Evaluation and optimization of tritium breeding,shielding and nuclear heating performances of the helium cooled solid breeder blanket for CFETR

Chinese Fusion Engineering Test Reactor (CFETR) is a new test tokamak device being designed in China aimed to bridge gaps between ITER and DEMO. As one of the candidate tritium breeding blankets, a conceptual design scheme of the helium cooled solid breeder blanket has been proposed and a series of preliminary analyses have been carried out to access the performances. However, both the required global tritium breeding ratio for CFETR not less than 1.2 and its poor working conditions under intense radiation need further thorough neutronics analyses and optimizations during the design phase. In this work, first, three-dimensional neutronics model of CFETR was built, and the neutron wall loading and global TBR were obtained. The nuclear and thermal calculation results were automatically coupled, which could make the neutronics calculation results more accurate and guaranteed they could always satisfy the corresponding thermal limits during the whole process. Then, the tritium breeding and shielding performances of both the outboard and inboard equatorial blanket modules were optimized for the comprehensive optimal schemes. The influences of Be/W armors on the shielding performance and TBR were also investigated. Finally, the nuclear heating rates and the neutron flux densities in different components were calculated based on the obtained comprehensive optimal scheme. In this paper, the neutronics analyses and optimizations verified that the optimized conceptual design could well meet the tritium self-sufficiency and neutron shielding requirements, and this could provide a valuable reference for the further thermal-hydraulic analysis and structural optimization of the CFETR helium cooled solid breeder blanket.     

A new method for improving the tritium breeding and releasing performance of China Fusion Engineering Test Reactor phase II helium-cooled ceramic breeder blanket

China Fusion Engineering Test Reactor (CFETR) is a brand new Tokamak reactor which is currently being designed to fill the intervals between ITER and future DEMO fusion reactor. It has two operation phases: phase I with P_f = 200 MW is to demonstrate steady-state operation; phase II with P_f = 1000 MW to validate DEMO technology. Helium-cooled ceramic breeder (HCCB) blanket is one of the candidate blanket concepts for CFETR. Until now, there are many research institutes which have performed conceptual designs and comprehensive analyses works for phase I HCCB blanket in detail. However, lately, the operational stage of CFETR has transformed from phase I to phase II, and the latest core design parameters have been just determined (major radius equals to 7.2 m, minor radius equals to 2.2 m). Therefore, the design and analyses work for HCCB blanket should also be initialized. In this work, based on the original integrated optimization method, the radial structure layout optimization of the outboard equatorial phase II HCCB blanket module is conducted by NTCOC. However, the calculation results show that the original optimization method provides inadequately optimized tritium breeding performance and insufficient tritium releasing ability under the condition of high fusion power. To solve these problems, a dynamic feedback is incorporated into the original optimization method for the first time, and then the calculation results show that the problem has been solved well after revision. This work can supply worthy guidance and reference for the conceptual design and comprehensive analyses of CFETR phase II HCCB blanket.     

Safety regulatory framework for hydrogen fusion reactors in China

China is very ambitious for the development of fusion energy system using the fusion reaction of hydrogen isotopes. According to China magnetic confinement fusion development roadmap, the China Fusion Engineering Test Reactor (CFETR) is expected to be constructed in 2020s, which will be the first hydrogen fusion reactor in China. However, China has not yet established the safety regulatory framework for the licensing of a hydrogen fusion reactor such as CFETR, while the existing regulatory framework is based on the fission reactor development and there are huge differences in the safety characteristics between fusion reactor and fission reactor. In this paper, China's nuclear safety regulation system is reviewed and then key issues are raised for adapting the regulatory framework to the hydrogen fusion reactor, considering the fusion safety characteristics as well as lessons learnt from the licensing of ITER in the French regulations. Gaps between China nuclear safety regulations and CFETR safety concerned demands are also identified and discussed. It is suggested that the engineering design of CFETR shall be done together with the development of corresponding safety regulatory framework.     

