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

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

Study of a self-regulated nuclear burn wave regime in a fast reactor based on a thorium–uranium cycle

The possibility of a wave of slow nuclear burning in a fast reactor in thorium–uranium fuel cycle is investigated. The calculations were performed using a model based on the solution of a nonstationary nonlinear diffusion equation for a cylindrical homogeneous reactor using the concept of a radial geometric factor (buckling) and the effective multigroup approximation taking account of the nuclear kinetics of the precursors of delay neutrons and burnup and production of the main nuclides of the thorium–uranium fuel cycle. The calculations showed that the generation and propagation of a wave of nuclear burning traveling with velocity approximately 2 cm/yr are possible in a thorium–uranium medium. However, the addition of even small quantities of a construction material and coolant to the composition of the reactor makes it impossible to obtain the burn wave regime. A self-maintained nuclear burn regime is also established in this case and exists for a long time (∼5 yr), but the system does not transition into a regime with a nuclear burn wave propagating along the axis of the reactor.     

Numerical simulations of gas–liquid–particle three-phase flows using a hybrid method

For the analysis of debris behavior in core disruptive accidents of liquid metal fast reactors, a hybrid computational tool was developed using the discrete element method (DEM) for calculation of solid particle dynamics and a multi-fluid model of a reactor safety analysis code, SIMMER-III, to reasonably simulate transient behavior of three-phase flows of gas–liquid–particle mixtures. A coupling numerical algorithm was developed to combine the DEM and fluid-dynamic calculations, which are based on an explicit and a semi-implicit method, respectively. The developed method was validated based on experiments of water–particle dam break and fluidized bed in systems of gas–liquid–particle flows. Reasonable agreements between the simulation results and experimental data demonstrate the validity of the present method for complicated three-phase flows with large amounts of solid particles.     

CFD investigation of flow inversion in typical MTR research reactor undergoing thermal–hydraulic transients

Three dimensional CFD full simulations of the fast loss of flow accident (FLOFA) of the IAEA 10 MW generic MTR research reactor are conducted. In this system the flow is initially downward. The transient scenario starts when the pump coasts down exponentially with a time constant of 1 s. As a result the temperatures of the heating element, the clad, and the coolant rise. When the flow reaches 85% of its nominal value the control rod system scrams and the power drops sharply resulting in the temperatures of the different components to drop. As the coolant flow continues to drop, the decay heat causes the temperatures to increase at a slower rate in the beginning. When the flow becomes laminar, the rate of temperature increase becomes larger and when the pumps completely stop a flow inversion occurs because of natural convection. The temperature will continue to rise at even higher rates until natural convection is established, that is when the temperatures settle off. The interesting 3D patterns of the flow during the inversion process are shown and investigated. The temperature history is also reported and is compared with those estimated by one-dimensional codes. Generally, very good agreement is achieved which provides confidence in the modeling approach.     

Numerical and theoretical investigations of heat transfer characteristics in helium–xenon cooled microreactor core

Helium–xenon cooled microreactors are a vital technological solution for portable nuclear reactor power sources. To examine the convective heat transfer behavior of helium–xenon gas mixtures in a core environment, numerical simulations are conducted on a cylindrical coolant channel and its surrounding solid regions. Validated numerical methods are used to determine the effect and mechanisms of power and its distribution, inlet temperature and velocity, and outlet pressure on the distribution and...     

Transient behavior of the sodium–potassium alloy heat pipe in passive residual heat removal system of molten salt reactor

High temperature heat pipes, as highly-effective heat transfer elements, have been extensively employed in thermal management for their remarkable advantages in conductivity, isothermality and self-actuating. It is of significance to apply heat pipes to new concept passive residual heat removal system (PRHRS) of molten salt reactor (MSR). In this paper, the new concept PRHRS of MSR using sodium–potassium alloy (NaK) heat pipes is proposed in detail, and then the transient behavior of high temperature NaK heat pipe is numerically investigated using the Finite Element Method (FEM) in the case of MSR accident. The two-dimensional transient conduction model for the heat pipe wall and wick structure is coupled with the one-dimensional quasi-steady model for the vapor flow when vaporization and condensation occur at the liquid–vapor interface. The governing equations coupled with boundary conditions are solved by FORTRAN code to obtain the distributions of the temperature, velocity and pressure for the heat pipe transient operation. Numerical results indicated that high temperature NaK heat pipe had a good operating performance and removed the residual heat of fuel salt significantly for the accident of MSR.     

