An approach to improve the economy of desalination plants with a nuclear heating reactor by coupling with hybrid technologies

Nuclear heating reactor is a new type of power plant that uses nuclear energy as heat source. Low temperature nuclear heating reactor should be the forerunner and main force for developing nuclear heating plant in China. Due to the lower water temperature required by the heating system, this dedicated, non-power generating nuclear reactor works at low temperatures and pressures with inherent safety features. The design, construction and operation of the nuclear heating reactors in various countries in the world were reviewed in this paper, and China’s new demonstration nuclear heating project and NHR-200 low-temperature heating reactor which would be used was discussed in the paper. We put forward the developing route and suggestion for the development of low-temperature heating reactor in China.

Coupling of nuclear heating reactor with desalination process

Some features of the nuclear heating reactor (NHR) design in China

Economic analysis of a 2×200 MW nuclear heating reactor for seawater desalination by multi-effect distillation (MED)

Small break loss of coolant accident (SBLOCA) is one of the most important severe accidents in nuclear heating reactor. Nuclear heating reactor designed by Tsinghua University, whose primary loop is integrated layout and designed without main pump. The initial water volume in the reactor vessel is important to determine whether the reactor will be cooled or not as no safety injection system is designed for coolant makeup during the whole scenario. This paper simulates SBLOCA in nuclear heating reactor based on RELAP5. Transient behavior of relevant thermal parameters is specifically analyzed. Moreover, investigation also has been made on SBLOCA scenario based on different residual heat removal correlations and found the long-term residual heat removal capacity is decisive in determining the loss of coolant. The mathematical form of residual heat removal correlation is specifically deducted and can be widely applied to different situations. The envelope line that differentiates the region whether the core is safe or not under different maximum PRHRS capacity is also given.

Adaptive state-observer for monitoring flexible nuclear reactors

The 200 MWth nuclear heat reactor II (NHR200-II) is a typical integral pressurized water reactor (iPWR) being developed by the institute of nuclear and new energy technology (INET) in Tsinghua university. The NHR200-II, which has inherent safety features such as full-range natural circulation, passive residual heat removal, self-pressurization and control rod hydraulically driving, can be adopted as a clean base-load energy source for a sea-water desalination plant having the process of multi-effect desalination with thermal vapor compression (MED-TVC). Dynamic modelling of the sea-water desalination plant coupled by the NHR200-II and MED-TVC is necessary for the design of its plant control strategy, which is important for the stable and efficient operation. In this paper, a lumped parameter dynamic model of NHR200II-based sea-water desalination plant with the process of MED-TVC is proposed based upon the conservation laws of mass, momentum and energy. The modeling verification in both the steady-state and open-loop dynamic-state are given, which show the suitability of applying this model for control system design. Finally, the closed-loop responses in the case of power-level maneuver from 100% to 50% full power is given.

The nuclear heating reactor (NHR) is a typical integral pressurized water reactor (iPWR) developed by the institute of nuclear and new energy technology (INET) of Tsinghua University, which has the safety advanced features such as the primary circuit integral arrangement, full-range natural circulation, self-pressurization. Power-level control is crucial for the operational stability and efficiency of the NHR, and the dynamic modeling is a basis for control system design and verification. From the conservation laws of mass, energy and momentum, a lumped-parameter dynamical model is proposed for the nuclear steam supply system (NSSS) based on the 200MWth nuclear heating reactor II (NHR200-II). The steady-state model validation is given by the comparing the parameter values of this model and that for plant design. Then, both the open-loop responses under the disturbances of reactivity and coolant flowrates as well as the closed-loop responses under the case of power ramp are given, where the rationality of the responses are analyzed from the viewpoint of plant physics and thermal-hydraulics. This model can be utilized for not only the control system design but also the development of a real-time simulator for the hardware-in-loop control system verification.

Nuclear heating reactor is integrated designed without main pump and safety injection system. The loss of coolant accidents are mainly in the form of small break LOCA. As no safety injection system is designed for coolant makeup, the water volume in the reactor vessel is critical since it determines whether the reactor will be submerged during the whole scenario. Therefore, the study on coolant loss in this pool system is indispensable. The RELAP5 code has been developed for best-estimate transient simulation of light water reactor coolant systems during postulated accidents. The long term effect in nuclear heating reactor is important. In this paper we investigated the influential factors on SBLOCA scenario and found the long term residual heat removal capacity is decisive in determining the loss of coolant. The residual heat removal capacity should be greater than 2% of reactor thermal power if ensuring the core submerged in the long run.

Seismic responses of a fuel assembly in nuclear heating reactor

Fuel storage facilities for 200-MW nuclear heating reactor

Flexible control of nuclear cogeneration plants for balancing intermittent renewables

The nuclear heating reactor (NHR), which is independently developed and designed by Tsinghua University, adopts full power natural circulation and has the characteristic of low mass flow rate. Due to the different operating conditions between NHR and PWR, it is necessary to analyze the thermal hydraulic performance of the fuel assembly in NHR core. The thermal hydraulic performance of the fuel assembly is mainly analyzed by the subchannel analysis code which can accurately calculate the two-phase parameters. Therefore, this research employs subchannel analysis code COBRA-TF to simulate radial uniform and non-uniform heating assembly under the operating conditions of low mass flow rate from 300kg/m2s to 700kg/m2s. The simulations are validated on the thermal hydraulics experiments of GE and Studsvik. The result shows that the calculated mass flow rate and void fraction are in good agreement with the experimental results. Under radial nonuniform heating condition, high outlet mass velocity appears in low heating power region, while the high outlet void fraction appears in high heating power region. Compared with the uniform heating condition, the distribution of the mass flow rate and the void fraction in subchannels under non-uniform heating conditions varies greatly. In addition, the transverse flow and the counterflow also can be observed in non-uniform heating conditions under low mass flow rate conditions. The appearance of such transverse flow mainly results from the radial pressure difference, rather than the turbulent mixing and void drift. Therefore, it is essential to predicate the distribution of the mass flow rate and the void fraction under non-uniform heating and low mass flow rate conditions.

Power-level regulation is a key technique that provides both economic and stable operation for any nuclear reactors. Since the pressurized water reactors (PWR) are not only the most widely utilized nuclear reactors but also complex nonlinear systems, efficient nonlinear power-level control strategy is meaningful to not only the development of nuclear power technique but also the renaissance of nuclear energy. Further, since self-stability is the most crucial dynamic feature of nuclear reactors, it is necessary to give self-stability analysis which gives the means to strengthen the closed-loop stability. In this paper, a sufficient condition of globally asymptotic self-stability for PWR is firstly given. Based on this self-stability analysis, a nonlinear state-feedback power-level control law with L2 disturbance attenuation performance is then proposed. This new controller is applied to the power-level regulation of a nuclear heating reactor (NHR), and simulation results show its feasibility and high performance.

Growing electricity requirement and the serious pollution caused by burning petroleum and coal give the current rebirth of nuclear energy industry. State observation is one of the key and basic technologies of system monitoring which is very necessary to the safe and effective operation of today's nuclear reactors. Since nuclear reactors are complex and nonlinear systems, it is quite necessary to design a nonlinear state-observer with high-performance for nuclear reactors. A dissipation-based high gain filter (DHGF) is presented for nonlinear systems in this paper, and robustness analysis is also given. The DHGF is then applied to the state-observation for a nuclear heating reactor (NHR), and simulation results show the feasibility of the DHGF.