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.     

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.     

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.     

小型高温气冷堆的堆型方案比较

本文对模块式高温气冷堆的棱柱状和球床两种堆芯型式和一体化与肩并肩分置式两种总体设计方案分别进行了技术特点、设计制造、运行经验和安全性与经济性的比较,提出了在我国发展高温气冷堆的堆型选用原则和建议.     

高温气冷堆在我国应用的预可行性研究摘要

本文概述了高温气冷堆在我国核能发展中的地位和作用,并对国内应用高温气冷堆的市场做了初步调查和分析,同时,做了小型高温气冷堆核电站 HTR100 型的初步适应性设计和经济估算。结果表明,在我国建造小型高温堆核电站在技术上是可行的。在一些地区,其发电成本可与同规模的煤电厂相接近。     

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.     

keff uncertainty quantification and analysis due to nuclear data during the full lifetime burnup calculation for a small-sized prismatic high temperature gas-cooled reactor

To benefit from recent advances in modeling and computational algorithms,as well as the availability of new covariance data,sensitivity and uncertainty analyses are needed to quantify the impact of uncertain sources on the design parameters of small prismatic high-temperature gas-cooled reactors(HTGRs).In particular,the contribution of nuclear data to the keff uncertainty is an important part of the uncertainty analysis of small-sized HTGR physical calculations.In this study,a small-sized HTGR designed by China Nuclear Power Engineering Co.,Ltd.was selected for keff uncertainty analysis during full lifetime burnup calculations.Models of the cold zero power(CZP)condition and full lifetime burnup process were constructed using the Reactor Monte Carlo Code RMC for neutron transport calculation,depletion calculation,and sensitivity and uncertainty analysis.For the sensitivity analysis,the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Track length importance Characterization(CLUTCH)method was applied to obtain sensitive infor-mation,and the"sandwich"method was used to quantify the keff uncertainty.We also compared the keff uncertainties to other typical reactors.Our results show that 235U is the largest contributor to keff uncertainty for both the CZP and depletion conditions,while the contribution of 239Pu is not very significant because of the design of low discharge burnup.It is worth noting that the radioactive capture reaction of 28Si significantly contributes to the keff uncer-tainty owing to its specific fuel design.However,the keff uncertainty during the full lifetime depletion process was relatively stable,only increasing by 1.12％owing to the low discharge burnup design of small-sized HTGRs.These numerical results are beneficial for neutronics design and core parameters optimization in further uncertainty prop-agation and quantification study for small-sized HTGR.     

keff uncertainty quantification and analysis due to nuclear data during the full lifetime burnup calculation for a small-sized prismatic high temperature gas-cooled reactor

To benefit from recent advances in modeling and computational algorithms,as well as the availability of new covariance data,sensitivity and uncertainty analyses are needed to quantify the impact of uncertain sources on the design parameters of small prismatic high-temperature gas-cooled reactors(HTGRs).In particular,the contribution of nuclear data to the keff uncertainty is an important part of the uncertainty analysis of small-sized HTGR physical calculations.In this study,a small-sized HTGR designed by China Nuclear Power Engineering Co.,Ltd.was selected for keff uncertainty analysis during full lifetime burnup calculations.Models of the cold zero power(CZP)condition and full lifetime burnup process were constructed using the Reactor Monte Carlo Code RMC for neutron transport calculation,depletion calculation,and sensitivity and uncertainty analysis.For the sensitivity analysis,the Contribution-Linked eigenvalue sensitivity/Uncertainty estimation via Track length importance Characterization(CLUTCH)method was applied to obtain sensitive infor-mation,and thesandwichmethod was used to quantify the keff uncertainty.We also compared the keff uncertainties to other typical reactors.Our results show that 235U is the largest contributor to keff uncertainty for both the CZP and depletion conditions,while the contribution of 239Pu is not very significant because of the design of low discharge burnup.It is worth noting that the radioactive capture reaction of 28Si significantly contributes to the keff uncer-tainty owing to its specific fuel design.However,the keff uncertainty during the full lifetime depletion process was relatively stable,only increasing by 1.12％owing to the low discharge burnup design of small-sized HTGRs.These numerical results are beneficial for neutronics design and core parameters optimization in further uncertainty prop-agation and quantification study for small-sized HTGR.     

