Research and development on nuclear hydrogen production using HTGR at JAERI

JAERI has been conducting R&D on HTGR and on hydrogen production using HTGR. The reactor technology has been developed using HTTR installed at Oarai site of JAERI. HTTR reached its full power operation of 30MW in 2001 and demonstrated reactor outlet helium temperature of 950°C in April 2004. As for the hydrogen production technology, the thermo-chemical IS process is under study. The process control method for continuous hydrogen production has been examined using a bench-scale apparatus. Also, studies are underway on process improvement and on materials of construction to be used in the corrosive environment. As for the system integration of HTGR and the hydrogen production plant, R&D is underway aiming to develop technologies for safe and economical connection. It covers safety technology against explosion, safety technology against radioactive materials release, control technology to prevent the thermal disturbance from hydrogen production plant to reactor, etc.     

Control strategies for transients of hydrogen production plant in VHTR cogeneration systems

Operability of Very High Temperature Reactor (VHTR) hydrogen cogeneration systems in response to abnormal transients initiated by the hydrogen production plant is one of the important concerns from economical and safety points of views. The abnormal events in the hydrogen production plant could initiate load changes and induce temperature variations in a primary cooling system. Excessive temperature increase in the primary cooling system would cause reactor scrams since the temperature increase in the primary cooling system is restricted in order to prevent undue thermal stresses from reactor structures. Also, temperature decrease has a potential propagation path for reactor scrams by reactivity insertions as a consequence of the reactivity feedbacks. Since suspensions of reactor operation and electricity generation should be avoided even in case of abnormal events in the hydrogen production plant from an economical point of view, an establishment of a control scheme against abnormal transients of hydrogen production plant is required for plant system design.In the present study, basic controls and their integration for the GTHTR300C, a VHTR cogeneration system designed by JAEA with a direct Brayton cycle power conversion unit and thermochemical Iodine-Sulfur process hydrogen production plant (IS hydrogen production plant), against abnormal transients of IS hydrogen production plant are presented. Transient simulations for selected load change events in the IS hydrogen production plants are performed by an original system analysis code which enables to evaluate major phenomena assumed in process heat exchangers of the IS hydrogen production plant.It is shown that abnormal load change events are successfully simulated by the system analysis code developed. The results demonstrated the technical feasibility of proposed controls for continuous operation of the reactor and power conversion unit against load change events in the IS hydrogen production plant.     

Safety aspects of the combined HTTR/steam reforming complex for nuclear hyrogen production

The High-Temperature Engineering Test Reactor (HTTR) in Oarai, Japan, has the potential to demonstrate the production of hydrogen by steam reforming and using nuclear process heat as primary energy input. Particular safety aspects for such a combined nuclear/chemical complex have been investigated such as fire and explosion hazard at presence of flammable gases (LNG, H₂, CO) near the reactor building. A methane vapor cloud in the open atmosphere or partially obstructed areas is highly unlikely to detonate and damage the reactor building. Theoretical assessments and experimental studies significant to the HTTR-steam reforming system, include the spreading and combustion behavior of cryogens and flammable gases providing the basis for a comprehensive safety analysis of the nuclear/chemical facility.     

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.     

