High Breeding Ratio Research Articles

Pre-conceptual design and proof of principle assessments of self-cooled Toroidally symmetric lead-lithium (TSLL) blanket

A new self-cooled liquid metal blanket concept called TSLL (Toroidally Symmetric Lead-Lithium) blanket is proposed and assessed, including analysis for magnetohydrodynamic (MHD) flows, structural analysis, and heat transfer and neutronics assessments using the ARC reactor with demountable magnets designed by the Commonwealth Fusion Systems (CFS) as a testbed. The proposed blanket utilizes lead-lithium (PbLi) alloy as breeder/coolant and reduced activation ferritic/martensitic (RAFM) steel as structural material. A special feature of the new concept is the toroidally symmetric flow in the blanket integrated first wall and the breeding zone to reduce the MHD pressure drop, while using anchor links to strengthen the first wall construction. Provided analysis suggests acceptable MHD pressure drop, required mechanical integrity and high tritium breeding ratio of ∼ 1.64. As a result of these assessments, the new blanket concept can be recommended for more detailed studies as a promising blanket candidate for implementation in future fusion devices.

Whole core analysis of the single-fluid double-zone heavy water-moderated molten salt reactor (SD-HWMSR)

Thermal-spectrum molten salt reactor (MSR) concepts usually adopt graphite as a neutron moderator. Although graphite as a moderator has many advantages, it also has drawbacks, including a relatively short lifespan (it has to be replaced), positive temperature feedback, and loss of impermeability due to expansion. Replacing graphite with heavy water in MSRs can effectively solve the problems introduced by graphite. The SD-HWMSR is a Single-fluid Double-zone Heavy Water-moderated Molten Salt Reactor. Optimization of SD-HWMSR’s fuel channel pitch and radius has been conducted in this work. As a result, the SD-HWMSR with a high breeding ratio (1.07465 ± 0.00060), low initial 233U loading (1.43 t), and negative temperature and void reactivity coefficients is put forward. The SERPENT-2 is used to analyze the neutronics parameters of the reactor design. The current work investigates the change in the multiplication factor, breeding ratio, and the accumulation of significant nuclides in the core. The suitable 232Th and 233U refill rates needed to maintain criticality and enable analysis of the whole core of SD-HMMSR are thoroughly determined in this study. Uranium-233 and thorium-232 both have a maximum refill rate of 3.49 and 3.20 kg/d, respectively. While the average refill rate for 233U and 232Th during the 60 years of operation is 1.90 and 2.35 kg/d, respectively. The net production of 233U rises with time, and by the end of the 60 years, it is around 1.98 t. The doubling time for the SD-HWMSR is 31 years, which is consistent with previous results.

Tritium Transport Modeling at the Pore Scale in Ceramic Breeder Materials Using TMAP8

Fusion reactors depend on the blanket material to breed and release tritium at the same rate or faster than it is consumed by the fusion reaction. Cellular ceramic breeders (CCBs) are dense materials that can maintain a high tritium breeding ratio while promoting tritium release because of highly connected pores. Assessing the tritium breeding capabilities of these materials requires a combination of extensive experimental and modeling efforts. In this work, we develop and calibrate a multiphysics model of tritium transport. This novel model accounts for ceramic and pore diffusion, trapping and detrapping, and several surface reactions at the pore surface. We perform a sensitivity analysis and calibrate the model by comparing its predictions against experimental measurements of deuterium absorption. The calibrated model is then used to model tritium absorption in samples with different pore microstructures to investigate the effect of pore interconnectivity on tritium absorption. The model is part of the development of the multiscale, multiphysics framework for tritium transport [i.e., the Tritium Migration Analysis Program (TMAP8)], which is itself built on top of the finite-element multiphysics framework multiphysics object-oriented simulation environment (MOOSE). This study demonstrates some of TMAP8’s capabilities and is the first step toward assessing the tritium breeding capabilities of ceramic breeder material designs.

