High-temperature gas-cooled reactors — energy source for industrial production of hydrogen

High-temperature helium-cooled reactors are the best understood nuclear technology that can supply high-temperature heat for thermal processes for producing hydrogen. The GT-MGR reactor — an innovative international modular design of a helium-cooled reactor with a gas-turbine cycle — best meets the requirements for hydrogen production and is proposed as a basis for a nuclear energy source. In this paper, the technical aspects of the proposed application of HTGR as a source of energy for producing hydrogen are analyzed. The required parameters of the energy obtained from HTGR for the presently completed and future hydrogen-production technology are examined. The problems and additional R&D work on the use of HTGR at high helium temperatures are indicated.Translated from Atomnaya Énergiya, Vol. 97, No. 6, pp. 432–446, December, 2004.This revised version was published online in April 2005 with a corrected cover date.     

Evaluation of high temperature gas reactor for demanding cogeneration load follow

Modular nuclear reactor systems are being developed around the world for new missions among which is cogeneration for industries and remote areas. Like existing fossil energy counterpart in these markets, a nuclear plant would need to demonstrate the feasibility of load follow including (1) the reliability to generate power and heat simultaneously and alone and (2) the flexibility to vary cogeneration rates concurrent to demand changes. This article reports the results of JAEA's evaluation on the high temperature gas reactor (HTGR) to perform these duties. The evaluation results in a plant design based on the materials and design codes developed with JAEA's operating test reactor and from additional equipment validation programs. The 600 MWt-HTGR plant generates electricity efficiently by gas turbine and 900°C heat by a topping heater. The heater couples via a heat transport loop to industrial facility that consumes the high temperature heat to yield heat product such as hydrogen fuel, steel, or chemical. Original control methods are proposed to automate transition between the load duties. Equipment challenges are addressed for severe operation conditions. Performance limits of cogeneration load following are quantified from the plant system simulation to a range of bounding events including a loss of either load and a rapid peaking of electricity.     

安全经济高效的先进能源

具有第四代安全经济特性的核电应该是人们期待的先进的清洁低碳能源。高温气冷堆是当今研发的第四代核电堆型之一,但现有的设计还存在需要排除的严重的安全隐患。堆芯不熔化,不等于说不会有严重事故发生。需要吸取国外球床高温堆和柱状高温堆两种实验堆型运行的经验教训、扩展安全观念和应对安全低概率事件,确保反应堆不出现后果极其严重的放射性释放事故。当热电转换系统采用与燃气蒸汽联合循环耦合应用的技术以后,会发挥高温堆所长,更大地提升转换效率,形成一种高安全低投资和高效率的双燃料清洁能源,可用于大堆或小堆的应用环境,可满足电力系统基本负荷和调锋负荷的需要。在工程设计上采取一系列改进和创新措施,包括釆用规则床模块化及地下反应堆设计以后,可在提高反应堆核心部位安全防卫能力的同时,防范低概率事件,成为一种新的安全经济高效的先进能源。     

核能制氢技术的发展

氢是清洁能源,有非常好的应用前景.但氢是二次能源,需要利用一次能源来生产.以可持续的方式(原料来源丰富、无温室气体排放)实现氢的大规模生产是实现氢广泛利用的前提.核能是清洁的一次能源,核电已经成为世界电力生产的主要方式之一.正在研发的第四代核能系统除了要使核电生产更经济和更安全之外,还要为实现核能在发电之外的领域的应用...     

Materials and design concepts of gas-cooled reactor systems

The family of gas-cooled reactors being developed in the United States by Gulf General Atomic consists of the steam-raising and direct cycle versions of the high temperature gas-cooled reactor (HTGR) for electric power generation, the hydrogen-producing HTGR for chemical process applications, and the gas-cooled fast reactor (GCFR), a high gain breeder. The aim of this paper is to describe the underlying design concepts that are common to all of these reactors and relate these design concepts to the choice of both structural and fuel materials for the wide variety of environmental conditions encountered throughout the world. Interwoven with this discussion are typical examples of the interaction of design activities and materials selection required to give a reactor system of maximum safety and reliability, favourable environmental features, and minimum cost.     

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.     

