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Pebble beds are used in many tritium breeder designs of fusion blanket, which is purged by a flow of low velocity helium. In this paper, experimental study of the pressure drop of helium through packed sphere beds was carried out. The pebbled beds were sealed in a rectangular container. The spheres of five different diameters varied from 0.5 mm to 2.0 mm. The Darcy regime and Forchheimer regime were identified within the range 2.47 < Re < 27.59. The transition region was identified and discussed in detail. The pressure drop was characterized by Kozeny-Carman constants or Ergun's constants for the experimental range. Some commonly used correlations were evaluated by the experimental data. [ABSTRACT FROM AUTHOR]

A conceptual design of helium cooled solid breeder (HCSB) blanket, one of three blanket candidates for Chinese Fusion Engineering Test Reactor (CFETR), has been performed recently. The transient analysis for different possible accidents should be carried out to assess its safety performance. In this paper, the ITER-like conditions are adopted since the associated system for CFETR is missing, such as helium cooling system, plasma shutdown condition, pump behavior, etc. The complete model is recreated including the optimized typical outboard HCSB blanket (NO.12) and its ancillary helium cooling loop, and accident analyses of two loss of coolant accident (LOCA) cases are investigated using RELAP5. The influences of different break areas under in-vessel LOCA are compared, and the accident consequence after small area break is chosen to be investigated. Regarding the ex-vessel LOCA, the influences of different break locations are thoroughly analyzed. Since the plasma cannot terminate passively after ex-vessel pipe break, the plasma termination behaviors are investigated with different shutdown time. The computational results show that with the safety criteria for ITER the HCSB blanket can be cooled down effectively by the helium cooling system (HCS) and the integrity of pressure barriers can be guaranteed for both accidents. [ABSTRACT FROM AUTHOR]

Highlights: [•] A thermal-hydraulic analysis code named TSACO-HCCB TBM was developed. [•] The code was verified by comparing with RELAP5. [•] The design basis accident in-vessel LOCA analysis was performed with this code. [•] This article is useful for design and operation of helium cooling system. [Copyright &y& Elsevier]

Highlights: [•] Thermal-hydraulic and safety analysis on DFLL-TBM system is performed. [•] The TBM FW maximum temperature is 541°C under steady state condition. [•] The TBM FW maximum temperature does not exceed the melt point of CLAM steel 1500°C. [•] Neither the VV pressurization nor vault pressure build-up goes beyond 0.2MPa. [ABSTRACT FROM AUTHOR]

Abstract: The thermal–hydraulic behavior and safety performance of the Chinese helium-cooled solid breeder (CH HCSB) test blanket module (TBM) with helium cooling system (HCS) has been studied using RELAP5/Mod3.4 code. According to accident analysis specification for TBM, two design basis accidents including loss of off-site power and TBM first wall (FW) ex-vessel coolant pipe break are investigated. The influences of different break locations and plasma termination behaviors are analyzed comprehensively. The results show that natural circulation is established in helium cooling circuit and the TBM can be cooled effectively after loss of off-site power. It is much more critical when the pipe break occurs at the downstream side of the circulator compared with that of upstream side of the circulator. In case of a more serious accident that the ex-vessel break extends to the TBM FW, the results reveal that TBM could be cooled down by natural circulation and radiation. In addition, at the beginning of ex-vessel loss of coolant accident (LOCA), large temperature difference between break and intact TBM FW pipes is found. The accidental results finally show that the integrity of the FW can be guaranteed if the plasma is terminated with a 3s delay time by fusion power shutdown system (FPSS) in the case of ex-vessel LOCA. [Copyright &y& Elsevier]