高温气冷堆(HTGR)釷、鈾燃料元件后处理萃取流程实验研究

针对HTGR钍、铀燃料元件高燃耗、~(232)U含量高的特点提出了酸式进料的单循环溶剂萃取流程,并进行了串级实验。钍、铀收率达到>99.6％,钍、铀产品对Cs,Sr,Zr—Nb,Ru的去污满足远距离操作条件下再制造核燃料元件的要求。     

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.     

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.     

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.     

对我国核临界安全工作的思考

对我国的核临界安全工作进行全面思考,肯定了４０多年来我国核燃料循环生产系统的安全业绩,指出可能导致核临界事故的着急环节及存在问题,对于如何进一步加强核临界安全工作,从多角度提出对策。     

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.     

核安全文化的发展与应用

安全文化已对核能企业的安全性产生了重大影响。本文分析了核安全文化产生的背景，介绍了核安全文化在一些国家和组织应用发展的状况，提出了推行安全文化过程中应注意的 上问题，讨论了评价安全文化绩效的原则。     

世界核电发展趋势与高温气冷堆

核能的发展面临经济竞争力、核安全、核废物的最终处置及防止核武器材料扩散的挑战。为改善公众的可接受性 ,核电厂的安全性进一步改进。电力市场体制的非管制化改革加剧了电力技术的竞争。环境保护意识增强使核废物的处置倍受关注。 80年代中期以来发展的先进轻水堆核电厂如ABWR ,System 80 ,EPR ,AP60 0等是今后一段时期内商用核电的主力堆型。进入 2 0 0 0年之际 ,美国能源部正在规划发展第四代先进核能系统 ,目标是在 2 0 2 0年或之前 ,向市场提供经过验证的成熟的第四代核电厂技术 ,以替代美国退役的核电容量。球床高温气冷堆被认为是第四代先进核能系统的优选技术。南非ESKOM电力公司选择了球床高温气冷堆作为今后核电发展的堆型。清华大学承担设计和建设的 10MW高温气冷实验堆计划在 2 0 0 0年内临界。通过10MW高温气冷堆的建造 ,我国已形成了高温气冷堆技术的自主知识产权 ,初步具备了自主设计、制造和建造的能力     

低温核供热堆热网回路的辐射安全评价

本文主要介绍低温核低热堆为防止放射性物质进入热网回路的安全措施，在正常和事故工况下热网回路水中的放射性水平及其对用户所受最大可能剂量的估计。结果表明，即使在发生概率极低的最严重事故条件下，热网回路水最大可能达到的放射性核素浓度也比天然水的本底浓度低１个量级，足以保证广大用户的安全。     

世界核电史三次严重事故的反思

1979年美国三哩岛核事故、1986年前苏联切尔诺贝利核事故和2011年日本福岛核事故是世界核电发展史上三次严重核事故。前事不忘,后事之师。从核事故中总结吸取经验和教训,有利于进一步加强核安全,更好地发展和利用核能为民造福。     

切尔诺贝利事故及其影响与教训

本文从核安全与辐射防护的角度出发,根据几年来国际的研究与报道以及现场访问所了解的情况,对前苏联切尔诺贝利核电站发生的原因、影响及其教训进行了简要回顾。     

核电厂应急计划中若干问题和概念的讨论

本文讨论了我国核电厂应急计划和应急准备中新所关心的若干问题,阐述了应急计划中一些基本概念;弄清这些问题和概念,对推动和恰当地进行应急准备可能是重要的。