Utilization of TRISO Fuel with LWR Spent Fuel in Fusion-Fission Hybrid Reactor System

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 Li₂₀Sn₈₀ 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.     

RETRACTED ARTICLE: The Evaluation of Reactor Performance by using Flibe and Flinabe Molten Salts in the APEX Hybrid Reactor

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.     

Advanced Power Conversion Efficiency in Inventive Plasma for Hybrid Toroidal Reactor

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.     

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.     

Effect of Nuclear Data Libraries on Tritium Breeding in a (D–T) Fusion Driven Reactor

In design a Deuterium–Tritium (D–T) fusion driven hybrid reactor, neutronics and nuclear data libraries have an essential role for reliable neutronics calculations. Therefore, nuclear data libraries are very important to calculate of the neutronic parameters and selection of tritium breeder materials to be used in the blanket. In this study tritium breeding performances of candidate tritium breeding materials, namely, Li₂O, LiH, Li₂TiO₃, Li₂ZrO₃ and Li₄SiO₄ in a (D–T) driven fusion–fission (hybrid) reactor is investigated based on three dimensional (3-D) and one dimensional (1-D) neutronic calculations. 3-D and 1-D neutron transport calculations are performed with Monte Carlo transport code (MCNP 4C), SCALE 5 and ANISN nuclear data codes to determine the tritium breeding ratio (TBR) of the blanket. The effects of different nuclear data libraries on TBR are examined and TBR calculation results are comparatively investigated.     

Utilization of Heavy Metal Molten Salts in the ARIES-RS Fusion Reactor

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.     

Monte Carlo calculations of the nuclear effects of certain fluids in a hybrid reactor

This study analyzes the effects of certain heavy-metal-salt fluids on nuclear parameters in a fusion–fission hybrid reactor. Calculated parameters include the tritium breeding ratio (TBR), energy multiplication factor (M), heat deposition rate, fission reaction rate, and fissile fuel breeding in the reactor's liquid first wall, blanket, and shield zones; gas production rates in the structural material of the reactor were calculated, as well. The fluid mixtures consisted of 93–85% Li₂₀Sn₈₀ + 5% SFG-PuO₂ and 2–10% UO₂, 93–85% Li₂₀Sn₈₀ + 5% SFG-PuO₂ and 2–10% NpO₂, and 93–85% Li₂₀Sn₈₀ + 5% SFG-PuO₂ and 2–10% UCO. The fluids were used in the liquid first wall, blanket, and shield zones of a fusion–fission hybrid reactor system. A 3 cm wide beryllium (Be) zone was used for neutron multiplier between the liquid first wall and the blanket. The structural material used was 9Cr2WVTa ferritic steel, measuring 4 cm in width. Three-dimensional analyses were performed using the Monte Carlo code MCNPX-2.7.0 and the ENDF/B-VII.0 nuclear data library.     

On the radiation damage characterization of candidate first wall materials in a fusion reactor using various molten salts

Evaluating radiation damage characteristics of structural materials considered to be used in fusion reactors is very crucial. In fusion reactors, the highest material damage occurs in the first wall because it will be exposed to the highest neutron, gamma ray and charged particle currents produced in the fusion chamber. This damage reduces the lifetime of the first wall material and leads to frequent replacement of this material during the reactor operation period. In order to decrease operational cost of a fusion reactor, lifetime of the first wall material should be extended to reactor’s lifetime. Using a protective flowing liquid wall between the plasma and first wall can decrease the radiation damage on first wall and extend its lifetime to the reactor’s lifetime. In this study, radiation damage characterization of various low activation materials used as first wall material in a magnetic fusion reactor blanket using a liquid wall was made. Various coolants (Flibe, Flibe + 4% mol ThF₄, Flibe + 8% mol ThF₄, Li₂₀Sn₈₀) were used to investigate their effect on the radiation damage of first wall materials. Calculations were carried out by using the code Scale4.3 to solve Boltzmann neutron transport equation. Numerical results brought out that the ferritic steel with Flibe based coolants showed the best performance with respect to radiation damage.     

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.     

3D Neutronic Analysis in MHD Calculations at ARIES-ST Fusion Reactors Systems

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).     

