FLiBe salt infiltration for the ultrafine-grained graphite (NG-CT-50) was greater than 600 kPa and much higher than those of the other three grades (medium-grained/fine-grained graphites), indicating that this graphite grade more probably resists salt infiltration in FHRs than other grades of graphite. However, if the threshold pressure is exceeded, it has the highest potential capacity for infiltrated salt over the pressures tested. The four grades were also characterized using mercury intrusion porosimetry. It was found that the infiltration curves of these two unwetting liquids were very similar. Scanning electron microscopy characterization showed that the FLiBe salt was distributed relatively uniformly in all four grades, indicating the presence of interconnected networks of open pores throughout the samples. X-ray diffraction patterns showed that infiltration test at high pressure led to an improved structural order and a decreased d-spacing in graphite.

The impregnation behavior of molten 2LiF–BeF2 (FLiBe) salt into a graphite matrix of fuel elements for a solid fuel thorium molten salt reactor (TMSR-SF) at pressures varying from 0.4 to 1.0 MPa was studied by mercury intrusion, molten salt impregnation, X-ray diffraction, and scanning electron microscopy techniques. It was found that the entrance pore diameter of the graphite matrix is less than 1.0 µm and the contact angle is about 135°. The threshold impregnation pressure was found to be around 0.6 MPa experimentally, consistent with the predicted value of 0.57 MPa by the Washburn equation. With the increase of pressure from 0.6 to 1.0 MPa, the average weight gain of the matrix increased from 3.05 to 10.48%, corresponding to an impregnation volume increase from 2.74 to 9.40%. The diffraction patterns of FLiBe are found in matrices with high impregnation pressures (0.8 MPa and 1.0 MPa). The FLiBe with sizes varying from tens of nanometers to a micrometer mainly occupies the open pores in the graphite matrix. The graphite matrix could inhibit the impregnation of the molten salt in the TMSR-SF with a maximum operation pressure of less than 0.5 MPa.

The 10 MWth solid-fueled thorium molten salt reactor (TMSR-SF1) is a FLiBe salt-cooled pebble bed reactor to be deployed in 5–10 years, designed by the TMSR group. Due to a large amount of beryllium in the core, the photoneutrons are produced via (γ, n) reactions. Some of them are generated a long time after the fission event and therefore are considered as delayed neutrons. In this paper, we redefine the effective delayed neutrons into two fractions: the delayed fission neutron fraction and the delayed photoneutron fraction. With some reasonable assumptions, the inner product method and the k-ratio method are adopted for studying the effective delayed photoneutron fraction. In the k-ratio method, the Monte Carlo code MCNP6 is used to evaluate the effective photoneutron fraction as the ratio between the multiplication factors with and without contribution of the delayed neutrons and photoneutrons. In the inner product method, with the Monte Carlo and deterministic codes together, we use the adjoint neutron flux as a weighting function for the neutrons and photoneutrons generated in the core. Results of the two methods agree well with each other, but the k-ratio method requires much more computing time for the same precision.

The solid-fueled thorium molten salt reactor (TMSR-SF1) is a 10 MWth test reactor design to be deployed in 5–10 years by the TMSR group. Its design combines coated particle fuel and molten FLiBe coolant for great intrinsic safety features and economic advantages. Due to a large amount of beryllium in the coolant salt, photoneutrons are produced by (γ, n) reaction, hence the increasing fraction of effective delayed neutrons in the core by the photoneutrons originating from the long-lived fission products. Some of the delayed photoneutron groups are of long lifetime, so a direct effect is resulted in the transient process and reactivity measurement. To study the impact of photoneutrons for TMSR-SF1, the effective photoneutron fraction is estimated using k-ratio method and performed by the Monte Carlo code (MCNP5) with ENDF/B-VII cross sections. Based on the coupled neutron–photon point kinetics equations, influence of the photoneutrons is analyzed. The results show that the impact of photoneutrons is not negligible in reactivity measurement. Without considering photoneutrons in on-line reactivity measurement based on inverse point kinetics can result in overestimation of the positive reactivity and underestimation of the negative reactivity. The photoneutrons also lead to more waiting time for the doubling time measurement. Since the photoneutron precursors take extremely long time to achieve equilibrium, a “steady” power operation may not directly imply a “real” criticality.

