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.

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.

Dynamic Model for the Control System Simulation and Design of a 200 MW Nuclear Heating Reactor

The paper develops a nonlinear dynamic modelusedin a widerange controlsystem simulationfor a200 MWNuclear Heating Reactor (NHR 200).Besides a one pointneutron kinetics equation and temperature feedback based onthe lumped fuel and coolanttemperature , which are the usual methods used in modeling of PWR,two otherfactors are also considered in orderto suitthe wide range operation.The first considerationis the natural circulationinthe primary loop because it affectsthe heattransfer coefficients in the core and in the primary heat exchanger(PHE).The second consideration isthe flow rate variation in the secondaryloop which leads to some nonlinear properties.The simulationresultsshow thatthe modelis accurate enoughfor control system simulation. Some modelreduction basis can be obtained throughthe dynamic analysis.

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.

Design of Tokamak ELM Coil Support in High Nuclear Heat Environment

In Tokomak, the support of the ELM coil, which is close to the plasma and subject to high radiation level, high temperature and high magnetic field, is used to transport and bear the thermal load due to thermal expansion and the alternating electromagnetic force generated by high magnetic field and AC current in the coil. According to the feature of ITER ELM coil, the mechanical performance of rigid and flexible supports under different high nuclear heat levels is studied. Results show that flexible supports have more excellent performance in high nuclear heat condition than rigid supports. Concerning thermal and electromagnetic （EM） loads, optimized results further prove that flexible supports have better mechanical performance than rigid ones. Through these studies, reasonable support design can be provided for the ELM coils or similar coils in Tokamak based on the nuclear heat level.

Prediction of the Average Decay Heat per Fission for MOX Nuclear Fuel

MIXED Oxide Nuclear fuel (MOX) contains both uranium and plutonium in oxidized form. It is important to calculate the nuclear decay heat due to the single thermal fission (fission due to 0.0235 eV neutron) for all fissile nuclei in the MOX fuels (U235, Pu239, and Pu241). These fissile nuclei are the main source of the decay heat in MOX fuel. Decay heat calculation of the weighted fissile material content in MOX fuel is also important. A numerical method was used in this work to calculate the concentrations of all fission products due to the individual thermal fission of the three fissile materials as a function of time N(t). The decay heat calculations for the three fissile materials are directly calculated using the summation method by knowing the different concentrations of fission products over time. The average decay heat of the MOX fuel in induced thermal fission is also concluded. The most influential nuclei in the decay heat were also identified. The method used has been validated by several comparisons before, but the new in this work is using the most recent Evaluated Nuclear Data Library ENDF/B-VIII.0. Calculations of decay heat show very common trends for a period of 107 sec after the fission burst of thermal fissions of individual fissile nuclei. Moreover, the code showed high capability in calculating the fission fragments inventories and decay heats due to the decay of fission fragments of 31 fissionable nuclei.

Fission Fragment Decay Heat by Using the Most Recent Evaluated Nuclear Data Library ENDF/B-VIII

In this paper, a home-made code was designed to calculate the decay heat emitted by fission fragments as a result of successive radioactive emissions after a fission burst. The nuclear data necessary for the calculations was extracted from the latest version of the Evaluated Nuclear Data Library ENDF/B-VIII.0. The code can calculate the decay heat of thermal and fast neutron-induced fission reactions on the isotopes of Thorium, Protactinium, Uranium, Neptunium, Plutonium, Americium, Curium, California, Einsteinium, and Fermium. A numerical method was used in this work to calculate the decay heat of all fission fragments due to the individual thermal or fast fissions of the isotopes of the previous ten actinides. The most influential nuclei in the decay heat were also identified at different times after the fission event. Moreover, the code showed high capability in calculating the fission fragments inventories and decay heats due to the decay of fission fragments of 31 fissionable nuclei.

Neutronics Analysis of Water-Cooled Ceramic Breeder Blanket for CFETR

In order to investigate the nuclear response to the water-cooled ceramic breeder blanket models for CFETR, a detailed 3D neutronics model with 22.5° torus sector was developed based on the integrated geometry of CFETR, including heterogeneous WCCB blanket models, shield, divertor, vacuum vessel, toroidal and poloidal magnets, and ports. Using the Monte Carlo N-Particle Transport Code MCNP5 and IAEA Fusion Evaluated Nuclear Data Library FENDL2.1, the neutronics analyses were performed. The neutron wall loading, tritium breeding ratio, the nuclear heating, neutron-induced atomic displacement damage, and gas production were determined. The results indicate that the global TBR of no less than 1.2 will be a big challenge for the watercooled ceramic breeder blanket for CFETR.

Neutronic analysis of Indian helium-cooled solid breeder tritium breeding module for testing in ITER

India has proposed the helium-cooled solid breeder blanket concept as a tritium breeding module to be tested in ITER.The module has lithium titanate for tritium breeding and beryllium for neutron multiplication.Beryllium also enhances tritium breeding.A design for the module is prepared for detailed analysis.Neutronic analysis is performed to assess the tritium breeding rate,neutron distribution,and heat distribution in the module.The tritium production distribution in submodules is evaluated to support the tritium transport analysis.The tritium breeding density in the radial direction of the module is also assessed for further optimization of the design.The heat deposition profile of the entire module is generated to support the heat removal circuit design.The estimated neutron spectrum in the radial direction also provides a more in-depth picture of the nuclear interactions inside the material zones.The total tritium produced in the HCSB module is around 13.87 mg per full day of operation of ITER,considering the 400 s ON time and 1400 s dwell time.The estimated nuclear heat load on the entire module is around 474 kW,which will be removed by the high-pressure helium cooling circuit.The heat deposition in the test blanket model(TBM)is huge(around 9 GJ)for an entire day of operation of ITER,which demonstrates the scale of power that can be produced through a fusion reactor blanket.As per the Brayton cycle,it is equivalent to 3.6 GJ of electrical energy.In terms of power production,this would be around 1655 MWh annually.The evaluation is carried out using the MCNP5 Monte Carlo radiation transport code and FEDNL 2.1 nuclear cross section data.The HCSB TBM neutronic performance demonstrates the tritium production capability and high heat deposition.

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.