CFETR聚变发电厂的储能技术适用性分析

  目的  中国聚变工程实验堆（CFETR）具有功率周期性输出特性,无法直接满足常规发电设备的稳定连续输入需求,CFETR聚变发电厂需要在核岛和常规岛之间配置针对性的工艺系统。  方法  通过调研光热发电工程应用以及聚变发电相关研究,基于CFETR聚变堆初步预期性能,从储能与补能对比、储能技术类型、储热介质、储热系统方案等方面进行分析讨论。  结果  对于采用水冷包层、氦冷包层的聚变堆,分别推荐配置以氢化三联苯型导热油、Solar salt熔融盐为储热介质的间接式双罐显热储热系统,以保证发电侧系统的稳定连续运行。  结论  上述储能技术方案具备规模化商业应用条件,可为CFETR聚变发电厂的整体设计提供支撑。     

CFETR聚变发电厂概念设计技术研究

  目的  基于中国聚变工程实验堆（CFETR）功率输出特性,在聚变堆最大热功率1.25 GW条件下,提出一种CFETR聚变发电厂概念设计方案。  方法  针对CFETR聚变堆周期性长脉冲输出特性与常规岛设备连续稳定输入需求之间的矛盾,采用储能技术方案以“削峰填谷”的方式实现CFETR聚变发电安全稳定运行。讨论了CFETR聚变水冷与氦冷两种不同包层方案下的储能以及一回路与二回路之间耦合与解耦运行模式的组合方案,并通过方案比较,给出CFETR聚变堆不同包层对应的最佳储能运行组合建议。  结果  结果表明:水冷包层、氦冷包层聚变堆分别对应的最佳组合方案是导热油+解耦储热方案与熔盐+耦合储热方案。  结论  从常规岛侧而言,相对于水冷包层方案,采用氦冷包层方案的CFETR聚变发电厂技术经济性更好。研究提出的CFETR聚变发电厂概念设计方案对后续聚变发电技术研究以及工程设计,具有较高的参考价值。     

A new direct external recycling concept for the tritium fuel cycle of CFETR

Tritium self-sustaining operation, which is achieved by a successful closed tritium fuel cycle, is one of the key scientific objectives of China Fusion Engineering Test Reactor (CFETR). The initial tritium start-up inventory and the required TBR are two key parameters which directly affect the possibility of achieving tritium self-sufficiency. In this paper, a new purge gas scheme of He + 0.n% vol. D₂ is proposed for tritium extraction of solid breeder blanket, which can avoid the existence of protium in the mixed gas. In this way, the isotope separation system (ISS) in outer fuel cycle (OFC) can be omitted and a new concept of direct external recycling (DER) has been first put forward for OFC, which aims to skip ISS-O and reduce the processing time of OFC; and the feasibility of DER is preliminarily analyzed from four aspects. On this basis, the influences of DER on both the initial start-up inventory and the required TBR have been preliminarily analyzed for different fueling efficiency and tritium burning fraction (TBF) products, and the calculation results are compared with those corresponding to the same direct internal recycling (DIR) fractions. The results indicate that the application of DER can reduce the demand for both the initial start-up inventory and the required TBR, which can improve the possibility of achieving tritium self-sufficiency appropriately. In particular, the amount of “start-up inventory saved by DER” is still attractive under the condition of high fueling efficiency and TBF, and it is in direct proportion to the fusion power. Both of these make DER technology even more attractive for DEMO reactor and future power plant (FPP). These studies can provide some valuable suggestions for the design and optimization of the tritium fuel cycle system of CFETR and FPP.     

Design and thermal‐hydraulic evaluation of helium‐cooled ceramic breeder blanket for China fusion engineering test reactor

China Fusion Engineering Test Reactor, a fusion tokamak device, is proposed to provide complementary technology and experience for ITER and the future fusion power plant. A helium‐cooled ceramic breeder blanket concept is adopted as the candidate tritium breeding blanket for China Fusion Engineering Test Reactor. Detailed design of the blanket structure located at the outboard equatorial plane is presented. The coolant flow characters in the blanket were calculated by the theoretical method and the finite element method. The comparison of the calculated results was done, and it has a good agreement between theoretical results and simulation results. The results show that the pressure drop is 0.13 MPa and the total temperature rise is 194.6°C.     