• The influence of poloidal nonuniform NWL is studied for HCCB blanket. • The symmetrical and stepped NWL distributions are selected. • The overall influence of nonuniform neutron distribution is limited. • The local influence of nonuniform NWL is obvious with the difference over 20 %. After its conceptual design, the Chinese Fusion Engineering Test Reactor (CFETR) is now at the stage of detailed engineering design. During this stage, many detailed factors should be considered for the components in the CFETR. In this paper, the neutronics influence of poloidal nonuniform neutron wall loading (NWL) is studied for helium-cooled ceramic breeder (HCCB) blanket of the CFETR. The blanket is divided into five parts along the poloidal direction. Each part is composed of the armor, the first wall, tritium breeders, neutron multipliers, and the manifold in the flow distribution zone. The nuclear heat deposition and the tritium breeding performance of the CFETR HCCB blanket are studied with symmetrical and stepped NWL distributions. The results obtained from the uniform NWL distribution are also presented as the reference. For each NWL distribution, the nuclear heat deposition and the tritium breeding capability in different blanket parts are compared. The results show that the nonuniform NWL distribution has almost no influence on the total nuclear heat deposition or the total tritium production. The overall variation is less than 1 ‰ compared to the results under the uniform NWL distribution. However, the local performances of different blanket parts are affected obviously. The maximum variation is over 20 %, which should be focused on during the design for the flow distribution of helium coolant and purge gas. The present work aims to provide an important guidance for the detailed design and structural optimization of the blanket for the CFETR, especially for consideration of the high-level fusion power during phase II (over 1 GW). [ABSTRACT FROM AUTHOR]

• The research of flow instability in water cooled blanket is carried out. • The effect of channel arrangement is studied under BWR and PWR conditions. • The phenomena of flow excursion and flow oscillation are analyzed. • The flow stability boundaries are obtained. Water is adopted as coolant in many blanket concepts, and the flow condition is designed based on its use in boiling water reactor (BWR) and pressure water reactor (PWR). The flow instability may occur in the water cooled blanket when water undergoes two-phase flow during steady state or under some accidents. The flow rate in one channel will change and oscillate, while the pressure is still constant. In this study, the research of flow instability in parallel channels, including flow excursion and flow oscillation, is performed using the thermal-hydraulic analysis code RELAP5. The phenomenon of flow instability is detected by the method of small disturbance. The different conditions of BWR and PWR are investigated. The effect of arrangement of parallel channels, vertically and horizontally, is also studied under 7 MPa for BWR condition and 15.5 MPa for PWR condition. Finally, the stability map of the four conditions is obtained. The analysis shows that the flow is more stable under PWR condition since the equilibrium quality of stability boundary is higher. The arrangement of coolant channels has different influences under different conditions. Under PWR condition, it will stabilize the system if the flow channels are horizontally arranged. However, the influence of the channel arrangement on the flow boundary is limited under BWR condition. The research work can provide an important reference for design and operation of water cooled blanket, especially for the consideration of high heat removal. [ABSTRACT FROM AUTHOR]

In this study, a three-step approach of ordinal 3D-1D&2D-3D is independently proposed to derive a coupling optimization design rule of the Chinese Fusion Engineering Test Reactor (CFETR) helium cooled solid breeder (HCSB) blanket. Firstly, the three dimensional (3D) neutronics model of CFETR is built with all of the 15 blanket modules adopting the original radial build arrangement parameters. Based on this model, both the neutron wall loading (NWL), which will serve as the normalization factors for the following local neutronics calculations, and the tritium breeding ratio (TBR) distributions on each blanket module are obtained. Secondly, the radial build arrangements of all the 7 inboard and the 8 outboard blanket modules in an 11.25° toroidal CFETR sector are optimized by taking both the tritium breeding and thermal performances into account as the optimization objectives. NTCOC, a N eutronics/ T hermal-hydraulic C oupling O ptimization C ode, which has been independently developed by Xi’an Jiaotong University for accelerating the radial build design and optimization of the fusion reactor HCSB blanket concept with adopting the simplified 1D neutronics and 2D thermal-hydraulic models, is used for these work. Thirdly, based on the final optimized schemes of all the blanket modules, a comprehensive research on the coupling optimization design rule of the CFETR HCSB blanket is performed in-depth and a coupling optimization design integration approach has been finally summarized. Finally, the 3D CFETR neutronics model is updated by replacing all of the original blanket radial build arrangement parameters with the final optimized schemes. The calculation results show that after the radial optimizations, the total TBR of all the 15 blanket modules has increased from the previous 1.092–1.274, which can well meet the self-sufficiency tritium breeding requirement of CFETR. [ABSTRACT FROM AUTHOR]