Improved method for measuring the γ-ray absorbed dose rate in models of iron shielding of a thermonuclear reactor

An improved method of measuring the absorbed γ-ray dose rate usingCaSO ₄ andSrSO ₄ type thermoluminescent detectors in models of iron shielding of a thermonuclear reactor is described. The reactionT(d, n)⁴ He served as a neutron source. The method obtained makes it possible to determine the absorbed γ-ray dose rate in shielding without using computed information and relying only on experimental data on the rates of nuclear reactions in threshold detectors. 7 figures, 1 table, 9 references. Moscow Engineering-Physics Institute. Translated from Atomnaya énergiya, Vol. 86, No. 3, pp. 219–225, March, 1999.     

Treatment of radioactive waste seawater by coagulation–flocculation method using ferric hydroxide and poly acrylamide

This work studied a coagulation–flocculation system using ferric hydroxide and poly acrylamide flocculant to find the effective condition to remove the radionuclides of Co, Mn, Sb, Ru, and Sr remaining in solution after Cs removal by adsorption for the treatment of radioactive waste seawater generated in a disastrous nuclear power plant accident like Fukushima Daiichi. The coagulation–flocculation mechanism was studied, and the performance characteristics of the coagulation–flocculation system was evaluated in views of decontamination yield of the elements, residual turbidity of treated solution, settling speed of flocs, and generated total floc volume, etc. The total removal yield of target radio nuclides of Co, Mn, Sb, and Ru was more than 99% in seawater at pH 8.     

Characterization of a sodium-cooled fast reactor in an MHR–SFR synergy for TRU transmutation

In the task of destroying the light water reactor (LWR) transuranics (TRUs), we consider the concept of a synergistic combination of a deep-burn (DB) gas-cooled reactor followed by a sodium-cooled fast reactor (SFR), as an alternative way to the direct feeding of the LWR TRUs to the SFR. In the synergy concept, TRUs from LWR are first deeply incinerated in a graphite-moderated DB-MHR (modular helium reactor) and then the spent fuels of DB-MHR are recycled into the closed-cycle SFR. The DB-MHR core is 100% TRU-loaded and a deep-burning (50–65%) is achieved in a safe manner (as discussed in our previous work). In this analysis, the SFR fuel cycle is closed with a pyro-processing technology to minimize the waste stream to a final repository. Neutronic characteristics of the SFR core in the MHR–SFR synergy have been evaluated from the core physics point of view. Also, we have compared core characteristics of the synergy SFR with those of a stand-alone SFR transuranic burner. For a consistent comparison, the two SFRs are designed to have the same TRU consumption rate of ∼250 kg/GW EFPY that corresponds to the TRU discharge rate from three 600 MW DB-MHRs. The results of our work show that the synergy SFR, fed with TRUs from DB-MHR, has a much smaller burnup reactivity swing, a slightly greater delayed neutron fraction (both positive features) but also a higher sodium void worth and a less negative Doppler coefficients than the conventional SFR, fed with TRUs directly from the LWRs. In addition, several design measures have been considered to reduce the sodium void worth in the synergy SFR core.     

Fundamental burn-up mode in a pebble-bed type reactor

This paper deals with a pebble-bed type reactor, in which the fuel is loaded from one side (top) and discharged from the other side (bottom). A boundary value problem of a single group diffusion equation coupled with simplified burn-up equations is studied, where the natural radioactive decay processes are neglected in the burn-up modelling. An asymptotic burning wave solution is found analytically in the one-dimensional case, which is called as fundamental burn-up mode. Among this solution family there are two particular cases, namely, a classic fundamental solution with a zero burn-up and a partial solitary burn-up wave solution with a highest burn-up. An example of Th–U conversion is considered and the solutions are presented in order to show the mechanism of the burning wave.     

Numerical study on flow field in inlet plenum of a pebble-bed modular reactor

Flow distribution and pressure drop analysis in the inlet plenum of a pebble-bed modular reactor (PBMR) have been performed numerically. Three-dimensional Navier–Stokes equations have been solved in conjunction with the k– model as a turbulence closure. Non-uniformity in the flow distribution is assessed for the reference case, and parametric studies have been performed for rising channels diameter, Reynolds number, angle between the rising channels, angle between the inlet ports, and aspect ratio of the plenum cross-section. Also, two different shapes of the inlet plenum, namely, rectangular and oval shapes, have been analyzed. The relative flow mal-distribution parameter variation shows that the flow distribution in rising channels for the reference case is strongly non-uniform. As the rising channels diameter is decreased, the flow uniformity as well as the pressure drop is found to increase. The flow distribution in the rising channels is independent of Reynolds number. Increase in the angle between the inlet ports and aspect ratio is found to increase the uniformity in flow distribution.     