Safety study of the coupling of a VHTR with a hydrogen production plant

The present paper deals with specific safety issues resulting from the coupling of a nuclear reactor (very high temperature reactor, VHTR) with a hydrogen production plant (HYPP). The first part is devoted to the safety approach consisting in taking into account the safety standards and rules dedicated to the nuclear facility as well as those dedicated to the process industry. This approach enabled two main families of events to be distinguished: the so-called internal events taking place in the coupling circuit (transients, breaks in pipes and in heat exchangers) and the external events able to threat the integrity of the various equipments (in particular the VHTR containment and emergency cooling system) that could result from accidents in the HYPP. By considering a hydrogen production by means of the iodine/sulfur (IS) process, the consequences of the both families of events aforementioned have been assessed in order to provide an order of magnitude of the effects of the incidents and accidents and also in order to propose safety provisions to mitigate these effects when it is necessary. The study of transients induced by a failure of a part of the HYPP has shown the possibility to keep the part of the HYPP unaffected by the transient under operation by means of an adapted regulation set. Moreover, the time to react in case of transfer of corrosive products in the VHTR containment has been assessed as well as the thermohydraulic loading that would experience the coupling pipes in case of very fast uncoupling of the facilities aiming at avoiding an excessive pressurization of the VHTR containment. Regarding the external events, by applying a method used in the process industries, the bounding representative scenarios have been identified on the basis of their consequences but also on the basis of their occurrence frequency. The consequences of the selected bounding scenarios, calculated taking into the source-term, the atmospheric dispersion and the pressure and toxic effects induced respectively by a hydrogen unconfined vapour cloud explosion (UVCE) and a sulfur dioxide release have been assessed. The resulting safety distance of about 100 m for the UVCE is fairly acceptable in terms of performance (head loss and thermal loss) of the coupling system. However, the longer safety distance (about 1.5 km) calculated for a SO₂ release implies to foresee a long distance to settle the control room of the site or to foresee provisions able to stop very fast the SO₂ leak.     

聚变高温制氢反应堆概念设计研究

在深入分析相关领域研究发展状况的基础上,提出了具有较好技术可行性的聚变高温制氢反应堆概念(称之为FDS-Ⅲ),包括具有先进等离子体物理和技术水平的聚变堆芯、先进高温锂铅包层(HTL)、可减少热流分布密度的"垂直靶板"偏滤器以及相应的功率转换系统。尤其是提出了HTL包层新概念,其特点是选用技术基础相对成熟的低活化铁素体/马氏体钢作结构材料,在锂铅流道中使用可耐高温的多层流道插件,实现约1000℃的出口温度,可应用于制氢。初步性能分析表明FDS-Ⅲ制氢堆及其包层概念具有较好的技术可行性。     

耦合高温堆的氦加热无机膜重整器的数值研究

基于拟均相一维模型,对耦合高温堆的氦加热无机膜重整器建立了一个稳态的数学模型,开发了相应的计算程序.结果表明:氦加热无机膜重整器的平均热流密度比常规的氦加热重整器高约25%,这样可实现紧凑的重整器结构,对于整个系统的安全性和经济性具有很重要的意义.利用氦加热无机膜重整器能获得高达95%的甲烷转化率,而压力损失只是略有增加. 当膜厚度比较小和吹扫比比较大时,甲烷转化率随着重整压力的增加而增大,这样高温堆甲烷蒸汽重整制氢系统的高压将由不利条件变为有利条件.     

Thermodynamic and Economic Analyses of HTGR Cogeneration System Performance at Various Operating Conditions for Proposing Optimized Deployment Scenarios

Several potential deployment scenarios of high-temperature gas reactor (HTGR) cogeneration systems for the simultaneous production of hydrogen and electricity have been proposed recently. They can operate under different conditions and thereby satisfy different hydrogen and electricity demand scenarios, but their performance must be studied to demonstrate thermodynamic feasibility and economic profitability to attract sufficient investment for their deployment. Therefore, this study uses exergy analysis and exergybased costing analysis methods to analyze HTGR cogeneration system performance thermodynamically and economically over the entire range of potential operating conditions by calculating performance indicators, exergy efficiency, and the specific cost per unit product. Furthermore, three optimized deployment scenarios with high performance for satisfying three different hydrogen demand scenarios are proposed based on the analysis results. The proposed three deployment scenarios show that the HTGR cogeneration system is thermodynamically efficient and economically competitive compared with other hydrogen and electricity generation systems. The feasibility of exergy analysis and exergy-based costing analysis methods for analyzing the HTGR cogeneration system so as to propose optimized deployment scenarios is demonstrated by the obtained findings practically.     

Study on efficiency of DCP for nuclear hydrogen production

With many advantages, hydrogen is considered as the fuel of the future. But there is no natural resource of hydrogen and it must be produced by other kinds of energy. As for the primary energy, nuclear energy is a promising alternative. Using heat from nuclear reactor to produce hydrogen is receiving more and more concerns in recent years. This paper mainly emphasizes the study of the direct contact pyrolysis （DCP） of methane using heat from nuclear reactor. A facility was designed to investigate the efficiency of DCP process in certain conditions. The experimental results show that this process produces only hydrogen and carbon. The conversion efficiency increases with temperature and residence time, but decreases as flow rate increases. The highest efficiency of DCP obtained in this exoedment is about 22%.     