Hydrogen permeation from F82H wall of ceramic breeder pebble bed: The effect of surface corrosion

Understanding the permeation behavior of tritium from a pebble bed breeding blanket is essential for establishing a self-sufficient fuel cycle in a nuclear fusion reactor. It is known that double corrosion layers forms on reduced activation ferritic-martensitic (RAFM) steel surface by gas release from a ceramic breeder material, however, its effect on hydrogen permeation behavior has not been elucidated. Herein, in-situ measurement of hydrogen permeation through an F82H RAFM wall of a ceramic breeder pebble bed was performed under H₂-added sweep gas conditions. The corrosion layer formed on the F82H sample had a dense microstructure, which reduced hydrogen permeation flux at least by one order of magnitude. The permeation reduction factors were 20–50 at the water-coolant temperature of a blanket. A self-repairing ability is expected for the surface oxide layer as the corrosion occurs spontaneously inside a breeding blanket.     

Thermal hydraulic analysis of the HECLIC blanket breeder unit for CFETR

China Fusion Engineering Test Reactor (CFETR) is proposed by the China National Integration Design Group. A helium‐cooled lithium ceramic (HECLIC) blanket for CFETR has been designed by Institute of Plasma Physics Chinese Academy of Sciences (ASIPP) and University of Science and Technology of China (USTC). The HECLIC blanket includes the tungsten armor, first wall (FW), breeder units (BUs), caps, stiffening grids and back plates. The BU consists of the beryllium pebble beds, lithium ceramic pebble beds and cooling plates. The stiffening grid reinforces the blanket structure and also cools the BU together with cooling plates. Thermal hydraulic analysis has been performed to assess the heat transfer coefficient (HTC) and temperature distribution. The results indicate that the velocity is fairly uniform in the cooling loops, and the HTC is high enough to remove the heat timely. The maximum temperatures are within the design temperature limits. Besides, optimization has been done to improve the layout of the cooling channels. In addition, thermal mechanical analysis has been carried out. The maximum stress can satisfy design limits, and it proves the feasibility of the design. Copyright © 2014 John Wiley & Sons, Ltd.     

A neutronic investigation on a helium cooled hybrid reactor using nitride fuels containing reactor grade plutonium

There has been a significant amount of reactor grade (RG) plutonium accumulated from the conventional nuclear reactors’ spent fuel. Destruction or reducing this RG plutonium is very important to prevent its misuse and/or release accidentally into the environment. Using very energetic fusion neutrons in fusion–fission (hybrid) reactors can burn the RG plutonium effectively. This study presents the burning of the mixed fuel containing RG plutonium and uranium in a helium cooled hybrid reactor for an operation period of 24 months. Effect of various fuel mixtures and tritium breeders on the neutronic performance of the reactor as well as the burning of the RG plutonium was investigated. Calculations were carried out on an experimental hybrid blanket with the aid of SCALE4.3 by solving the Boltzmann transport equation with the XSDRNPM code. Numerical results showed that increasing RG plutonium content in the fuel increased the burning of plutonium and fusion energy multiplication substantially. The tritium breeder having the lowest lithium atomic density allowed the highest fission of the fuel in the blanket.     

超导磁体技术与磁约束核聚变

  目的  磁约束核聚变是解决能源问题的有效途径之一。为了实现准稳态运行,超导磁体（特别是高场高温超导磁体）已成为未来托卡马克设计的首选方案。  方法  介绍了EAST的最新实验进展及未来研究计划,并从超导磁体技术方面总结了未来聚变装置CFETR的最新进展。  结果  2021年底,世界首个全超导托卡马克EAST（Experimental Advanced Superconducting Tokamak）成功实现1056 s长脉冲高参数等离子体运行,创造最长运行时间的世界记录。  结论  中国聚变工程试验堆（CFETR, China Fusion Engineering Test Reactor）的设计已经完成,它将填补国际热核聚变实验堆（ITER, International Thermonuclear Experimental Reactor）和示范堆（DEMO）间的空白。     

Integral measurements in a simulated fast breeder reactor blanket

Much of the breeding of fissile material and a significant fraction of the energy production in proposed fast breeder reactors will occur in a blanket of fertile material surrounding the reactor core. Current uncertainties in neutron cross sections and calculational methods have much greater significance for blanket design than for core design. The Purdue University Fast Breeder Blanket Facility (FBBF), designed to simulate the neutron behavior in fast breeder reactor blankets, has recently been completed. The facility is now being used to perform integral neutron reaction rate measurements. Such measurements provide tests of nuclear cross sections and calculational methods used in the design of fast breeder reactor blankets.     