The LIBRA Experiment: Investigating Robust Tritium Accountancy in Molten FLiBe Exposed to a D-T Fusion Neutron Spectrum

The Liquid Immersion Blanket: Robust Accountancy (LIBRA) experiment will be a first-of-a-kind experiment to explore and develop the liquid immersion blanket (LIB) concept. The LIB is a radically simple molten–LiF-BeF2 (FLiBe)–salt tritium breeding blanket for deuterium-tritium (D-T)–fueled fusion power plants (FPPs) achieving a high tritium breeding ratio (TBR) in neutronics models. However, tritium breeding in FLiBe is inherently difficult to study experimentally. As a result, the coupled issues of FLiBe radiochemistry and tritium (T) transport are poorly understood. LIBRA approaches this challenge by simulating an FPP blanket environment using a D-T neutron generator and 1000 kg of FLiBe. LIBRA will investigate T breeding, containment, and extraction, coupled with FLiBe redox control and radiochemistry. The primary goal of LIBRA is to demonstrate robust T accountancy in blanket prototypical conditions. Here, T accountancy encompasses accurate predictions of T breeding in the FLiBe; detection and measurement of all T bred in LIBRA; and speciation of the T extracted from the FLiBe. Initial neutronics simulations of LIBRA indicate that a global TBR of 1 is possible, where the TBR is defined as the number of tritons bred and extracted from FLiBe relative to the number of neutrons produced by D-T fusion reactions in the neutron generator. In this paper, we present the LIBRA concept and its scientific goals in the context of T breeding experiments. We also consider the potential impact of the LIB on the future fusion power industry, motivating further development of FLiBe-based T breeding research activities such as LIBRA.

Coordinated design of the cartridge-type blanket and the ceramic pebble divertor for the helical reactor FFHR-b3

The cartridge-type blanket for the helical reactor has been updated and named the CARDISTRY-B3. Together with this, the ceramic pebble ergodic limiter/divertor system named the REVOLVER-D3 has been adopted in the design. There was a difficulty in fabricating the blanket cartridges due to the first wall shape with three-dimensional surface. This difficulty has been removed by introducing the slit first wall, where the first wall surface is formed by alternately stacking solid metal plates and porous plates made of metal or ceramic. The liquid metal flowing inside the blanket cartridge oozes out through the porous plates and forms the surface flow covering the first wall. Solid target plates for the helical divertor has been omitted, because the divertor region is already covered and protected by the liquid metal surface flow. The liquid metal freely falls inside the blanket cartridges by the gravity force and then exhausted to the pool set on the lower port. The ceramic pebbles, which are dropped to the inboard-side ergodic layer to form the flowing limiter divertor, also flow into the liquid metal pool. The pebbles and liquid metal are individually elevated again by screw conveyers. Since the free surface of the liquid metal is exposed to the plasma, the vapor pressure of the working liquid metal should be low enough. Ternary or quadruple alloys, which include Li, Sn, Pb (or Bi), and Er, have been selected as the candidates of the functional liquid metal for the CARDISTRY-B3 that satisfies the requirements of low vapor pressure, low density, low melting point, and high tritium breeding ratio. In this study, the coordinated design of CARDISTRY-B3 and REVOLVER-D3 for the FFHR-b3 is presented. The FFHR-b3 is one of the design options that aims at demonstration of 100 MW net electricity with two times larger devise size than LHD.

Novel features of the helical volumetric neutron source FFHR-b2

The compact helical fusion reactor FFHR-b2 is a multipurpose HElical Volumetric Neutron Source available as the component test facility to test the divertor and blanket systems. In the latest design, three new technologies of the high-temperature superconducting conductor, the ceramic pebble divertor, and the cartridge-type blanket are adopted. Tin-based functional liquid metals (FLMs) including lithium are considered as the working fluid for the blanket. The FLMs have multi functions of low vapor pressure, low density, low melting point, low corrosiveness, high tritium breeding ratio, and so on. The target construction cost of the FFHR-b2 has been set to 200 billion JPY. The main target of FFHR-b2 is to demonstrate the fusion gain larger than one. To achieve this in a relatively small device, the tangential neutral beam injection (NBI) heating with a moderate energy of 80 keV is adopted for the FFHR-b2. The beam-plasma fusion with 5 MW of NBI heating can produce a fusion power of 5 MW that is larger than the absorbed power of 3 MW.