Derivation of ideal power distribution to minimize the maximum Kernel migration rate for nuclear design of pin-in-block type HTGR

ABSTRACT

Suppressing the kernel migration rates under normal operation condition is quite important from the viewpoint of the fuel integrity for High-Temperature Gas-cooled Reactors. It depends on both the fuel temperature and the fuel temperature gradient. The presence of the ideal axial power distribution to minimize the maximum kernel migration rate allows us to improve the efficiency of design work. Therefore, we propose a new method based on Lagrange multiplier method in consideration of thermohydraulic design in order to obtain the ideal axial power distribution to minimize the maximum kernel migration rate. For one of the existing conceptual designs performed by JAEA, the maximum kernel migration rate for the power distribution to minimize the maximum kernel migration rate proposed in this study is lower by approximately 10% than that for the power distribution as a conventional design target to minimize the maximum fuel temperature.     

Hydrogen production by high temperature electrolysis with nuclear reactor

High Temperature Electrolysis (HTE) is a promising method because its most parts consist of environmentally sound and common materials. Hydrogen production efficiency of HTE was evaluated about the process coupling with high temperature gas cooled reactor. This process can be expected to accomplish over 53% hydrogen production efficiency at HTE operating temperature of 800 °C. As a demonstration of hydrogen production by HTE, a unit housing 15 tubular cells, where yttria-stabilized zirconia (YSZ) was used as electrolyte, was constructed, and accomplished 130 NL/h hydrogen production. In this experiment, measured hydrogen production rate has good agreement with calculated hydrogen production rate based on applied current. To design and construct large amount of hydrogen production unit, it is important to predict the thermal and electrochemical features of the unit. To predict them, the simulation technology has been developed. From the comparison between single tubular cell experimental result and simulation result, good agreement based on current–voltage characteristic was acquired.     

Numerical Study on Tritium Behavior by Using Isotope Exchange Reactions in Thermochemical Water-Splitting Iodine—Sulfur Process

One potential problem in the hydrogen production system coupled with the high-temperature gascooled reactor (HTGR) is transmission of tritium from the primary coolant to the product hydrogen by permeation through the heat transfer tubes. Tritium accumulation in the process chemicals in the components of a hydrogen plant, a thermochemical water-splitting iodine-sulfur (IS) process, will also be a critical issue in seeking to license the hydrogen plant as a non-nuclear plant in the future. A numerical analysis model for tritium behavior in the IS process was developed by considering the isotope exchange reactions between tritium and the hydrogen-containing process chemicals, i.e., H₂O, H₂SO₄ and HI. The tritium activity concentration in the IS process coupled with the high-temperature engineering test reactor (HTTR), the HTTR-IS system, was preliminarily evaluated in regard to the effects of some indeterminate parameters, i.e., equilibrium constants of the isotope exchange reactions, permeability of tritium through heat transfer tubes, tritium and hydrogen concentrations in the secondary helium coolant, and the leak rate from the secondary coolant loop. The results describing how the tritium activity concentration changes with variations in these parameters and which component has the maximum tritium activity concentration in the IS process are described in this paper.     

European passive plant program: A design for the 21st century

In 1994, a group of European utilities initiated, together with Westinghouse and its industrial partner GENESI (an Italian consortium including ANSALDO and FIAT), a program designated EPP (European Passive Plant) to evaluate Westinghouse passive nuclear plant technology for application in Europe. The following major tasks were accomplished: (1) the impacts of the European utility requirements (EUR) on the Westinghouse nuclear island design were evaluated; and (2) a 1000 MWe passive plant reference design (EP1000) was established which conforms to the EUR and is expected to be licensable in Europe. With respect to safety systems and containment, the reference plant design closely follows that of the Westinghouse simplified pressurized water reactor (SPWR) design, while the AP600 plant design has been taken as the basis for the EP1000 reference design in the auxiliary system design areas. However, the EP1000 design also includes features required to meet the EUR, as well as key European licensing requirements.     

