Reactors for low-temperature nuclear heat supply

Nuclear sources are not only covering more than 16% of today's electricity production but can also supply heat for district heating and industrial needs. Thus the nuclear generated heat substitutes for fossil fuels with good efficiency and economy and with much higher environmental cleanliness. Low-temperature nuclear heat is gained in several countries from the reactors of nuclear power plants by co-generation of heat and electricity which is already a proven technology. Specialized nuclear heating plants are in an early stage of development. The paper gives an overview of the situation worldwide and shows also specific common safety characteristics of these reactors.     

热管空间核反应堆电源的研究进展

随着人类对宇宙太空的深入探索,对于提供能量的电源要求也在逐步提高,空间核反应堆电源在执行深空探测任务中脱颖而出。热管核反应堆由于具备非能动性、寿期长、可靠性高等优势,成为目前空间核反应堆领域的研究热点。本文通过重点介绍典型热管堆的概念设计系统以及在Kilopower中的应用,对热管电源系统(Heatpipe Power System,HPS)、由热管控制的火星探索反应堆(The Heatpipe-operated Mars Exploration Reactor,HOMER)、安全可负担裂变引擎方案(Safe Affordable Fission Engine,SAFE)以及Kilopower进行了重点调研,归纳总结了各个反应堆的结构设计、燃料选择、热管排布、功率设计等,以期对未来热管空间核反应堆电源的设计研究提供思路和参考。     

The encapsulated nuclear heat source reactor for low-waste proliferation-resistant nuclear energy

The encapsulated nuclear heat source (ENHS) is a new Pb-Bi cooled modular reactor concept that features a combination of the following useful features that may make nuclear energy more attractive: (1) 20 years of full power operation without refueling. (2) Nearly constant fissile fuel contents and k_eff. (3) No on-site refueling and fueling hardware. (4) The ENHS modules are factory manufactured and transported already fueled to the site. (5) No access to neutrons. (6) No mechanical connections between the ENHS module and the energy conversion plant (The ENHS module has the function of a nuclear battery — with 20 years of full power operation at 125 MW_th). (7) At end of life, the ENHS module serves as a spent fuel storage cask and, later, as a spent fuel shipping cask. That is, the fuel is locked inside the ENHS from “cradle to grave”. (8) 100% natural circulation resulting in passive load following capability and autonomous control. This combination of features offers a highly safe nuclear energy system that is characterized by low waste, high proliferation resistance and high uranium utilization. The low waste and high uranium ore utilization are achieved by recycling the Pu and MA many times using a proliferation-resistant dry process; only fission products are to be extracted between cycles. Spent LWR fuel can provide for the HM make-up. The high level of proliferation resistance is obtained by restricting access to the fuel and neutrons and by eliminating the economic incentive of the client country to invest in sensitive technologies or infrastructure that can be used for clandestine production of strategic nuclear materials.     

Simple and safe deep pool reactor for low-temperature heat supply

The deep pool reactor (DPR) is a novel design of pool type reactor for heat only supply. The reactor core is put in a deep pool and it is within the low working temperature range. By only putting light hydrostatic pressure on the core coolant, the DPR will be able to meet the temperature requirements of heat supply for district heating and seawater desalination.

This paper will cover its design characteristics and main conclusions of safety analyses and economic evaluation. The nuclear heating system provides a safe feasibility and economically attractive solution for supplying low-temperature heat and reducing greenhouse gas emissions.     

Experimental simulation study on start-up of the 5 MW nuclear heating reactor

An experimental simulation study on the start-up of a low temperature, natural circulation nuclear heating reactor (5 MW developed by the Institute of Nuclear Energy of Tsinghua University, Beijing) is presented. The experiment was performed on the test loop (HRTL-5), which simulates the geometry and system design of the 5 MW reactor. The manifestation of different kinds of two-phase flow instability, namely geysering, flashing instability and low steam quality density wave instability on the start-up are described. The mechanism of flashing instability, which has never been well studied in this field, is especially interpreted. Based on the study of these instabilities, it is suggested that the start-up process, from initial condition to boiling operation condition, should consist of three steps: increasing of initial pressure by means of a noncondensable gas (N₂), start-up of the reactor at this pressurized condition (single-phase regime operation), and transition to a lower pressure, boiling operation. Three transition methods are discussed. As a result of these studies, the method of transition with low heat flux and low inlet subcooling is proposed. A stable start-up process of the 5MW reactor is achieved by careful selection of the thermohydraulic parameters.     

Utilization of high temperature heat from nuclear reactor using inorganic chemical heat pump

A new chemical heat pump designed to utilize high-temperature heat generated from high-temperature gas reactor (HTGR) is discussed. The calcium oxide / lead oxide / carbon dioxide reaction system was found to be a suitable reaction system for the desired heat pump from experimental survey of inorganic oxide / carbon dioxide reaction systems. The proposed heat pump using the reaction system was expected to be heat storage and heat transform system for HTGR. To demonstrate the validity of the heat pump, equilibrium relationship and kinetics of the reaction system was studied experimentally. The practical operation conditions of the heat pump were revealed from the experiment. This system was available to store heat above 800°C and transform it to higher temperature under a thermal driving condition. The heat output of the heat pump was valued enough compared to a common system. The applied system of the heat pump combined with HTGR was proposed to show the way of effective utilization of HTGR.     

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。     

Design concept of self-contained low power reactor master for heat supply

This paper demonstrates technical features and conceptual scheme of innovative self-contained low power reactor MASTER for heat supply. Neutron-physical and thermo-hydraulic characteristics of this reactor are analyzed. The possibility of power self-control and minimization of reactivity swing during fuel burnup are considered.     

Calculation of the magnitude of depressurization (loss of sealing) of the reactor vessels of nuclear heat supply plants (NHP)

Conclusions Thus, we suggested an engineering approach for evaluating the strength of cylindrical vessels containing cracks and for determining the approximate magnitude of the possible depressurization.Fatigue growth of a technological (an already existing) crack or a cracklike defect nucleated during the service period forms the principal mechanism determining the destruction process (damage) of the reactor vessels right up to the stage of loss of tightness.The detrimental effect of the factors such as radiation embrittlement, hydrogenation, or stress corrosion requires additional studies.In the case of the vessels of NHP, realization of the criterion of prefacture flow (and not catastrophic fracture) is expected, during crack growth through the reactor wall.Basically, depressurization appears to be highly improbable since growth of an elliptical crack from the initial sizes (0.1S) up to the final sizes (0.8S) is possible only when the number of loading cycles is large (N>10⁶).If cyclic crack growth occurs through the wall, the magnitude of depressurization of the reactor vessels of NHP is insignificant (M=1·10^–5 m²).The large reserve (safety factor) with respect to the number of cycles to realization of the criterion of prefacture flow=rs makes it possible to indicate, in advance, the loss of integrity of the reactor vessel using the available methods of nondestructive control and the small magnitude of depressurization forms a basis for working out the variations of localization of the consequences of the anticipated accident without compulsory use of complex and costly structures such as special (insured) vessels.Translated from Atomnaya Énergiya, Vol. 64, No. 2, pp. 87–90, February, 1988.