Research and Development of Nuclear Heating Reactors in China

The research and development (R & D) of nuclear heating reactors (NHRs) have been conducted as one of the national key projects in science and technology in China since the 1980s. This paper presents the development status. main design featur and safety concepts of the NHR.

Thermal-hydraulic Stability Analysis of Nuclear Heating Reactors

The two-phase flow instability that can occur in a natural circulation system is of importance in the design of nuclear heating reactors. The time domain code RETRAN-02 and the frequency domain code NUFREQ were applied to estimate the instability boundary and the effects of such parameters as pressure, inlet resistance and riser height in NHR-5 and an experimental loop. The results of the calculations and the experiments are in good agreement. Nonlinear density wave oscillations were analyzed using the RETRAN-02 code. The theory of nonequilibrium thermodynamics was used to find an explicit criterion to estimate the threshold of the stability. Experimental simulation of the nuclear feedback density wave instability was also carried out in a test loop using. computer controlled electric power.

Loss of Coolant Experiments for the Test Nuclear Heating Reactor

A series of tests were completed for three types of loss of coolant accidents (LOCAs) (pipe break in the gas plenum. near the liquid level and submerged under water) in the test nuclear heating reactor (NHR). Experiments show that the three cases of LOCAs (loss of coolant accidents) have different patterns. In the case of a pipe break connected to the gas plenum, the quantity of water lost is independent of the diameter of the broken pipe. In the case of a pipe located near the liquid level. the quantity of water lost depends on the location of the pipe. In the case of a pipe break below the water level. all the water above the break will be discharged. The discharge patterns for all three cases are analyzed in detail.

Experimental Study of a Stoppage Natural Circulation during a Nuclear Heating Reactor LOCA

The 5MW nuclear heating reactor is an integral natural circulation reactor. The rupture of the coolant penetrating tube is a typical accident causing coolant loss. When the water level drops down to the upper edge of the inlet of the heat exchanger, the natural circulation stops. This influences the core cooling and the stability of the main loop. A series of tests showed that there is a stable drop of pressure, and the heated element temperature is not too high to cause burnout. But the backward flow or flow oscillation in the primary coolant circuit occurs when the flow breaks completely in the end. The whole flow process is described and the mechanism is discussed.

Simultaneous production and utilization of methanol for methyl formate synthesis in a looped heat exchanger reactor configuration

In this investigation, a novel thermally coupled reactor （TCR） containing methyl formate （MF） production in the endothermic side and methanol synthesis in the exothermic side has been investigated. The interesting feature of this TCR is that productive methanol in the exothermic side could be recycled and used as feed of endothermic side for MF synthesis. Other important advantages of the proposed system are high production rates of hydrogen and MF. The configuration consists of two thermally coupled concentric tubular reactors. In these coupled reactors, autothermal system is obtained within the reactor. A steady-state heterogeneous model is used for simulation of the coupled reactor. The proposed model has been utilized to compare the performance of TCR with the conventional methanol reactor （CMR）. Noticeable enhancement can be obtained in the performance of the reactors. The influence of operational parameters is studied on reactor performance. The results show that coupling of these reactions could be feasible and beneficial. Experimental proof-of-concept is required to validate the operation of the novel reactor.

Comparative studies for two different orientations of pebble bed in an HCCB blanket

The Indian Test Blanket Module（TBM） program in ITER is one of the major steps in its fusion reactor program towards DEMO and the future fusion power reactor vision. Research and development（RD） is focused on two types of breeding blanket concepts： lead–lithium ceramic breeder（LLCB） and helium-cooled ceramic breeder（HCCB） blanket systems for the DEMO reactor. As part of the ITER-TBM program, the LLCB concept will be tested in one-half of ITER port no. 2, whose materials and technologies will be tested during ITER operation. The HCCB concept is a variant of the solid breeder blanket, which is presently part of our domestic RD program for DEMO relevant technology development. In the HCCB concept Li_2TiO_3 and beryllium are used as the tritium breeder and neutron multiplier, respectively, in the form of a packed bed having edge-on configuration with reduced activation ferritic martensitic steel as the structural material. In this paper two design schemes, mainly two different orientations of pebble beds, are discussed. In the current concept（case-1）, the ceramic breeder beds are kept horizontal in the toroidal–radial direction. Due to gravity, the pebbles may settle down at the bottom and create a finite gap between the pebbles and the top cooling plate, which will affect the heat transfer between them. In the alternate design concept（case-2）, the pebble bed is vertically（poloidal–radial） orientated where the side plates act as cooling plates instead of top and bottom plates. These two design variants are analyzed analytically and 2 D thermal-hydraulic simulation studies are carried out with ANSYS, using the heat loads obtained from neutronic calculations.Based on the analysis the performance is compared and details of the thermal and radiative heat transfer studies are also discussed in this paper.