Estimation of Neutronic Performance in a Hybrid Reactor with Regression Analysis

This study presents regression analysis method used for prediction and investigation of neutronic performance in a hybrid reactor using UO₂ fuel and Flibe (Li₂BeF₄) coolant. The ²³⁵U fraction is increased gradually from 0 to 4% stepped by 1% and the ⁶Li fraction within the Flibe coolant is enriched gradually to 30, 60 and 90% from 7.5%. Relations between ²³⁵U fuel fraction and lithium (⁶Li) enrichment are investigated for the estimation of neutronic performance as the tritium breeding ratio (TBR), energy multiplication factor (M), total fission rate (Σ_f), ²³⁸U (n,γ) reaction and fissile fuel breeding (FFB) in the hybrid reactor. Regression analysis by results obtained by using the code (XSDRNPM/SCALE5) for TBR, M, Σ_f, ²³⁸U (n,γ) and FFB are performed. The results of the regression analysis and the values obtained by using the code (XSDRNPM/SCALE5) are compared with respect to the TBR, M, Σ_f, ²³⁸U (n,γ) and FFB of the reactor. The values calculated from the obtained formulations with regression analysis are found to be in good agreement with results obtained by using the code (XSDRNPM/SCALE5). It is observed that the derived equations from regression analysis could provide an accurate computation of the neutronic performances so that these equations could use for the prediction of TBR, M, Σ_f, ²³⁸U (n,γ) and FFB. In addition, correlation matrix is calculated to determine the degree of relationship between variables as TBR, M, Σ_f, ²³⁸U (n,γ) and FFB.     

氦气、水和熔盐冷却的混合堆中子学性能研究

氦气、水、熔盐(Flibe)在强磁场中流动不存在严重的MHD问题,因此适合在基于磁约束的聚变-裂变混合堆中作为冷却剂.针对氦气、水、Flibe这3种冷却剂对混合堆包层中子学性能的影响进行研究,分析包层中能谱特点及燃料增殖特性.通过燃耗计算,研究氚增殖率(TBR)、能量倍增因子(M)、keff等随运行时间的变化.中子学输运采用三维蒙特卡罗程序MCNP.计算结果表明,不同的冷却剂对混合堆系统中子能谱影响很大:氦冷系统的能谱最硬,主要发生快中子裂变,氚增殖效果最好;水冷系统的能谱最软,产能最多,但需提高TBR;Flibe冷系统的能谱较硬,产能最少.     

Investigation on the Neutronic Performance of a Fusion Reactor Using Flibe with Heavy Metal Fluorides

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.     

Neutronic performance of HYLIFE-II fusion reactor using various thorium molten salts

HYLIFE-II is one of the major inertial fusion energy reactor design concepts in which a thick molten salt layer (Flibe = Li₂BeF₄) is injected between the reaction chamber walls and the explosions. Molten salt coolant eliminates the frequent replacement of solid first wall structure during reactors lifetime by decreasing intense neutron flux. This study presents the neutronic analysis of HYLIFE-II fusion reactor using various liquid wall coolants, namely, 75% LiF–25% ThF₄, 75% LiF–24% ThF₄–1% ²³³UF₄ or 75% LiF–23% ThF₄–2% ²³³UF₄. Neutron transport calculations for the evaluation of neutron spectra were conducted with the help of Scale 4.3 by solving the Boltzmann transport equation in S₈–P₃ approximation. The effects of flowing liquid wall thickness and type of coolant on the neutronic performance of the reactor were investigated. Furthermore, radiation damage calculations at the first wall structure with respect to type and thickness of liquid wall were carried out. Numerical results showed that using the flowing liquid wall containing the molten salt, 75% LiF–23% ThF₄–2% UF₄ with a thickness of 70 cm maintained tritium self-sufficiency of the (DT) fusion driver and extended the first wall lifetime to the reactors lifetime (30 full power years). In addition significant amount of high quality fissile fuel was bred through (n, γ) reaction of ²³²Th. Moreover, energy multiplication factor (M) was increased to 12 by high rate fission reactions of ²³³U occurring in the flowing wall. On the other hand, it was concluded that using the other two coolants, 75% LiF–25% ThF₄ or 75% LiF–24% ThF₄–1% ²³³UF₄, as liquid wall did not satisfy the radiation damage and the tritium sufficiency criteria together at any thickness, so that these two coolants were not suitable to improve neutronic performance of HYLIFE-II reactor.     