Feasibility study on the applicability of a diffusion-welded compact intermediate heat exchanger to next-generation high temperature gas-cooled reactor

The development of an intermediate heat exchanger (IHX) transferring high temperature heat to a process heat application is of prime importance for a next-generation high temperature gas-cooled reactor (HTGR). The IHX needs high structural integrity and reliability over 900°C for a long duration. A plate fin type compact heat exchanger (PFCHX) has a large heat transfer area per heat exchanger volume and is expected to be used as the IHX. However, the brazing for connecting fins and plate is not reliable when existing PFCHXs are used in a high temperature condition for a long time. We have proposed a concavo-convex plate type compact heat exchanger (CPCHX) which consists of concavo-convex plates (CPs) welded by solid state diffusion and made of nickel-based superalloy Hastelloy XR. In our study, first, an optimized condition for the solid state diffusion welding between the CPs of the CPCHX was found by experiments using test pieces made of Hastelloy XR. Second, small-scale diffusion-welded CPCHXs were designed, manufactured and installed in a test loop to investigate the reliability of the diffusion welding. As a result of leakage tests, it was confirmed that the reliability of the solid state diffusion welding between the CPs of the small-scale CPCHX is sufficient. A thermal performance test revealed that the thermal conductance of the small-scale CPCHX was better than calculated. In addition, a design study for the CPCHX was performed to investigate the feasibility of the diffusion-welded CPCHX to the IHX in a next-generation HTGR.     

Small high temperature gas-cooled reactors with innovative nuclear burning

Since the innovative concept of CANDLE (Constant Axial shape of Neutron Flux, nuclide densities and power shape During Life of Energy producing reactor) burning strategy was proposed, intensive research works have been continuously conducted to evaluate the feasibility and the performance of the burning strategy on both fast and thermal reactors. We learned that one potential application of the burning strategy for thermal reactors is for the High Temperature Gas-Cooled Reactors (HTGR) with prismatic/block-type fuel elements. Several characteristics of CANDLE burning strategy such as constant reactor characteristics during burn-up, no need for burn-up reactivity control mechanism, proportionality of core height with core lifetime, sub-criticality of fresh fuel elements, etc. enable us to design small sized HTGR with a high degree of safety, easiness of operation and maintenance, and long core lifetime which are required for introducing the reactors into remote areas or developing countries with limited infrastructures and resources. In the present work, we report our evaluation results on small sized block-type HTGR designs with CANDLE burning strategy and compared with other existing small HTGR designs including the ones with pebble fuel elements, under both uranium and thorium fuel cycles.     

Nuclear powered heat pumps for near-term process heat applications

There is a substantial market for nuclear energy in non-electric applications such as hydrogen production or water desalination. Among the Generation IV reactor concepts, the very high temperature reactor (VHTR) with a reactor outlet temperature close to 1000 °C and a power conversion efficiency of approximately 50% is believed to be the most suitable concept for co-generation of process heat. Its high coolant exergy would enable centralized hydrogen production and other process heat applications. In this paper it is shown that a reactor with lower coolant outlet temperature or another near-term heat source can also meet the VHTR objectives which are high power conversion efficiency and capability to deliver high temperature process heat in the narrow temperature window required by thermochemical hydrogen production cycles. The approach was to separate the requirement for high temperature process heat production from the nuclear part of the plant, in other words the nuclear part of the power plant would run at acceptably low temperature while the high temperature heat production via a heat pump system would be limited to a conventional external circuit, thus avoiding nuclear constraints. The separation of these high temperature constraints from the reactor would avoid massive R&D requirements on materials, components and fuel with uncertain outcome thus unnecessarily delaying introduction of this otherwise very attractive reactor concept.We then show that the proposed technology is equally suitable for the generation of cold (e.g. for air conditioning) and for desalination of seawater.     

Calcium oxide/carbon dioxide reactivity in a packed bed reactor of a chemical heat pump for high-temperature gas reactors

The thermal performance of a chemical heat pump that uses a calcium oxide/carbon dioxide reaction system was discussed as a heat storage system for utilizing heat output from high temperature gas reactors (HTGR). Calcium oxide/carbon dioxide reactivity for the heat pump was measured using a packed bed reactor containing 1.0 kg of reactant. The reactor was capable of storing heat at 900 °C by decarbonation of calcium carbonate and generating up to 997 °C by carbonation of calcium oxide. The amount of stored heat in the reactor was 800–900 kJ kg⁻¹. The output temperature of the reactor could be controlled by regulating the carbonation pressure. The thermal storage performance of the reactor was superior to that of conventional sensible heat storage systems. A heat pump using this CaO/CO₂ reactor is expected to contribute to thermal load leveling and to realize highly efficient utilization of HTGR output due to the high heat storage density and high-quality temperature output of the heat pump.     