Hydrogen isotopes separation validation of frontal displacement chromatography for various compositions of feed gas and tritium extraction simulation for TBM

Based on the previous study in frontal displacement chromatography (FDC) packed with Pd-Al₂O₃, three groups of separation tests were carried out to verify the separation performance of the constituted FDC device for various compositions of feed gas and to validate the application probability of FDC in the Tritium Extraction System (TES) of ITER and China Fusion Engineering Test Reactor (CFETR). The separations were conducted by the FDC procedure with characteristics of the feed gas one-time flushing though the column and then reasonable separation performance had been obtained. The results indicate that the FDC process could be applied to deal with the desorbed gas mixtures from TES and/or further to extract and thereafter enrich the breeding tritium in ITER or CFETR, which would take the advantages of system compactness and efficiency over the present route of TES. Comparing to other related displacement chromatography procedures, the FDC process could be applied in tritium pre-enrichment for the mixtures of low tritium concentrations, which is highlighted by the outstanding merit of operation simplicity.     

Core design of long‐cycle small modular lead‐cooled fast reactor

A core design of small modular liquid‐metal fast reactor (SMLFR) cooled by lead‐bismuth eutectic (LBE) was developed for power reactors. The main design constraint on this reactor is a size constraint: The core needs to be small enough so that (1) it can be transported in a spent nuclear fuel (SNF) cask to meet the electricity demands in remote areas and off‐grid locations or so that (2) it can be used as a power source on board of nuclear icebreaker ships. To satisfy this design requirement, the active core of the reactor is 1 m in height and 1.45 m in diameter. The reactor is fueled with natural and 13.86% low‐enriched uranium nitride (UN), as determined through an optimization study. The reactor was designed to achieve a thermal power of 37.5 MW with an assumption of 40% thermal efficiency by employing an advanced energy conversion system based on supercritical carbon dioxide (S‐CO₂) as working fluid, in which the Brayton cycle can achieve higher conversion efficiencies and lower costs compared to the Rankine cycle. The outer region of the core with low‐enriched uranium (LEU) performs the function of core ignition. The center region plays the role of a breeding blanket to increase the core lifetime for long cycle operation. The core working fluid inlet and outlet temperatures are 300°C and 422°C, respectively. The primary coolant circulation is driven by an electromagnetic pump. Core performance characteristics were analyzed for isotopic inventory, criticality, radial and axial power profiles, shutdown margins (SDM), reactivity feedback coefficients, and integral reactivity parameters of the quasi‐static reactivity balance. It is confirmed through depletion calculations with the fast reactor analysis code system Argonne Reactor Computation (ARC) that the designed reactor can be operated for 30 years without refueling. Preliminary thermal‐hydraulic analysis at normal operation is also performed and confirms that the fuel and cladding temperatures are within normal operation range. The safety analysis performed with the ARC code system and the UNIST Monte Carlo code MCS shows that the conceptual core is favorable in terms of self‐controllability, which is the first step towards inherent safety.     

Hydrogen production via steam-methane reforming in a SOMBRERO fusion breeder with ceramic fuel particles

In the present study, hydrogen production potential of SOlid Moving BREeder ReactOr (SOMBRERO) fusion reactor and heat recovery of this system is investigated. The original SOMBRERO fusion reactor has a 1000 MWe KrF laser-driven IFE power plant. This reactors fusion power is 2677 MW and total thermal power is 2891 MW. The blanket is divided into three breeding zone and these breeding zones have different C, Li₂O and ceramic fuel particles. One-dimensional neutronic calculations of SOMBRERO fusion reactor have been performed by using XSDRNPM/SCALE4.4a neutron transport code. Steam Methane Reforming (SMR) method is used for large-scale hydrogen production and heat recovery of waste heat is analyzed. The numerical results show that the considered SOMBRERO fusion reactor has a good neutronic performance as well as the high hydrogen production potential with heat recovery of SMR process.     