Conceptual design of the supercritical CO2 cooled lithium lead blanket for CFETR

A conceptual design of the supercritical CO2 cOoled Lithium-Lead (COOL) blanket has been proposed as one advanced blanket candidate for the Chinese Fusion Engineering Testing Reactor (CFETR). At present, the COOL blanket is designed to fulfill the requirement of operating under the fusion power of 1.5 GW of CFETR while realizing the tritium self-sufficiency. Structural designs are carried out for outboard and inboard blanket segments and analyses with regard to neutronics and thermo-hydraulics are conducted to evaluate the blanket performance. Results indicate that a comparatively high Tritium Breeding Ratio (TBR) of 1.183 and a PbLi outlet temperature of 600 – 700°C can be achieved for this blanket concept. Besides, a Power Conversion System (PCS) based on the supercritical CO2 recompressing cycle is preliminarily designed for the blanket and the gross thermal efficiency is estimated to be 39% – 46% when assuming the turbine inlet temperature ranges between 550°C and 650°C.

Neutronic experiment and analyses of a hybrid tritium breeding blanket mockup for CFETR

A neutronic experiment has been performed using DT neutrons on a blanket mockup of the hybrid tritium breeding blanket (HTBB) at CAEP. As a backup blanket configuration for CFETR, HTBB design which uses uranium alloy, water coolant and ceramic breeder pebble bed has been proposed to obtain high tritium breeding ratio. The experiment has provided validation not only to the preliminary design of HTBB, but also to calculation code and nuclear data used in the simulations. Various neutron responses such as fission reaction rate, plutonium production rate, activation reaction rate and tritium production rate (TPR) have been measured. Numerical analyses were carried out with MCNP-4C code using ENDF/B-VII.1, TENDL-2017, IRDFF-v1.05 and FENDL3.1 nuclear data libraries. The comparisons between the calculation and experimental results show an agreement within about 10% for most of the measured quantities. The ratios of calculated to measured results (C/E) of tritium production rate are in 0.98–1.04 range which shows good agreement.

Pressure wave behavior and its effects on structure under In-box LOCA in a helium-cooled lead lithium blanket of hydrogen fusion reactors

The helium-cooled lead lithium (PbLi) blanket is considered as one of the candidate blanket concepts selected for the hydrogen fusion DEMO reactors and beyond, which has the advantages of simple structure, strong heat removal capacity and high tritium breeding ratio. However, due to the harsh environment such as high-energy neutron irradiation, high thermal load and great pressure gradient, there is a high possibility that one or some of the thousands of coolant channels will break in the breeding zone, which is so-called In-box Loss of Coolant Accident (In-box LOCA). When the accident occurs, the high pressure helium will rapidly inject into the lead lithium flow channel, generating a complex two-phase flow and great pressure shock effect, which may cause the peak pressure to exceed the design limit and threaten the integrity of the blanket structure. Therefore, it is of great significance to perform the transient analysis of in-box LOCA to improve the safety of the blanket and avoid the leakage of radioactive materials. In this paper, a two-way coupling model for fluid-solid interaction was established based on the ANSYS Workbench, and the model were validated through the experimental data obtained by injecting the high pressure helium gas into liquid lithium lead. Then the validated model was applied to the transient pressure wave propagation analysis and structural stress analysis of the Dual-Functional Lithium Lead (DFLL) blanket in order to explore the integrity of blanket structure under In-box LOCA. In addition, the effects of break location on pressure and structural stress was also investigated through six cases. The study found that the transient pressure in the DFLL blanket gone through three stages in any case: step rise, oscillate, and flatten out. Pressure peaks occurred during oscillations and their values were strongly dependent on the break location. The closer to the inlet/outlet, the higher the peak pressure was. The maximum pressure reached more than twice of the inlet pressure (up to ~16 MPa). As a result, the structural stress in some local areas has exceeded the allowable limits, and the corresponding suggestions for improvement have also been put forward. This study can provide guidance for safety design, operation and accident mitigation measures of helium-cooled lead lithium blankets.