Flibe (Li2BeF4) and Flinabe (LiFNaFBeF2) are molten salt which are used as a first wall and blanket in advanced power extraction. Since they have low vapor pressure at high temperature and don’t pollute plasma. Moreover, they are used as a coolant material due to their high thermal cycling capacity. There are lots of benefits of using liquid wall such as reducing cost, increasing the reactor power and extending the life time of first steel wall. Being in Flinabe (LiFNaFBeF2) molten salt, 23 Na is also considered as a coolant material in fast reactors. Therefore, the reaction including neutron and 23 Na is getting important. In this study, the cross section calculations of neutron induced reactions for 23 Na were performed. Using pre-equilibrium and equilibrium models, excitation functions of 23 Na(n,p)23 Ne, 23 Na(n,d)22 Ne, 23 Na(n,α)20 F, 23 Na(n,2n)22 Na reactions and 23 Na(n,xα), 23 Na(n,xn) emission spectrums were calculated. We calculated alpha emission spectrum obtained when 23 Na nuclei projectiled with 14.1 MeV neutrons and also neutron emission spectrum emitted when 23 Na nuclei was bombardment with 14.6 MeV neutrons.

In this study, some important thermodynamic properties of the fusion reactor have been analyzed. The physical and chemical properties of molten salts have been extensively studied in the nuclear fusion program. In recent years, molten salts technology began to be used in some engineering areas, in the advanced nuclear field and especially in nuclear fusion reactor systems. Nowadays, Aries team has developed advanced designs by using the molten salts technology in order to get high thermodynamic and structural advantage on nuclear technology areas (Tillack et al. in Fusion Energ Des 65:215–261, 2003; Tillack et al. in Fusion Energ Des 49–50:689–695, 2000; El-Guebaly et al. in Fusion Energ Des 65:263–284, 2003). The Aries-St reactors are a 1000 MW fusion reactor system that based on a low aspect ratio ST plasma (Tillack et al. in Fusion Energy Des 65:215–261, 2003; Tillack et al. in Fusion Energy Des 49–50:689–695, 2000). The Aries team studies especially on liquid walls concepts and this liquid are used to increase neutronic performance of various structures of Aries-St reactors. In this research, candidate molten salts have been studied neutron effects on reactor performance which are the first wall (FW) and blanket. There are various candidate liquids that meet all the criteria such as Li17Pb83, flibe(Li2BeF4) and LiNaBeF4, LiSn that are able to breed enough tritium. In this research, we used Li17Pb83, pure lithium and flibe as candidates that are in the Aries design. Montecarlo n-particle 4b-code is used for neutronics analysis and thermodynamic features. The value of tritium breeding ratio of the Aries-St reactors must be (TBR ≥ 1.1). This can be achieved in the region of LiPb/FW blanket of reactors. Aries-St spherical reactor has high heat flux (0.8 MW/m2) and NWL (6–8 MW/m2) in this region.

In this study, we developed new models for liquid wall (FW) state at ARIES-ST fusion reactor systems. ARIES-ST is a 1,000 MWe fusion reactor system based on a low aspect ratio ST plasma. In this article, we analyzed the characteristic properties of magnetohydrodynamics (MHD) and heat transfer conditions by using Monte-Carlo simulation methods (ARIES Team et al. in Fusion Eng Des 49–50:689–695, 2000; Tillack et al. in Fusion Eng Des 65:215–261, 2003) . In fusion applications, liquid metals are traditionally considered to be the best working fluids. The working liquid must be a lithium-containing medium in order to provide adequate tritium that the plasma is self-sustained and that the fusion is a renewable energy source. As for Flibe free surface flows, the MHD effects caused by interaction with the mean flow is negligible, while a fairly uniform flow of thick can be maintained throughout the reactor based on 3-D MHD calculations. In this study, neutronic parameters, that is to say, energy multiplication factor radiation, heat flux and fissile fuel breeding were researched for fusion reactor with various thorium and uranium molten salts. Sufficient tritium amount is needed for the reactor to work itself. In the tritium breeding ratio (TBR) >1.05 ARIES-ST fusion model TBR is >1.1 so that tritium self-sufficiency is maintained for DT fusion systems (Starke et al. in Fusion Energ Des 84:1794–1798, 2009; Najmabadi et al. in Fusion Energ Des 80:3–23, 2006).