Analysis of Tritium Behavior in Very High Temperature Gas-Cooled Reactor Coupled with Thermochemical Iodine-Sulfur Process for Hydrogen Production

The tritium concentration in the hydrogen product in Japan's future very high temperature gas-cooled reactor (VHTR) system coupled with a thermochemical water-splitting iodine-sulfur (IS) process (VHTRIS system), named GTHTR300C, was estimated by numerical analysis. The tritium concentration in the hydrogen product significantly depended on undetermined parameters, i.e., the permeabilities of a SO₃ decomposer and a H₂SO₄vaporizer made of SiC. Thus, the estimated tritium concentration in the hydrogen product for the conservative analytical condition ranged from 3.4 × 10^?3 Bq/cm³ at STP (38 Bq/g-H₂) to 0.18 Bq/cm³ at STP (2,000 Bq/g-H₂). By considering the tritium retained by core graphite and the reduction in permeation rate by an oxide film on the heat transfer tube of the IHX and the HI decomposer, the tritium concentration in the hydrogen product decreased to the range from 3.3 × 10^?5 Bq/cm³ at STP (0.36 Bq/g-H₂) to 5.6 × 10^?3 Bq/cm³ at STP (63 Bq/g-H₂), which were smaller than those for the conservative analytical condition by factors of about 3.2 × 10^?2 and 9.6 × 10^?3, respectively. The effectof the helium flow rate in the helium purification system on the tritium concentration in the hydrogen product was also evaluated.     

聚变制氢堆高温液态包层热工水力学新概念研究

在深入分析聚变堆包层设计要求和目前技术发展水平的基础上,根据热化学工艺制氢需要高温热的要求,提出了一个基于技术相对成熟的低活化铁素体/马氏体钢作为主要结构材料、高压氦气与液态LiPb合金作为冷却剂、具有创新性“多层流道插件”结构方案以获得高温热能的包层热工水力学概念,建立了热工水力学模型,在利用有限元数值模拟程序进行模拟计算的基础上分析了这种新概念包层的可行性。     

Cost evaluation for centralized hydrogen production

Nuclear energy can provide heat and electricity to produce hydrogen. Thermo-chemical and electrical decomposition of water have been studied as a hydrogen source. In this study, the cost of hydrogen supply for transportation usage was evaluated. The total cost for a centralized hydrogen production consisted of production cost, delivery cost, and station cost. The total cost for hydrogen production using nuclear energy can be at least comparable to that of steam-methane reforming, if the cost of carbon dioxide fixation was included. The delivery cost can be reduced by optimizing the size of hydrogen production and delivery distances. The hydrogen station cost was found out to be about 50% of the hydrogen supply cost. The optimum thermal power of nuclear power plants for hydrogen production was estimated based on the cost evaluation.     

核动力装置的基本安全原则

要确保核动力装置的安全运行,必须严格执行安全法规和建立本国的核安全法规体系,并且遵循行之有效的核动力装置的基本安全原则。本文扼要地介绍了核法规的重要意义、经济效果以及各国多年实践所积累的基本安全原则.     

核电厂安全级仪表在线监测系统技术研究

目前国内核电厂主要采取定期校准的方式对安全级仪表漂移进行管理,但该方法过于保守且经济性差。基于此,本文对安全级仪表在线监测系统技术进行了研究,首先对安全级仪表实际漂移数据进行了分析,明确了核电厂安全级仪表漂移的主要类型,证明了对安全级仪表开展在线监测的可行性。其次,通过对相关法规及标准的分析和研究,明确了核电厂安全级仪表在线监测技术的基本要求。最后,开展了在线监测系统技术的数据分析研究,对冗余仪表提出了等价平均算法,对非冗余仪表算法进行了分析并对多元状态估计模型(MSET)方法开展了基于电厂实际数据的建模验证,证明了该方法在核电厂应用的可行性。     

核电厂设备抗震鉴定的审评

介绍了核电厂设备抗震鉴定的标准和要求，鉴定的范围、步骤、方法和程序，以及抗震鉴定审评所依据的准则。详细论述了抗震试验鉴定中所采用的具体实施程序，并通过审评实践对核电厂设备抗震鉴定中常见的一些问题也进行了探讨。