As one of three candidate tritium breeding blanket concepts for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of helium cooled solid breeder (HCSB) blanket was recently proposed. In this paper, the preliminary thermal-hydraulic and safety analyses of the typical outboard equatorial blanket module (No.12) have been carried out using RELAP5/Mod3.4 code. Two design basis accidents are investigated based on the steady-state initialization, including Loss of Off-Site Power and In-Box Loss of Coolant Accident (LOCA). The differences between circulator coast down and circulator rotor locked under Loss of Off-Site Power are compared. Regarding the In-Box LOCA, the influences of different break sizes and locations are thoroughly analyzed based on a relatively accurate modeling method of the heat structures in sub-modules. The analysis results show that the blanket and the combined helium cooling system (HCS) are designed with sufficient decay heat removal capability for both accidents, which can preliminarily verify the feasibility of the conceptual design. The research work can also provide an important reference for parameter optimization of the blanket and its HCS in the next stage. [ABSTRACT FROM AUTHOR]

In this paper, the supercritical-water cooled solid breeder test blanket module (SWCB TBM), using the supercritical water as the coolant, Li 4 SiO 4 lithium ceramic pebbles as a breeder, and beryllium pebbles as a neutron multiplier, was designed and analyzed for ITER. The results of neutronics, thermo-hydraulic and thermo-mechanical analysis are presented for the SWCB TBM. Neutronics calculations show that the proposed TBM has high tritium breeding ratio and power density. The tritium breeding ratio (TBR) of the proposed design is 1.17, which is greater than that of 1.15 required for tritium self-sufficiency. The thermo-hydraulic calculation proved that the TBM components can be effectively cooled to the allowable temperature with the temperature of outlet reaching 500 °C. According to thermo-mechanics calculation results, the first wall with the width of 17 mm is safe and the deformation of first wall is far below the limited value. All the results showed that the current TBM design was reasonable under the ITER normal condition. [ABSTRACT FROM AUTHOR]

Abstract: A Water-cooled Pressure Tube Energy production blanket (WPTE) for fusion driven subcritical reactor has been designed to achieve 3000MW thermal power with self-sustaining tritium cycle. Pressurized water has great advantages in energy production; however the high pressure may cause some severe structural design issues. This paper proposes a new concept of water-cooled blanket. To solve the problem of the high pressure of the coolant, the pressure tube was adopted in the design and in the meantime, the thickness of the first wall can be significantly reduced as result of adopting pressure tube. The numerically simulating and calculating of temperature, stress distribution and flow analyses were carried out and the feasibility of using water as coolant was discussed. The results demonstrated the engineering feasibility of the water-cooled fusion–fission hybrid reactor blanket module. [Copyright &y& Elsevier]

• Three-dimensional full-scale thermal-hydraulic analysis of fusion blanket is performed after validation of numerical method. • Specific structure modifications are proposed for initial blanket components. • The optimized blanket can satisfy temperature limits for all materials. Fusion energy is widely considered as the most promising method to solve the energy crisis, while the blanket is of vital importance in the fusion device due to its main function of thermal extraction and energy conversion. The high heat flux from plasma and the internal nuclear heat deposition are imposed on fusion blanket. Slice model or two-dimensional model is usually adopted for traditional design of fusion blanket, in which most side effects are neglected, introducing relatively large uncertainty for thermal removal capability of the blanket. In this study, the integral three-dimensional full-scale model is established based on the split-zone method for mesh generation. Based on the validated numerical method, the preliminary thermal-hydraulic analysis of the full-scale model shows that inadequate heat transfer and significant flow distribution difference are encountered for the initial blanket design, in which the full-scale effect is not considered. After that, the specific structure modifications are proposed for different blanket components to solve the abovementioned problems. The maximum temperature of first wall is reduced from 671°C to 639°C, which is under the material temperature limit of 650°C; the inverse V-shaped flow distribution in the cap is modified as more even flow in cap channels; the maximum temperature of the function zone in blanket is reduced from 999°C to 759°C, which is an important structure modification for the blanket. The integral three-dimensional full-scale thermal-hydraulic analysis is performed after the local structure modification. The results show that the function zone is cooled down obviously by the coolant in the side wall. Compared to the temperature limit of 900°C for lithium orthosilicate, there is a more rigorous limit for the beryllium material, since they share the common cooling capability layer by layer while the temperature limit for the latter is only 650°C. The three-dimensional full-scale thermal-hydraulic analysis in our research is valuable to evaluate the integral reliability of the fusion blanket. [ABSTRACT FROM AUTHOR]