Apex hybrid reactor has a good potential to utilize uranium and thorium fuels in the future. This toroidal reactor is a type of system that facilitates the occurrence of the nuclear fusion and fission events together. The most important feature of hybrid reactor is that the first wall surrounding the plasma is liquid. The advantages of utilizing a liquid wall are high power density capacity good power transformation productivity, the magnitude of the reactor’s operational duration, low failure percentage, short maintenance time and the inclusion of the system’s simple technology and material. The analysis has been made using the MCNP Monte Carlo code and ENDF/B–V–VI nuclear data. Around the fusion chamber, molten salts Flibe (LI2BeF4), lead–lithium (PbLi), Li–Sn, thin-lityum (Li20Sn80) have used as cooling materials. APEX reactor has modeled in the torus form by adding nuclear materials of low significance in the specified percentages between 0 and 12 % to the molten salts. In this study, the neutronic performance of the APEX fusion reactor using various molten salts has been investigated. The nuclear parameters of Apex reactor has been searched for Flibe (LI2BeF4) and Li–Sn, for blanket layers. In case of usage of the Flibe (LI2BeF4), PbLi, and thin-lityum (Li20Sn80) salt solutions at APEX toroidal reactors, fissile material production per source neutron, tritium production speed, total fission rate, energy reproduction factor has been calculated, the results obtained for both salt solutions are compared.

Fertile fuel, such as thorium or depleted uranium, can be bred into fissile fuel and burnt in a breed-and-burn (B&B) reactor. Modeling a full core with fertile fuel can assess the performance of a B&B reactor with exact quantitative estimates, but costs too much computation time. For simplicity, performing the recently developed neutron balance method with a zero-dimensional (0-D) model can also provide a reasonable result. Based on the 0-D model, the feasibility of the B&B mode for thorium fuel in a fast reactor cooled by sodium was investigated by considering the (n, 2n) and (n, 3n) reaction rates of fuel and coolant in this work, and compared with that of depleted uranium fuel. Afterward, the performance of the same thorium-based fuel core, but cooled by helium, lead-bismuth, and FLiBe, respectively, is discussed. It is found that the (n, 2n) and (n, 3n) reactions should not be neglected for the neutron balance calculation for thorium-based fuel to sustain the B&B mode of operation.

The fluoride salt-cooled high-temperature reactor (FHR) uses molten FLiBe salt as the coolant, which introduces a corrosive effect on the alloy-N structure material. Fission neutrons activate the corroded alloy-N, along with alloy-N structures inside the reactor vessel. The activation products of the alloy-N have a big impact on radiation protection during operation, maintenance, and decommissioning of the reactor. We have constructed a SCALE 6.1 model for the core of a typical 10 MWth FHR and analyzed the activity of each constituent of the irradiated alloy-N. The results show that the activity is predominantly due to short-lived 28Al, 60mCo, 56Mn, 51Ti, and 52V, as well as long-lived 60Co, 51Cr, 55Fe, 59Fe, and 54Mn. Furthermore, because of their relatively long half-life and high-energy γ-rays emissions, 60Co and 54Mn are the major contributors to the radiation source terms introduced by alloy-N activation. The yield of 60Co and 54Mn per unit mass of alloy-N under the current core design is 5.58 × 105 and 1.55 × 103 Bq MWd−1 g−1, respectively. The results of this paper, combined with future corrosion studies, may provide a basis for evaluating long-term radiation source terms of the primary loop salt and components.

