The fourth generation nuclear power of the fourth generation nuclear energy system

The original intention of the United States to develop the fourth generation nuclear power plant is mainly to prevent nuclear proliferation, and the goal is to develop a closed small reactor nuclear power plant with long-life core for developing countries. However, after the international working group meeting in May 2000, the GIF meeting in Seoul in August 2000 and the Paris meeting in March 2006, 5438+0 and so on. The United States adopted the opinions of other member States and decided to develop a new generation of nuclear energy system with a broader concept. The development goal of the fourth generation nuclear power plant can be divided into four aspects.

The sustainable development of nuclear energy provides a means of sustainable energy production through the effective use of nuclear fuel; Minimize the amount of nuclear waste, strengthen management, reduce long-term management affairs, ensure public health and protect the environment.

Improve safety and reliability to ensure higher safety and reliability; Greatly reduce the probability and degree of core damage, and have the ability to quickly resume reactor operation; Cancel the necessity of taking emergency measures outside the factory.

Improving the cost of economic power generation is superior to other energy sources; The risk level of capital can be compared with other energy sources.

Nuclear proliferation prevention utilizes the characteristics of the reactor system itself, and the materials processed in the commercial nuclear fuel cycle have higher nuclear proliferation prevention ability, ensuring that it is difficult to be used in nuclear weapons or stolen; In order to evaluate the nuclear non-proliferation of nuclear energy, the Ministry of Energy is developing a quantitative evaluation method to prevent nuclear proliferation for the fourth generation nuclear power plants. In April of 20001year, the US Department of Energy collected 94 fourth-generation nuclear power plant reactor systems from 12 countries, including 28 water-cooled reactors, 32 liquid metal-cooled reactors, 17 gas-cooled reactors and 17 other reactor types.

At the GIF conference held in Tokyo from September/KLOC-0 to September 20, 2002, 10 countries unanimously agreed to develop the following six conceptual reactor systems for the fourth generation nuclear power plants based on the above 94 conceptual reactors.

(1) Air-cooled fast reactor system

The gas-cooled fast reactor (GFR) system is a fast neutron spectrum helium-cooled reactor, which adopts a closed fuel cycle, and the fuel can choose composite ceramic fuel. It uses a direct cycle helium turbine to generate electricity, or uses its process heat for thermochemical hydrogen production. Through the comprehensive utilization of fast neutron spectrum and the complete recovery of actinides, GFR can minimize the generation of long-lived radioactive waste. In addition, its fast neutron spectrum can also make use of existing fissile materials and convertible materials (including depleted uranium). The reference reactor is a 288 MW helium cooling system with an outlet temperature of 850℃.

(2) Lead alloy liquid metal cooling fast reactor system

Lead-cooled fast reactor (LFR) system is a liquid metal-cooled reactor with fast neutron spectrum of lead (lead/bismuth * * *). Closed fuel cycle is adopted to realize effective conversion of convertible uranium and control actinides. Fuel is a metal or nitride containing convertible uranium and transuranic elements.

The characteristic of LFR system is that it can choose from a series of rated power of power plants. For example, the LFR system can be a large-scale comprehensive power plant with 65,438+0,200 MW, or a combination of a modular system with a rated power of 300-200 MW and a battery pack with a long refueling interval of 50-65,438+0,000 MW (65,438+0.5-20 years). LFR battery pack is a turnkey power plant manufactured by a small factory, which can meet the market demand for small power grid power generation.

(3) molten salt reactor system

Molten salt reactor (MSR) system is a kind of superheated neutron spectrum reactor, and its fuel is a circulating liquid mixture of sodium, zirconium and uranium fluoride. Molten salt fuel flows through the graphite channel in the core, producing superheated neutron spectrum. The liquid fuel of MSR system does not need to make fuel elements, and actinides such as plutonium are allowed to be added. Actinides and most fission products will form fluoride in liquid coolant. Molten fluoride has good heat transfer characteristics, which can reduce the pressure on pressure vessels and pipelines. The power level of the reference power station is 1000 MW, and the outlet temperature of coolant is 700 ~ 800℃, with high thermal efficiency.

④ Liquid sodium cooled fast reactor system

The liquid sodium-cooled fast reactor (SFR) system is a fast neutron spectrum sodium-cooled reactor. The closed fuel cycle can effectively control the transformation of actinides and convertible uranium. SFR system is mainly used to manage high-level radioactive waste, especially plutonium and other actinides. The system has two main schemes: a medium-sized nuclear power plant with power of 150 ~ 500 MW, and the fuel is uranium-plutonium-actinide-zirconium alloy; Medium and large nuclear power plants, that is, nuclear power plants with a power of 500 ~ 1 500 MW, use uranium-plutonium oxide fuel.

The system has the characteristics of long thermal response time, large coolant boiling margin, the primary loop system running near atmospheric pressure, and an intermediate sodium system between radioactive sodium in the loop and water and steam in the power plant, so the safety performance is good.

⑤ Ultra-high temperature gas cooled reactor system

The ultra-high temperature reactor (VHTR) system is a graphite moderated helium cooled reactor with direct uranium fuel cycle. The reactor core can be a prismatic block core (such as Japanese HTTR core) or a pebble bed core (such as China HTR- 10 core).

The VHTR system provides heat, and the core outlet temperature is 1 000℃, which can produce hydrogen or process heat for petrochemical or other industries. Power generation equipment can also be added to the system to meet the needs of cogeneration. In addition, the system can flexibly use uranium/plutonium fuel cycle to minimize the amount of waste. The reference reactor uses 600 MW core.

(6) Supercritical water-cooled reactor system

Supercritical water-cooled reactor (SCWR) system is a high temperature and high pressure water-cooled reactor operating above the thermodynamic critical point of water (374℃, 22. 1 MPa). Supercritical water coolant can improve the thermal efficiency to about 1.3 times that of light water reactor. The characteristic of the system is that the coolant does not change its state in the reactor, and it is directly connected with the energy conversion equipment, so it can greatly simplify the supporting equipment of the power plant. The fuel is uranium oxide. There are two schemes for core design, namely thermal neutron spectrum and fast neutron spectrum. The reference system power is 1 700 MW, the operating pressure is 25 MPa, and the reactor outlet temperature is 5 10 ~ 550℃.