Highlights • A new pressure drop limit has been added into the optimization criteria. • The previous integrated optimization algorithm of NTCOC is improved. • The improved code has been applied on the HCSB blanket for CFETR Phase II. • The computational processes with two methods are compared. Abstract Blanket is the key component of Chinese Fusion Engineering Test Reactor (CFETR). In our previous work, considering that the radial arrangement of the internal functional zones has the greatest influence on the blanket performances and is also the most important basis of the 3D design, NTCOC, a Neutronics/Thermal-hydraulic Coupling Optimization Code which adopts 1D neutronics and 2D thermal-hydraulic simplified calculation models, has been independently developed for facilitating the radial build design and optimization of the CFETR helium cooled solid breeder (HCSB) blanket. Through setting three leading design requirements as the optimization criteria, this code has been applied on optimizing the radial build arrangement of a conceptual design of HCSB blanket for CFETR Phase I, which preliminarily verifies the reasonability of the code. However, recently, the design phase of CFETR has gradually changed from I to II, the fusion power of which will be quintuple to the previous I. This means that the internal nuclear power deposition in each blanket component under II operation condition will be much larger than the previous I. Therefore, much larger coolant inlet mass flowrate will be needed and more breeder units should to be inserted, both of which will make the total pressure drop of blanket for Phase II much larger than the previous I and even exceed the limit of CFETR HCSB blanket greatly. This can’t be sustained in CFETR, a power plant that should produce electricity. To solve this problem, in this paper, the previous integrated neutronics/thermal-hydraulic optimization algorithm is improved by adding the new pressure drop limit into the optimization criteria. Then the improved code has been applied on optimizing the radial build arrangement of a conceptual design of HCSB blanket for CFETR Phase II through constant and variable step length two methods. The calculation results verifies the feasibility and reasonability of the improved optimization algorithm preliminarily. [ABSTRACT FROM AUTHOR]

As one of the most important components of fusion reactor, breeding blanket mainly accounts for tritium self-sufficiency, shielding and fusion energy collection. The performance of the blanket is mainly determined by the internal structural arrangement and material composition. However, if we directly adopt a three-dimensional calculation model for the blanket design and optimization, not only will it consume extensive human and computer resources, but also make the computational process extremely complicated and lengthy. Considering that the radial arrangement of the functional zones, which has the greatest influence on the performance, is the basis of the detailed and accurate 3D design of the blanket, NTCOC, a N eutronics/ T hermal-hydraulics C oupling O ptimization C ode, which adopts simplified 1D neutronics and 2D thermal-hydraulic calculation models, is developed for accelerating the radial build design and optimization of the blanket in this work. This code realizes the functions of both the automatic iterative updating the temperature-dependent cross-section libraries and the data transmission between the neutronics and thermal-hydraulic calculation results through the inner and external coupling between the MCNP4C and CFD software, respectively. Therefore, this can make the comprehensive optimization analyses progress much faster and more accurate. In addition, a variable step length method is adopted to further improve the coupling optimization speed of this code, and several leading design objectives are set as the optimization criteria instead of taking all the possible practical engineering constrains into account. Finally, this code has been applied to optimize the radially arranged zones of a conceptual design scheme of Chinese Fusion Engineering Test Reactor (CFETR) helium cooled solid breeder blanket, which adopts mixed pebble beds as both the tritium breeder and neutron multiplier, under both the normal and critical conditions. [ABSTRACT FROM AUTHOR]