Certain light nuclei such as Lithium (Li), Beryllium (Be), Fluorine (F) (which are known as FLIBE) and its molten salt compounds (LiF, BeF2 and NaF) can serve as a coolant which can be used at high temperatures without reaching a high vapor pressure. These molten salt compounds are also a good neutron moderator. In this study, cross sections of neutron induced reactions have been calculated for fluorine target nucleus. The new calculations on the excitation functions of 19F(n,2n), 19F(n,p), 19F(n,xn), 19F(n,xp) have been made. In these calculations, the pre-equilibrium and equilibrium effects have been investigated. The pre-equilibrium calculations involve the full exciton model and the cascade exciton model. The equilibrium effects are calculated according to the Weisskopf–Ewing model. Also in the present work, the (n,2n) and (n,p) reaction cross sections have calculated by using evaluated empirical formulas developed by Tel et al. at 14–15 MeV energy. The multiple pre-equilibrium mean free path constant from internal transition have been investigated for 19F nucleus. The obtained results have been discussed and compared with the available experimental data.

In this study, the improvement of neutronic performance of a dual purpose modified PACER concept has been investigated. Flibe as the main constituent are fixed as 92% coolant. ThF4 is mixed with increased mole-fractions of RG-PuF4 starting by 0 mol % up to 1 mol %. TBR variations for all the investigated salts with respect to the RG-PuF4 contents are computed. Tritium self-sufficiency is provided with the ThF4 when the adding RG-PuF4 content is higher than 0.75%. The energy multiplication of the blanket is increased as 70% with adding RG-PuF4 contents to ThF4. High quality fissile isotope U-233 are produced with increasing RG-PuF4. DPA and helium production increases with increased RG-PuF4 content in molten salt. Radiation damage with dpa < 1.7 and He < 3.3 ppm after a plant operation period of 30 years will be well below the damage limit values.

HTRs use a high performance particulate TRISO fuel with ceramic multi-layer coatings due to the high burn up capability and very neutronic performance. TRISO fuel because of capable of high burn up and very neutronic performance is conducted in a D-T fusion driven hybrid reactor. In this study, TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The neutronic effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on the fuel performance has been investigated for Flibe, Flinabe and Li(20)Sn(80) coolants. The reactor operation time with the different first neutron wall loads is 24 months. Neutron transport calculations are evaluated by using XSDRNPM/SCALE 5 codes with 238 group cross section library. The effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on tritium breeding (TBR), energy multiplication (M), fissile fuel breeding, average burn up values are comparatively investigated. It is shown that the high burn up can be achieved with TRISO fuel in the hybrid reactor.

Utilization of nuclear explosives can produce a significant amount of energy which can be converted into electricity via a nuclear fusion power plant. An important fusion reactor concept using peaceful nuclear explosives is called as PACER which has an underground containment vessel to handle the nuclear explosives safely. In this reactor, Flibe has been considered as a working coolant for both tritium breeding and heat transferring. However, the rich neutron source supplied from the peaceful nuclear explosives can be used also for fissile fuel production. In this study, the effect of using thorium molten salts on the neutronic performance of the PACER was investigated. The computations were performed for various coolants bearing thorium and/or uranium-233 with respect to the molten salt zone thickness in the blanket. Results pointed out that an increase in the fissile content of the salt increased the neutronic performance of the reactor remarkably. In addition, higher energy production was obtained with thorium molten salts compared to the pure mode of the reactor. Moreover, a large quantity of (233)U was produced in the blanket in all cases.

ARIES-RS is one of the major magnetic fusion energy reactor designs that uses a blanket having vanadium alloy structure cooled by lithium [1, 2]. It is a deuterium–tritium (DT) fusion driven reactor, having a fusion power of 2170 MW [1, 2]. This study presents the neutronic analysis of the ARIES-RS fusion reactor using heavy metal molten salts in which Li2BeF4 as the main constituent was mixed with increased mole fractions of heavy metal salt (ThF4 or UF4) starting by 2 mol.% up to 12 mol.%. Neutron transport calculations were carried out with the help of the SCALE 4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S 8–P 3 approximation. According to the numerical results, tritium self-sufficiency was attained for the coolants, Flibe with 2% UF4 or ThF4 and 4% UF4. In addition, higher energy multiplication values were found for the salt with UF4 compared to that with ThF4. Furthermore, significant amount of high quality nuclear fuel was produced to be used in external reactors.