高温和超高温气冷堆动力转换方案研究

动力转换单元是高温和超高温气冷堆的重要组成部分。本文对高温和超高温气冷堆的动力转换单元进行研究。从4个关键参数（反应堆出口温度、反应堆入口温度、压缩比和主蒸汽参数）入手，对5个循环方案进行比较分析。综合考虑各种工程因素，上位循环为简单氦气透平循环、下位循环为有再热的蒸汽轮机循环的联合循环方案是具有竞争力的，其中下位循环在高温气冷堆范围是亚临界参数循环，在超高温气冷堆范围是超临界参数循环。联合循环可实现高温和超高温气冷堆热量的高效率转化，且反应堆入口温度在反应堆压力壳材料允许的范围内，具有足够的安全性。     

Leak rate calculation for LBB analysis in high temperature gas-cooled reactors

Leak rate calculation is very important for Leak Before Break (LBB) analysis. Helium is used as coolant in high temperature gas-cooled reactor (HTGR). Therefore the flows in the cracks of HTGR vessels and pipes are single phase, which are different from the two phase critical flows in the cracks of water reactors. In the present paper, simple leak rate calculation formulae for compressible laminar and turbulent flows in HTGR cracks are introduced. The velocity and pressure distributions in cracks as well as the leak rates are calculated using the formulae. Numerical simulations are also conducted for compressible laminar, turbulent and critical flows with different crack widths and depths. The results of the numerical simulation and theoretical formulae are compared with experimental data. The comparison shows that both the simple theoretical formulae and the numerical simulation can achieve good results.     

Development of high temperature gas-cooled reactor (HTGR) fuel in Japan

This paper describes experiences and present status of research and development works for the high temperature gas-cooled reactor (HTGR) fuel in Japan. Recently, Very High Temperature Reactor (VHTR) is evaluated highly worldwide, and is a principal candidate for the Generation IV reactor systems. In Japan, HTGR fuel fabrication technologies have been developed through the High Temperature Engineering Test Reactor (HTTR) project in Japan Atomic Energy Agency since 1960’s. In total about 2 tons of uranium of the HTTR fuel has been fabricated successfully and its excellent quality has been confirmed through the long-term high temperature operation. Based on the HTTR fuel technologies, SiC TRISO fuel has been newly developed for burnup extension targeted VHTR. For ZrC-TRISO coated fuel as an advanced fuel designs, R&Ds for fabrication and inspection have been carried out in JAEA. The irradiation with the Japanese uniform stoichiometric ZrC coating has been completed in the cooperation with Oak Ridge National Laboratory of the United States.     

Present status of the High Temperature Engineering Test Reactor (HTTR)

In Japan, the research and development on the High Temperature Gas-cooled Reactors (HTGRs) had been carried out for more than fifteen years since 1969 as the multi-purpose Very High Temperature gas-cooled Reactor (VHTR) program for direct utilization of nuclear process heat such as nuclear steel making. Recently, reflecting the change of the social and energy situation and with less incentives for industries to introduce such in the near future, the JAERI changed the program to a more basic ‘HTTR program’ to establish and upgrade the HTGR technology basis.The HTTR is a test reactor with a thermal output of 30 MW and reactor outlet coolant temperature of 950°C, employing a pin-in-block type fuel block, and has the capability to demonstrate nuclear process heat utilization using an intermediate heat exchanger. Since 1986 a detailed design has been made, in which major systems and components are determined in line with the HTTR concept, paying essential considerations into the design for achieving the reactor outlet coolant temperature of 950°C. The safety review of the Government started in February 1989. By request of the Science and Technology Agency the Reactor Safety Research Association reviewed the safety evaluation guideline, general design criteria, design code and design guide for the graphite and the high-temperature structure of the HTTR.The installation permit of the HTTR was issued by the Government in November 1990.     

Present status of energy in Japan and HTTR project

An amount of primary energy supply in Japan is increasing year by year. Much energy such as oil, coal and natural gas is imported so that the self-sufficiency ratio in Japan is only 20% even if including nuclear energy. An amount of energy consumption is also increasing especially in commercial and resident sector and transport sector. As a result, a large amount of greenhouse gas was emitted into the environment. Nuclear energy plays the important role in energy supply in Japan.Japan Atomic Energy Research Institute (JAERI) has been carried out research and development of a hydrogen production system using a high temperature gas cooled reactor (HTGR). The HTTR project aims at the establishment of the HTGR hydrogen production system. Reactor technology of the HTGR, hydrogen production technology with thermochemical water splitting process and system integration technology between the HTGR and a hydrogen production plant are developed in the HTTR project.     