Neutronics analysis of the stellarator-type fusion-fission hybrid reactor based on the CAD optimization method

The stellarator plasmas devices have some advantages in steady state operation with less magneto hydrodynamic activities, and the development of stellarator-type FFHR (Fusion-Fission Hybird Reactor) can provide viable energy in the foreseeable future with less requirements to the plasmas technology compared with the pure fusion system. However, the neutronics design of the stellarator-type FFHR has some challenges and difficulties, since the construction of helical structures with large number of high order spline surfaces are the extremely complicated process, and the physical parameters of the complex model have to be changed frequently in the primary design stage, which make the corresponding manual input simulation work into the time consuming and error prone tasks. In this context, the CAD optimization method has been used in converting the dense mesh oriented CAD structure to the low poly style, and the CAD-PSFO code has been employed for the modeling transformation process in Monte Carlo calculation. The thorium resource has been selected as the breeding material for the stellarator-type FFHR, and the polonium and natural abundance of uranium have been used as the seed fuel for the initial loading in order to get the high energy multiplication factor. The OMCB (OpenMC basedBurnupcode) has been introduced to study the burnup features, and stellarator-type FFHR has been designed as the multi-purposes facility that can keep tritium self-sufficiency and get the high breeding ratio with steady energy output.

Interference of Zr(IV) during the extraction of trivalent Nd(III) from the aqueous waste generated from metallic fuel reprocessing

Abstract A metallic alloy of uranium–zirconium and uranium–plutonium–zirconium has been proposed as a fuel for fast reactors, owing to the possibility of achieving high breeding ratio in a short span of time. About 6–10 wt.% of zirconium has been added to these actinide fuels to increase the melting temperature and thermal-mechanical stability. Aqueous reprocessing of the spent metallic fuel generates the high-level liquid waste (HLLW) that contains about 60 % of the total zirconium from the fuel. In view of this, the extraction behavior of a trivalent representative ion, Nd(III) in the presence of Zr(IV) was studied from nitric acid medium using the candidate ligands proposed for trivalent actinide separation from HLLW, such as N,N,N′N′-tetraoctyldiglycolamide (TODGA), and N,N-di-octyl-2-hydroxyacetamide (DOHyA). The extraction was studied as a function of nitric acid concentration, zirconium and neodymium concentration and Nd(III) to Zr(IV) ratio. The findings of dynamic light scattering (DLS) and ATR-FTIR spectral techniques were used for understanding the complex chemistry of Zr(IV) extraction under different conditions. Poor extraction of nitric acid, smaller aggregate size, no third phase formation during the extraction of Zr(IV) and Nd(III) and other unique solvent properties favor the DOHyA molecule in n-dodecane as a solvent for partitioning of trivalent actinides from HLLW generated from metallic fuel reprocessing.

New Exploration on TMSR: Redesign of the TMSR Lattice

Molten salt reactor (MSR) has been recognized as one of the next-generation nuclear power systems. Most MSR concepts are the variants evolved from the Oak Ridge National Laboratory (ORNL's) molten-salt breeder reactor (MSBR), which employs molten-salt as both fuel and coolant, and normally graphite is used as moderator. Many evaluations have revealed that such concepts have low breeding ratio and might present positive power coefficient. Facing these impediments, thorium molten salt reactor (TMSR) with redesigned lattice is proposed in this paper. Based on comprehensive investigation and screening, important lattice parameters including molten salt fuel composition, solid moderator material, lattice size, structure and lattice pitch to channel diameter (P/D) ratio are redesigned. In this paper, a fuel composition without BeF2 is adopted to increase the solubility for actinides (ThF4, UF4), and BeO is introduced as moderator to improve neutron economy. Moreover, lattice size and structure with cladding to separate fuel and moderator were also optimized. With these lattice parameters, TMSR has a high breeding ratio close to 1.14 and a short doubling time about 15 years. Meanwhile, a negative power coefficient is maintained. Based on this lattice design, TMSR can have excellent performance of safety and sustainability. SONG/TANG-MSR codes system is applied in the simulation, which is independently developed by Shanghai Nuclear Engineering Research & Design Institute (SNERDI).