As one of the key components of fusion reactor, breeding blanket is of vital importance for the functions of tritium self-sufficiency and fusion energy transformation. The performances of the blanket mainly depend on the internal structural arrangement and material composition. However, if we directly adopt the actual 3D calculation model for the blanket design and optimization, not only will it cost extensive human and computer resources, but also make the computational process extremely complicated and lengthy. Considering that the radial build arrangement of the functional zones is the most important determinant for the detailed 3D design of the blanket, NTCOC, a N eutronics/ T hermal-hydraulic C oupling O ptimization C ode, which adopts 1D neutronics and 2D thermal-hydraulic simplified models, is developed for accelerating the radial build design and optimization of the blanket in this work. This code realizes the iterative updating the temperature cross-section libraries through inner coupling, which can make the calculation more accurate. Besides, it realizes the automatic data transmission between the nuclear and thermal calculation results, which can facilitate the coupling optimization process greatly. In addition, a variable step length method is adopted for further improving both the coupling optimization efficiency and accuracy of this code, and several leading design requirements are set as the optimization criteria instead of taking all the practical engineering constrains into account. Finally, after the improvement of the algorithm of the integrated neutronics/thermal-hydraulic optimization for NTCOC, this code has been applied on optimizing the radial build arrangement of a conceptual design of Chinese Fusion Engineering Test Reactor (CFETR) helium cooled solid breeder blanket, which adopts separated Li 4 SiO 4 and Beryllium pebble beds as tritium breeder and neutron multiplier respectively, under normal, critical and DEMO three conditions. [ABSTRACT FROM AUTHOR]

As one of the candidate tritium breeding blankets for Chinese Fusion Engineering Test Reactor (CFETR), a conceptual structure of the helium cooled solid breeder blanket has recently been proposed. The neutronic, thermal-hydraulic and mechanical characteristics of the blanket directly affect its tritium breeding and safety performance. Therefore, neutronic/thermal-hydraulic/mechanical coupling analyses are of vital importance for a reliable blanket design. In this work, first, three-dimensional neutronics analysis and optimization of the typical outboard equatorial blanket module (No. 12) were performed for the comprehensive optimal scheme. Then, thermal and fluid dynamic analyses of the scheme under both normal and critical conditions were performed and coupled with the previous neutronic calculation results. With thermal-hydraulic boundaries, thermo-mechanical analyses of the structure materials under normal, critical and blanket over-pressurization conditions were carried out. In addition, several parametric sensitivity studies were also conducted to investigate the influences of the main parameters on the blanket temperature distributions. In this paper, the coupled analyses verify the reasonability of the optimized conceptual design preliminarily and can provide an important reference for the further analysis and optimization design of the CFETR helium cooled solid breeder blanket. [ABSTRACT FROM AUTHOR]

• The neutronics and the thermal-hydraulic analyses for multiple operation modes under both the symmetrical and the stepped NWL distributions are conducted. • Under the actual nonuniform NWL distributions, there is obvious poloidal temperature nonuniformity in every functional zone. • The thermal-hydraulic results calculated under the simplified uniform NWL distribution are found to be not conservative. • The degree of the influences of the poloidal nonuniform NWL on blanket performances for multiple operation modes are compared. Helium Cooled Ceramic Breeder (HCCB) blanket is one significant component of China Fusion Engineering Test Reactor (CFETR). The neutronics and thermal-hydraulic performances of the blanket are the bases of the safe operation for CFETR, and both of them are influenced by the neutron wall loading (NWL) distribution on the blanket. In our previous work, the influences of the poloidal nonuniform NWL on the neutronics performances of the CFETR HCCB blanket have been preliminarily investigated. However, as the nuclear heating rate in each blanket component is adopted as the internal heat source for the thermal-hydraulic analyses, it's necessary to further investigate the influences of the poloidal nonuniform NWL on the thermal-hydraulic performances of the blanket. Besides, CFETR aims to cover multiple fusion powers (from 200 MW to 1.5 GW) at the present stage. Therefore, it's also essential to compare the different influences of the poloidal nonuniform NWL on blanket performances for multiple operation modes. [ABSTRACT FROM AUTHOR]