Above-ground HTGR design for passive decay heat removal

The design of a small high-temperature gas-cooled reactor (HTGR) for passive decay heat removal which could be located deeply underground was proposed previously. In the present work, analogue design analyses of passive decay heat removal for an above-ground long-life small prismatic HTGR was carried out to obtain the conditions for successful decay heat removal by radiation and conduction inside the reactor building, and by radiation and natural cooling by air at the outer surface of the reactor building. Sensitivity analysis of the peak temperatures of both the core and the reactor building after reactor shutdown was performed by changing the physical characteristics of the reactor regions. Enlarging the reactor building was found to be an effective way to reduce the peak reactor building temperature to within its design limit. By using the obtained condition for design parameters, the appropriate sizes of reactor core and reactor building were evaluated for some reactors. Consequently, criticality and burnup analyses for the proposed reactors were performed to confirm the possibility of designing a long-life core for the core size and reactor power which meet the condition of removing decay heat successfully. Using our design, all the reactors with 20 wt% uranium enrichment could be critical for over nine years.     

Economic viability of small to medium-sized reactors deployed in future European energy markets

Future plans for energy production in the European Union as well as other locations call for a high penetration of renewable technologies (20% by 2020, and higher after 2020). The remaining energy requirements will be met by fossil fuels and nuclear energy. Smaller, less-capital intensive nuclear reactors are emerging as an alternative to fossil fuel and large nuclear systems. Approximately 50 small (<300 MWe) to medium-sized (<700 MWe) reactors (SMRs) concepts are being pursued for use in electricity and cogeneration (combined heat and power) markets. However, many of the SMRs are at the early design stage and full data needed for economic analysis or market assessment is not yet available. Therefore, the purpose of this study is to develop “target cost” estimates for reactors deployed in a range of competitive market situations (electricity prices ranging from 45-150 €/MWh). Parametric analysis was used to develop a cost breakdown for reactors that can compete against future natural gas and coal (with/without carbon capture) and large nuclear systems. Sensitivity analysis was performed to understand the impacts on competitiveness from key cost variables. This study suggests that SMRs may effectively compete in future electricity markets if their capital costs are controlled, favorable financing is obtained, and reactor capacity factors match those of current light water reactors. This methodology can be extended to cogeneration markets supporting a range of process heat applications.     

Integrity assessment of the high temperature engineering test reactor (HTTR) control rod at very high temperatures

The high temperature engineering test reactor (HTTR) is the first high temperature gas-cooled reactor (HTGR) in Japan with a reactor outlet coolant temperature of 950°C at high temperature test operation. The HTTR contains 16 pairs of control rods for which Alloy 800H is chosen of the metallic parts. Because the maximum temperature of the control rods reaches about 900°C at reactor scrams, structural design guideline and design material data on Alloy 800H are needed for the high temperature design. The design guideline for the HTTR control rod is based on ASME Code Case N-47-21. Design material data is also determined and shown in this paper. Under the guideline, temperature and stress analysis was conducted, and it is confirmed that the target life of the control rods of 5 years can be achieved.     

Small PWRs using coated particle fuel for district heating, PFPWR50

It has been said that nuclear energy is an important option for especially developing countries to satisfy their increasing energy demand. However, it will be difficult to deploy first of a kind nuclear power plant in developing countries because extensive safety demonstration has to be conducted in industrialized countries. On the other hand, it will be essential to present rigid proof of reliable operational experience to develop proper understanding of the safety features of new reactor systems among the people around the demonstration plant sites. One of the ways to solve the issue is to integrate existing technologies supported by a great deal of data and experience into a new reactor design. Based on the consideration, a small-sized district heating reactor system based on the pressurized water reactor (PWR) technologies combined with the fuel concept of high temperature gas cooled reactors (HTGRs) has been studied. The purpose of the combination of these two existing concepts is to take the best advantages of both excellent operational experience of PWRs and the integrity of HTGR fuel, coated particle fuel, against fission products release even at high temperature. We expect that this approach will help create a breakthrough to the current stagnation of nuclear power deployment.     

高温气冷堆用碳毡材料导热系数测量及反问题计算

碳纤维材料已成为核能、航天等领域不可或缺的重要功能材料,在高温气冷堆及其相关实验中需要使用大量碳纤维保温材料。但由于目前测试方法的限制,相关材料物性参数测量数据严重不足,尤其是缺乏高温1000℃以上的热物性参数,致其使用受到限制。为此,清华大学核能与新能源技术研究院研制了模拟高温气冷堆温度、环境氛围的材料测试装置,可提供1600℃以下的材料性能测试。根据该装置一次典型实验过程的测量数据,详细介绍了采用非线性导热反问题方法确定材料温度相关导热系数的完整过程和具体算法。提出了一种依据稳态、非稳态热传导原理求解反问题的简明算法,该方法既可单独使用,也可为其他反问题算法提供良好的迭代初值。实验确定了高温气冷堆用碳毡保温材料在1600℃以下的导热系数,将为高温气冷堆相关实验和其他特高温条件下的应用提供重要参考。