Optimization of Th-U fuel breeding based on a single-fluid double-zone thorium molten salt reactor

Molten salt reactor (MSR) shows great promise of high thermal-electric conversion efficiency, inherent safety, and on-line reprocessing. Furthermore, fuel breeding can be realized in both thermal and fast spectrum reactors with Th-U fuel cycle, which can make use of the abundant resource of thorium.In order to combine the advantages of high breeding ratio (BR) in fast spectrum and low fissile inventory for criticality in thermal spectrum, a single-fluid double-zone thorium-based molten salt reactor (SD-TMSR) core configuration is proposed. First, assemblies in both the inner and the outer zones are optimized by adjusting the ratios of molten salt and graphite with consideration of BR, 233U inventory, double time (DT), and temperature coefficient of reactivity (TCR) at the startup time, aiming to have a minimal DT for fuel breeding and a negative TCR for security. The results show that DT has the optimum value when the ratios of molten salt and graphite in the inner zone and the outer zone are 0.357 and 1.162, respectively. And the TCR can be improved to −2 pcm/K when the side length of the graphite hexagonal prism is 7.5 cm. Then, based on the optimized geometry, the burn-up calculations with a series of on-line reprocessing rates are carried out with a self-developed MSR reprocessing sequence (MSR-RS). The results show that on-line reprocessing is beneficial to the Th-U fuel breeding. When the reprocessing rate is 200 l/d, iso-breeding can be achieved. When the reprocessing rate is 5 m3/d, the performance of Th-U breeding is improved significantly, and only 16 years is needed for the 233U doubling. The results also show that TCR remains negative and inherent safety is satisfied during all the operation time.

Influences of 7Li enrichment on Th–U fuel breeding for an Improved Molten Salt Fast Reactor (IMSFR)

The molten salt fast reactor (MSFR) shows great promise with high breeding ratio (BR), large negative temperature coefficient of reactivity, high thermal–electric conversion efficiency, inherent safety, and online reprocessing. Based on an improved MSFR optimized by adding axial fertile salt and a graphite reflector, the influences of $${}^{7}\hbox {Li}$$ enrichment on Th–U breeding are investigated, aiming to provide a feasible selection for the molten salt with high fissile breeding and a relatively low technology requirement for $${}^{7}\hbox {Li}$$ concentration. With the self-developed molten salt reactor reprocessing sequence based on SCALE6.1, the burn-up calculations with online reprocessing are carried out. Parameters are explored including BR, $${}^{233}\hbox {U}$$ production, double time (DT), spectrum, $${}^{6}\hbox {Li}$$ inventory, neutron absorption, and the tritium production. The results show that the $${}^{7}\hbox {Li}$$ enrichment of 99.95% is appropriate in the fast fission reactor. In this case, BR above 1.10 can be achieved for a long time, corresponding to the $${}^{233}\hbox {U}$$ production of 130 kg per year and DT of 36 years. After 80 years’ operation, the tritium production for 99.5% is only about 7 kg, and there is no obvious increase compared to that for 99.9995%.

Core scheme and depletion analysis of new thorium molten salt reactor

Core layout and depletion features of thorium molten salt reactor(TMSR) are analyzed using the self-developed lattice and core code SONG/TANG-MSR. Based on the results of previous lattice optimization research, fuel salt that contains no BeF2 is adopted, BeO is introduced as moderator instead of graphite, and SiC is chosen as cladding material. According to further lattice analysis at core level, the optimized three-zone layout of core is obtained, with high breeding ratio, negative power coefficient, flat temperature distribution. Thereafter, the depletion calculation and analysis is performed considering the online processing of fuel salt. The results show that the core has high breeding ratio, short doubling time, and long term stable operation. Thus, the design of TMSR has been greatly improved with increased breeding ability and sufficient safety margins.

Reactor core and actinide production evaluation based on different loading material of recyled spent nuclear fuel of LWR in FBR

Reactor core analysis is one of the important factor for evaluating reactor operation and safety concern as well as the evaluation on fuel breeding of nuclear fuel to show nuclear fuel sustainability. Actinide compositions including plutonium and minor actinides production are also evaluated for estimating the plutonium non-proliferation aspect as well as for nuclear fuel breeding aspect. Loaded initial fuels in fast breeder reactor (FBR) are based on the spent nuclear fuel (SNF) of light water reactor (LWR). Those SNF of LWR will vary depending on LWR operation such as fuel burnup levels as well as cooling time process after the reactors are shutdown. In order to optimize the analysis of FBR design, a large FBR type with core and blanket fuel arrangements as driver fuels and breeding fuel regions as well as adopted some specific fuel batches and cycle length systems were adopted as a basic analysis case. Some obtained results are showing that reactor core performance such as criticality and breeding ratio are depending on the loaded fuel composition as well as operating reactor operation. Less criticality condition and higher breeding ratio are obtained by longer cooling time process of loaded SNF LWR. In addition, higher burnup of loaded SNF LWR achieves less criticality condition, while fuel breeding ratio profile is obtained higher for higher burnup. Loaded fuel composition of U-TRU fuel gives higher breeding ratio and its breeding ratio become higher for longer cooling time of loaded fuels. Actinide element compositions become less at end of equilibrium cycle (EOEC) of FBR, except for plutonium (Pu) and curium (Cm) which obtains higher composition. It shows the utilization of trans-uranium fuel type such as U-TRU type gives a significant production of minor actinide which can be estimated to contribute for reducing the excess reactivity as well as to increase nuclear fuel breeding capability. Less criticality is effective to reduce the excess reactivity at beginning of cycle for safety concern and higher breeding profile is indicating that more fuel sustainability aspect of nuclear fuel is gained.