In design of deuterium-tritium (DT) fusion driven hybrid reactors, maintaining tritium self-sufficiency is very important to decrease operational costs. A (DT) fusion driven hybrid reactor must produce its own tritium to annihilate external investment on tritium production. Tritium Breeding Ratio (TBR) should be greater than 1.05 to maintain tritium self-sufficiency of the (DT) fusion plasma. Tritium is an artificial isotope that can be bred by using the breeding reactions of 6Li and 7Li isotopes. Therefore use and selection of lithium bearing coolants and blanket materials are very important with respect to tritium self-sufficiency of the (DT) fusion plasma. In this study, tritium breeding potential of various solid breeders Li2TiO3, Li8ZrO6, Li4SiO4, Li2ZrO3 or Li2O with different coolants Li2BeF4 (Flibe), NaF.LiF.BeF2 (Flinabe) or natural lithium in the hybrid blanket using different nuclear carbide fuels was investigated. Computations were carried out by using Scale4.3 solving Boltzmann transport equation. According to numerical results, among the solid breeders Li2O and Li8ZrO6 and among the liquid breeders, natural lithium had the highest contribution to TBR. Furthermore addition of uranium to thorium fuel improved tritium production.

In this study, neutronic analysis of the HYLIFE-II reactor was investigated by inserting fuel rods containing UC or mixed ThC–UC into reflector zone partially. Four different coolants, namely, flibe, helium, natural lithium, and light water were considered in the fissile fuel breeding zone for comparison. Neutron transport calculations per incident (D,T) fusion neutron were performed by using the code Scale 4.3 under resonance-effect and resonance-free conditions. Numerical results pointed out that replacing the reflector zone by fissile fuel breeding zone even with a thickness of 14 cm improved the neutronic performance remarkably with respect to energy amplification and fissile fuel breeding.

Using liquid wall between the plasma and solid first wall in a fusion reactor allows to use high neutron wall loads and could eliminate frequent replacement of the first wall structure during reactor’s lifetime. Liquid wall should have a certain effective or optimum thickness to extend solid first wall lifetime to reactor’s lifetime and supply sufficient tritium for deuterium–tritium (DT) fusion driver. This study presents the effect of thickness of flowing liquid wall containing 90 mol % Flibe+10 mol % UF4 or ThF4 on the neutronic performance of a magnetic fusion reactor design called APEX. Neutron transport calculations were carried out with the aid of code Scale4.3. Numerical results brought out that optimum liquid wall thickness of ∼38 cm was found for the blankets using Flibe+10% UF4 whereas, 56 cm for that with Flibe+10% ThF4. Significant amount of high quality fissile fuel was produced by using heavy metal salt.

One of the major inertial fusion energy reactor designs is HYLIFE-II which uses protective flowing liquid wall between fusion plasma and solid first wall. The most attractive aspect of this reactor is that protective liquid wall eliminates the frequent replacement of the first wall structure during reactor lifetime. Liquid wall thickness must be at least the thickness required for supplying sufficient tritium for the deuterium–tritium (DT) driver and satisfying radiation damage on the first wall below the limits. Reducing this thickness results less pumping power requirements and cost of electricity. In this study, investigation on potential of utilizing refractory alloys (W-5Re, TZM and Nb-1Zr) as first wall to reduce effective liquid wall thickness in HYLIFE-II reactor using liquid wall of Flibe + 10 mol % UF4 mixture. Neutron transport calculations were carried out 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. Numerical results showed that using W-5Re or TZM as first wall was effective in decreasing liquid wall thickness in contrast to Nb-1Zr.

In a commercial (DT) driven fusion reactor, the tritium breeding ratio per incident fusion neutron must be greater than 1.05 to maintain tritium self-sufficiency for the driver. In this study tritium breeding capability of three different coolants, namely Flibe (LiF.BeF2), Flinabe (LiF.NaF.BeF2), and Li20Sn80 in a (DT) driven fusion-fission (hybrid) reactor was investigated for different refractory alloys (W-5Re, TZM, T111, and Nb-1Zr) as structural material. Neutron transport calculations were conducted with the help of SCALE 4.3 SYSTEM by solving the Boltzmann transport equation with code XSDRNPM. The contribution of Flibe, Flinabe, and Li20Sn80 with respect to Li-6 enrichment in their lithium content to overall TBR was investigated. In addition, the effect of structural material type on TBR was examined.