• The influence of the poloidal nonuniform NWL on the thermal-hydraulic performances of the blanket is researched. • Two symmetrical and one stepped NWL have been proposed aggregately as the poloidal nonuniform NWL distribution forms. • Obvious poloidal temperature nonuniformity is observed in every functional zone under the nonuniform NWL distribution. • The thermal-hydraulic results calculated under the simplified uniform NWL distribution are not conservative. Helium Cooled Ceramic Breeder (HCCB) blanket, which has been determined as one of the blanket candidates for China Fusion Engineering Test Reactor (CFETR), is the critical component of the fusion reactor. The neutronics and thermal-hydraulic performances of the blanket are the bases of the safe operation, and they are also the initial boundary conditions for the other performance analyses. Both performances are influenced by the neutron wall loading (NWL) distribution on the blanket and the radial structural arrangement of the internal functional zones. In our previous work, the influences of the poloidal nonuniform NWL on the neutronics performances of the blanket have been investigated preliminarily. However, as the nuclear power deposition in each blanket component is the internal heat source for the thermal-hydraulic analyses, it's essential to further investigate the influence of the poloidal nonuniform NWL on the thermal-hydraulic performances of the blanket. In this paper, taking the typical CFETR HCCB blanket module as the research object, firstly, both the neutronics and the thermal-hydraulic models are established. For investigating the influence of the poloidal nonuniform NWL on blanket performances, both models are divided into five identical parts along the poloidal direction. Afterwards, by referencing the actual poloidal NWL distribution in CFETR, we have aggregately proposed three kinds of poloidal nonuniform NWL distribution forms, including two symmetrical and one stepped NWL. On these bases, both the neutronics and the thermal-hydraulic analyses under the poloidal nonuniform NWL distributions are conducted, and the calculation results are compared with those corresponding to the uniform NWL distribution. The comparison results show that the maximum temperature of every functional zone calculated under the three nonuniform NWL distributions is generally higher than that corresponding to the uniform NWL distribution, which means that the thermal-hydraulic results calculated under the simplified uniform NWL distribution are not conservative. Under the condition of nonuniform NWL distributions, there is obvious poloidal temperature nonuniformity in every functional zone. Besides, the temperature deviations between the stepped and the uniform NWL distributions are generally larger than those between the symmetrical and the uniform ones. This paper can provide some valuable references for the detailed engineering design, analyses, and optimizations of the CFETR HCCB blanket. [ABSTRACT FROM AUTHOR]

• The 3D neutronics analyses and validation of the optimized CFETR HCCB blanket are conducted. • The 3D thermal-hydraulic analyses and validation of the optimized CFETR HCCB blanket are conducted. • A 3D thermal revision is added to the previous integrated optimization method in the form of feedback. • After revision, the problem of not conservative 2D thermal-hydraulic calculation model has been well solved. The Helium Cooled Ceramic Breeder (HCCB) blanket is the critical component of the China Fusion Engineering Test Reactor (CFETR). The radial structural layout of the internal functional zones has the largest impact on the neutronics and thermal-hydraulic performance of the blanket, and it's also the major determinant for the detailed 3D design. In the previous work, NTCOC, a Neutronics/Thermal-hydraulic Coupling Optimization Code, has been developed for application in the radial structural layout optimization of the CFETR HCCB blanket. However, for increasing the optimization efficiency, the NTCOC employs the simplified 1D neutronics and 2D thermal-hydraulic calculation models, the results of which may differ from the calculation results of the 3D actual models. Therefore, it's essential to verify the reliability of the simplified models by comparing their calculation results with the corresponding 3D models. In this work, both the 3D neutronics and thermal-hydraulic analyses corresponding to the radial structural optimization processes of the CFETR HCCB blanket are performed and the corresponding calculation results are compared with NTCOC. The results show that adopting the 1D neutronics model in NTCOC can reflect the change tendency of the tritium breeding performance during the radial optimization processes accurately. However, the radial temperature distribution which is calculated by the 3D thermal-hydraulic model is generally higher than the NTCOC results, which means that the direct application of the 2D thermal-hydraulic model in NTCOC is not conservative. For solving this problem, a 3D thermal revision is added to the previous optimization method in the form of feedback. After revision, the 3D calculated temperature of the newly obtained blanket scheme is generally lower than the NTCOC's, which means that the previous problem has been well solved. This work comprehensively verifies the reliability of the NTCOC and the improved optimization method after the 3D thermal revision for application in the radial structural optimization of the CFETR HCCB blanket. [ABSTRACT FROM AUTHOR]