Impact of Solid Breeder Materials on Tritium Breeding in a Hybrid Reactor

Tritium breeding ratio (TBR) is one of the important parameters in design of a Deuterium–Tritium (DT) driven hybrid reactor. Therefore, selection of tritium breeder materials to be used in the blanket is very crucial. In this study, tritium breeding potential of the solid breeders, namely, or in a (DT) fusion driven hybrid reactor fuelled with or was investigated. For this purpose in addition to these solid breeders, different types of liquid breeders, namely natural lithium, Flibe, Flinabe and were used to examine the tritium breeding behavior of liquid–solid breeder couple combinations. Numerical calculations were carried out by using Scale 4.3. According to numerical results, the blanket with fuel using natural lithium as coolant and as solid breeder had the highest TBR value.     

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.     

Three-dimensional neutronic calculations for a fusion breeder APEX reactor using some libraries

The effects of evaluated nuclear data files on neutronics characteristics of a fusion–fission hybrid reactor have been analyzed; three-dimensional calculations have been made using the MCNP4C Monte Carlo Code for ENDF/B-VII T = 300 K, JEFF-3.0 T = 300 K, and CENDL-2 T = 300 K evaluated nuclear data files. The nuclear parameters of a fusion–fission hybrid reactor such as tritium breeding ratio, energy multiplication factor, fissile fuel breeding and nuclear heating in a first wall, blanket and shield have been investigated for the mixture components of 90% Flibe (Li₂BeF₄) and 10% UF₄ for a blanket layer thickness of 50 cm. The contributions of each isotope of Flibe (⁶Li, ⁷Li, ¹⁹F, ⁹Be) and UF₄ (²³⁵U, ²³⁸U) to the integrated parameter values were calculated. The neutron wall load is assumed to be 10 MW/m².     

Effect of water adsorption on tritium release from Li4SiO4

As a ceramic material proposed for tritium breeding in a fusion reactor blanket, lithium orthosilicate (Li₄SiO₄) is being examined in view of the influence of water uptake on tritium release behavior. In this work, out-of-pile tritium release experiments were performed on Li₄SiO₄ samples that were transferred and stored under different moisture environments. The water content was measured on the samples that were treated in similar conditions. Effects of water adsorption on the chemical form and temperature of released tritium were investigated. It is found that with the water content increases, the gaseous tritium fraction decreases and the proportion of low-temperature desorption of HTO increases. The results of this study can be used later for engineering and design activities for fusion reactor blankets.     

Reducing Effective Liquid Wall Thickness in a HYLIFE-II Fusion Breeder

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.     

Neutronic analysis of a thorium fusion breeder with enhanced protection against nuclear weapon proliferation

The fissile breeding capability of a (D,T) fusion-fission (hybrid) reactor fueled with thorium is analyzed to provide nuclear fuel for light water reactors (LWRs). Three different fertile material compositions are investigated for fissile fuel breeding: (1) ThO₂; (2) ThO₂ denaturated with 10% natural-UO₂ and (3) ThO₂ denaturated with 10% LWR spent fuel. Two different coolants (pressurized helium and Flibe ‘Li₂BeF₄’) are selected for the nuclear heat transfer out of the fissile fuel breeding zone. Depending on the type of the coolant in the fission zone, fusion power plant operation periods between 30 and 48 months are evaluated to achieve a fissile fuel enrichment quality between 3 and 4%, under a first-wall fusion neutron energy load of 5 MW/m² and a plant factor of 75%. Flibe coolant is superior to helium with regard to fissile fuel breeding. During a plant operation over four years, enrichment grades between 3.0 and 5.8% are calculated for different fertile fuel and coolant compositions. Fusion breeder with ThO₂ produces weapon grade ²³³U. The denaturation of the ²³³U fuel is realized with a homogenous mixture of 90% ThO₂ with 10% natural-UO₂ as well as with 10% LWR spent nuclear fuel. The homogenous mixture of 90% ThO₂ with 10% natural-UO₂ can successfully denaturate ²³³U with ²³⁸U. The uranium component of the mixture remains denaturated over the entire plant operation period of 48 months. However, at the early stages of plant operation, the generated plutonium component is of weapon grade quality. The plutonium component can be denaturated after a plant operation period of 24 and 30 months in Flibe cooled and helium cooled blankets, respectively. On the other hand, the homogenous mixture of 90% ThO₂ with 10% LWR spent nuclear fuel remains non-prolific over the entire period for both, uranium and plutonium components. This is an important factor with regard to international safeguarding.