The United States Department of Energy (DOE) sponsored two studies of inertial fusion energy (IFE) electric power plants, one led by McDonnell Douglas Aerospace (MDA) and the other led by W. J. Schafer Associates (WJSA). These studies, which were completed in March of 1992, included the conceptual design and analysis of two KrF-laser and two heavy-ion-driven power plants. The reactor and driver designs for the heavy-ion power plants are described in this paper. The MDA team design, Prometheus-H, features a wetted-wall chamber, a compact induction linac driver, and channel propagation for beam delivery. The single-beam driver uses storage rings to accumulate a series of pulses before delivery to the target. The use of storage rings and a folded linear section results in a compact footprint and significant cost savings compared to a straight multiple-beam linac. Fourteen beams of doubly-charged, 4 GeV Pb ions are used to deliver 7 MJ on the target at a pulse repetition rate of 3.5 Hz with an overall driver efficiency of 21%. The channel propagation mode requires only two small, diametrically opposed holes through the blanket and first wall; this approach maximizes the wall coverage and minimizes radiation leakage and contamination of the beamlines. Th resulting target gain is 103 for a yield of 719 MJ per pulse. The Prometheus chamber has a Pb-cooled, porous SiC first wall and a He-cooled Li2O breeding blanket. The blanket structural material is low-activation SiC. The thermal conversion efficiency is 43%, and the net electric power is 1000 MWe. The WJSA team design, Osiris, uses a wetted-wall chamber, a multiple-beam induction linac driver, and ballistic focus with auto-neutralization. The driver design is conservative in that it uses singly charged ions and does not use beam combination or splitting. Twelve beams are close-packed in order to use common cores and focusing magnets. The accelerator costs were reduced by operating at spacecharge-limited current density and standardizing the quadrupole and inductor designs. In this case, twelve beams of 3.8 GeV Xe ions provide a total of 5 MJ at a pulse repetition rate of 4.6 Hz. The overall driver efficiency is 28%, the target gain is 87, and the yield is 432 MJ per pulse. The Osiris chamber uses a flexible, porous carbon/carbon fabric to contain the flow of the molten salt Flibe, which serves as the primary coolant and breeding material. The vacuum vessel structure is a carbon/carbon composite. The thermal conversion efficiency is 45%, and the net electric power is 1000 MWe.

The modeling of APEX hybrid reactor, produced by using ARIES-RS hybrid reactor technology, has been performed by using the MCNP-4B computer code and ENDF/B-V-VI nuclear data. Around the fusion chamber, molten salts Flibe (Li2BeF4) and Flinabe (LiNaBeF4) were used as cooling materials. APEX reactor was modeled in the torus form by adding nuclear materials of low significance in the specified percentages between percent 0–12 to the molten salts. The result of the study indicated that fissile material production, UF4 and ThF4 heavy metal salt increased nearly at the same percentage and it was observed that the percentage of it was practically the same in both materials. In order for the hybrid reactor to work itself in terms of tritium, TBR (tritium breeding ratio) should be lower than 1.05. When flibe molten salt was utilized in the APEX hybrid reactor, TBR was calculated as >1, 22 and when flinabe molten salt was used, TBR was calculated as >1.06.

Neutronic calculations were performed to optimize the SENRI blanket in terms of energy multiplication as well as tritium breeding ratio. The blanket employs a thick (∼ 64-cm) Li layer as breeder/coolant. Three approaches were taken here to achieve the goal: (1) reduction of6Li in the lithium, (ii) replacement of the Li layer by a molten-salt (flibe) layer, and (iii) shipment of excess tritium to a nonbreeding blanket. It was found that the excess tritium produced in the SENRI blanket could be used effectively to obtain additional power by fueling a nonbreeding D-T reactor.

The objective of this study is to provide a comparison of thermal-hydraulic and structural performance of lithium, helium, and flibe cooled fusion blankets based on a tube/header geometry in a liquid lithium breeder. Type 316 stainless steel and TZM are considered as representative near-term and long-term, high temperature blanket structural materials, respectively, to show the potentials of each coolant. The flibe-TZM system has the best characteristics, while lithium-316SS, helium-316SS, and helium-TZM are comparable but definitely more limited in operating conditions. These results suggest that molten salt-refractory metal systems deserve more attention.