A space power system of free piston Stirling generator based on potassium heat pipe

The power system of a free piston Stirling generator (FPSG) based on potassium heat pipes has been developed in this paper. Thanks to the advantages of long life, high reliability, and high overall thermal efficiency, the FPSG is a promising candidate for nuclear energy, especially in space exploration. In this paper, the recent progress of FPSG based on nuclear reactor for space use was briefly reviewed. A novel FPSG weighted only 4.2 kg was designed, and one dimensional thermodynamic modeling of the FPSG using Sage software was performed to estimate its performance. The experiment results indicated that this FPSG could provide 142.4 W at a thermal-to-electric efficiency of nearly 17.4%. Besides, the power system integrated with four FPSGs and potassium heat pipes was performed and the single machine failure test was conducted. The results show that this system could provide an electrical power of 300 W at an overall thermal efficiency of 7.3%. Thus, it is concluded that this power system is feasible and will have a great prospect for future applications.     

Measurement of subcritical growth of defects in large components of nuclear power plants at elevated temperatures

An evaluation of fracture parameters of components to be integrated into high-power nuclear systems (1000 MW) is of a great importance because of extremely high safety requirements. Materials to be used for both pressure vessels and pipes including welds must comply with numerous requirements, fracture resistance being one of the most important of them. Properties of the high-performance steels to be used are experimentally verified on specimens. However, eventually, an important part of the experimental programme are fracture tests of full-scale models of real components.The paper deals with problems of measurement of initiation and subcritical growth of crack-like defects in large full-scale models of the connection of pipes to a reactor pressure vessel during fracture experiments at elevated temperatures. A modification of a direct current potential drop method (DCPD) was used to compensate influences of: (i) non-uniform cross-section of pipes, (ii) complicated shape and (iii) high temperatures. Methods how analytical calibration curves can be used in these specific complicated shapes are described. It was demonstrated that the high temperature is not a limiting factor for exact measurement of crack length if the compensation and computer controlled device is used. Results of a measurement on the component and an evaluation method of subcritical crack initiation and growth are presented.     

Status of helium-cooled nuclear power systems

Helium-cooled nuclear power systems offer a great potential for electricity generation when their long-term economic, environmental, conservation and energy self-sufficiency features are examined. The high-temperature gas-cooled reactor (HTGR) has the unique capability of providing high-temperature steam for electric power and process heat uses and/or high-temperature heat for endothermic chemical reactions. A variation of the standard steam cycle HTGR is one in which the helium coolant flows directly from the core to one or more closed cycle gas turbines. The effective use of nuclear fuel resources for electric power and nuclear process heat will be greatly enhanced by the gas-cooled fast breeder reactor (GCFR) currently being developed. A GCFR using thorium in the radial blanket could generate sufficient U-233 to supply the fuel for three HTGRs, or enough plutonium from a depleted uranium blanket to fuel a breeder economy expanding at about 10% per year. The feasibility of utilizing helium to cool a fusion reactor has been included in most research studies on thermonuclear fusion and is also discussed in this paper. This paper summarizes the status of helium-cooled nuclear energy systems as a basis for assessing their prospects.     

The effects of irradiation and high temperature on chemical states in Li2TiO3

In the future D-T fusion reactor, tritium will be bred mainly by the reaction of ⁶Li (n, α) T, as well as ⁷Li (n, n’ α) T in tritium breeding materials. Solid breeding materials will experience harsh conditions under both irradiations by energetic particles (neutron, tritium, helium and self-particles) and high temperature. The interactions of irradiations and high temperature on lithium ceramics will influence tritium breeding ratio (TBR). The changes of chemical states and its effects on release behavior of hydrogen isotopes in deuterium-irradiated Li₂TiO₃ and deuterium-exposed Li₂TiO₃ at high temperature have been investigated. The peak of O-1s shifted to higher binding energy by both irradiation and deuterium exposure, indicating that O-D bonds formed. The amount of O-D bonds enhanced as increase of irradiation fluence and exposure temperature. The main deuterium atoms were trapped by defects for irradiated samples. Annihilation of E-centers was thought to trigger the release of hydrogen isotopes. O-D bonds were the main deuterium trapping sites in deuterium-exposed Li₂TiO₃. Deuterium recovered by detrapping O-D bonds would require higher temperature. Both deuterium-irradiation and deuterium-exposure at high temperature could result in the change of chemical states in Li₂TiO₃. The changes in chemical states had effects on deuterium release. It illustrates that tritium breeding materials in fusion reactor will be modified by both irradiation and high temperature and could result in lower tritium recovery.