Trends of Publications and Patents on Metallic Fuel Development for Fast Reactors

Higher breeding ratio, high thermal conductivity, shorter doubling time and high plutonium production make metallic fuels a viable solution compared to oxide/nitride/carbide/silicide fuels for cost-effective commissioning of many power reactors. Metallic fuels lend themselves to compact and simplified reprocessing and re-fabrication technologies, a key feature in a novel concept for the deployment of fast reactors. Satisfactory physical and technical characteristics of fuel rods with metallic fuel have been demonstrated at high burn-ups, and comparatively easy reprocessing of spent fuel using the pyro-metallurgical method makes this fuel relevant in fast reactors development. The present work is complemented with a scientometric study.

Analysis of accidents and abnormal transients of a high breeding fast reactor cooled by supercritical-pressure light water

A high breeding core of supercritical water cooled fast reactor (Super FBR) is designed with the tightly packed fuel assembly for obtaining a high breeding ratio with negative void reactivity. The coolant volume fraction is substantially smaller than that of the tight lattice fuel assembly in Super FRs. The present study conducted the safety analysis of this reactor for the abnormal transients and accidents at supercritical pressure. The safety system and safety criteria are similar to those of Super FRs. The accident “control rod ejection” gives the highest fuel cladding temperature and the highest peak pressure, although which are still within the limit of safety criteria. The overall results show that all the safety criteria are satisfied at the selected events.

Neutronics and thermo-hydraulic design of supercritical-water cooled solid breeder TBM

In this paper, the supercritical-water cooled solid breeder test blanket module (SWCB TBM), using the supercritical water as the coolant, Li4SiO4 lithium ceramic pebbles as a breeder, and beryllium pebbles as a neutron multiplier, was designed and analyzed for ITER.The results of neutronics, thermo-hydraulic and thermo-mechanical analysis are presented for the SWCB TBM. Neutronics calculations show that the proposed TBM has high tritium breeding ratio and power density. The tritium breeding ratio (TBR) of the proposed design is 1.17, which is greater than that of 1.15 required for tritium self-sufficiency. The thermo-hydraulic calculation proved that the TBM components can be effectively cooled to the allowable temperature with the temperature of outlet reaching 500°C. According to thermo-mechanics calculation results, the first wall with the width of 17mm is safe and the deformation of first wall is far below the limited value. All the results showed that the current TBM design was reasonable under the ITER normal condition.

Optimization of temperature coefficient and breeding ratio for a graphite-moderated molten salt reactor

Molten salt reactor (MSR) has fascinating features: inherent safety, no fuel fabrication, online fuel reprocessing, etc. However, the graphite moderated MSR may present positive feedback coefficient which has severe implications for the transient behavior during operation. In this paper, the feedback coefficient and the breeding ratio are optimized based on the fuel-to-graphite ratio variation for a thorium based MSR (TMSR). A certain thermal core with negative feedback coefficient and relative high initial breeding ratio is chosen for the reprocessing scheme analysis. The breeding performances for the TMSR under different online fuel reprocessing efficiencies and frequencies are evaluated and compared with other MSR concepts. The results indicate that the thermal TMSR can get a breeding ratio greater than 1.0 with appropriate reprocessing scheme. The low fissile inventory in thermal TMSR leads to a short doubling time and low transuranic (TRU) inventory. The lifetime of graphite used for the TMSR is also discussed.