A transient multiphysics coupling method based on OpenFOAM for heat pipe cooled reactors

Differing from traditional pressurized water reactors(PWRs),heat pipe cooled reactors have the unique characteristics of fuel thermal expansion,expansion reactivity feedback,and thermal contact conductance.These reactors require a new multiphysics coupling method.In this paper,a transient coupling method based on OpenFOAM is proposed.The method considers power variation,thermal expansion,heat pipe operation,thermal contact conductance,and gap conductance.In particular,the reactivity feedback caused by working medium redistribution in a heat pipe is also preliminarily considered.A typical heat pipe cooled reactor KRUSTY(Kilowatt Reactor Using Stirling TechnologY)is chosen as the research object.Compared with experimental results of load following,the calculated results are in good agreement and show the validity of the proposed method.To discuss the self-adjusting capability of this type of reactor system,a hypothetical accident is simulated.It is assumed that at the beginning of this accident,loss of the heat sink occurs.After 1500 s of the transient process,the reactor system recovers immediately.During this hypothetical accident,the control rod is always out of the reactor core,and the reactor only relies on the reactivity feedback to regulate the fission power.According to the simulation,the peak temperature is only about 1112 K,which is far below the safety limit.As for system recovery,the reactor needs approximately 2500 s to return to a steady state and can realize effective power regulation by reactivity feedback.This study confirms the availability of this coupling method and that it can be an effective tool for the simulation of heat pipe cooled reactors.

3D CFD modeling of acetone hydrogenation in fixed bed reactor with spherical particles

Acetone hydrogenation in a fixed bed reactor packed with spherical catalyst particles was simulated to study the effects of inlet gas velocity and particle diameter on hydrogenation reaction. Computational results show that the catalyst particles in the reactor are almost isothermal, and the high isopropanol concentration appears at the lee of the particles. With the increase of inlet velocity, the outlet isopropanol mole fraction decreases, and the total pressure drop increases drastically. Small diameter catalyst particles are favorable for acetone hydrogenation, but result in large pressure drop.

Dynamic Programming Method to Optimize Control Rod Positions in NHR-200

A code. OPTROD. based on dynamic programming method is developed and utilized to optimize control rod positions throughout the fuel cycle of the 200 MW nuclear heating reactor (NHR-200). A fuel cycle is divided into many steps or stages. Optimization of the multistage process is solved iteratively in the forward direction throughout a fuel cycle. The dynamic programming method is much more efficient than the normal nonlinear programming method. Convergence is obtained even if poor initial control rod positions are given.

Fuel Assembly Arrangement Optimization for NHR-200

This study considers optimization of the fuel assembly arrangement in the initial core loading of the 200 MW nuclear heating reactor (NHR-200). The enrichment of the fuel assemblies is used as the control variable with the objective to minimize the power peaking factor. The optimization methods are applied indirectly because it is difficult to directly relate the control variable and the object function in a single equation. Therefore, the solution uses linearized functons which are solved with linear programming. The corrected simplex method is used to solve the optimal problem. Useful engineer software has been designed and used in reactor physics design.

HTGR Process Heat Application Study

The 10MW high temperature gas-cooled reactor test module (HTR-10) is currently under construction.One of its objectives is to develop high temperature process heat applications. To realize this target, various high temperature gas-cooled reactor (HTGR) process heat applications have been analyzed. This paper briefly describes the possibilities and experimental schemes for using the HTR-10 for process heat application studies.