Finite element nonlinear heat transfer analysis using a stable explicit method

The explicit stable method of Saulev is applied to nonlinear finite element heat conduction. Several nonlinear example problems are considered which include temperature-varying material properties and radiation boundary conditions.     

Reactivity enhancement of chemical materials used in packed bed reactor of chemical heat pump

In this study, a mixture of expanded graphite (EG) and magnesium hydroxide (Mg(OH)₂) was used to enhance the thermal conductivity and reactivity of a magnesium oxide/water (MgO/H₂O) chemical heat pump, because EG is chemically stable and has high thermal conductivity and high moldability to form the heat exchange structure. Calcium chloride (CaCl₂) was also introduced into the mixture of EG and Mg(OH)₂ to ensure smooth diffusion of vapor in materials and enhance the fittability between EG and Mg(OH)₂. The reaction kinetics of pure Mg(OH)₂, a mixed material containing Mg(OH)₂ and CaCl₂ (termed MC), and a mixed material containing EG, Mg(OH)₂, and CaCl₂ (termed EMC) were examined under the same reaction conditions by performing thermobalance measurements. EMC exhibited a higher dehydration rate than the other materials. It also exhibited hydration reactivity at temperatures of up to 200 °C; at this temperature, pure Mg(OH)₂ exhibited low reactivity. The addition of CaCl₂ also enhanced the hydration reactivity of MgO because of the high water adsorption ability of CaCl₂ in EMC. A reaction rate equation for the hydration of EMC was proposed on the basis of an assumed reaction model. The thermal performance of a MgO/H₂O chemical heat pump manufactured using EMC was evaluated from this equation. EMC was concluded to have good potential for use as a packed bed material in the MgO/water chemical heat pump owing to its low cost, high hydration reactivity, high thermal conductivity, and high moldability to form the heat exchange structure.     

CFD code application: calculating heat transfer in a reactor with supercritical parameters

Thermohydraulic calculations of isolated and communicating cells of a rod bundle were performed by the channel method for CANDU-X fuel assemblies and by a three-dimensional method. It was established that in solving the problem for the tightest cell in the case q = const the azimuthal nonuniformity of the temperature was found to decrease by 77°C but it too was inadmissibly large. The temperature distribution along the surface of a fuel element in the case q = const was found to be different from the solution of the adjoint problem. A region with elevated coolant temperature, impeding heat exchange between two neighboring cells, was found between two adjoining cells. It was found that to evaluate computational reliability an experimental study must be performed on rod assemblies with supercritical coolant parameters.     

圆柱形带反射层反应堆的数值传热计算

在结构化和非结构化网格中,采用有限容积方法,数值计算带有反射层(反射层布置分为轴向、径向和复合双向三种情况)的圆柱形反应堆的物理和热特性。首先采用单组法数值求解堆芯和反射层中的热中子注量率密度,并同其精确解相比较,验证彼此的正确性;然后用类似法确定堆内两区中热中子产生的热功率分布规律,并进行数值传热计算。所有结果都与没有反射层的反应堆(裸堆)状况进行比较,并且得到具有参考价值的结果。     

Boundary element method in nonlinear transient heat transfer of reactor solids with convection and radiation on surfaces

This paper deals with the boundary (integral) element method for non-steady conduction problems of solids, subject to non-linear convective and radiation conditions on surfaces. Boundary integral equations for the mixed-type and non-linear boundary conditions, both for the case with constant and variable heat conductivity are derived, modelled by non-conforming boundary elements, while domain integrals are evaluated within triangular cells. A test case is included to illustrate the described procedure.     

Analysis of heat source distribution in internal circulating surface heat transfer molten salt reactor

The Molten Salt Reactor (MSR) is one of the Generation IV nuclear reactor concepts that were selected by the Generation IV International Forum in 2000. The concept is based on liquid fuel instead of solid fuel assemblies. Besides the advantages, there are several aspects of operation that can hinder the realization of this reactor concept. In this paper, the authors investigate the neutronics behaviour of a new sub-concept that offers solutions for many of the technical problems. The analysis was performed using the particle transport code MCNPX 2.7. The paper focuses on the short-term and steady state heat source distribution in the fuel salt and in the graphite moderator. Accordingly, neither burn-up effects nor reactivity transients are considered. The sensitivity of the effective multiplication factor on different geometrical and material parameters was studied. The results obtained indicate that the main region of heat deposition is in the internal and external channels of the graphite moderator. Only a few percent of the total heat power is released in the graphite moderator, where the gamma and neutron related heat deposition is on the same scale. The results also prove that the heat source distribution does not change drastically upon the actuation of the control rods.