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• A design analysis code for sodium heated OTSG is developed based on the power mesh method. • The code is verified with various design data both of integral type and modular type OTSGs. • The influence of tube diameter on heat transfer capacity has been analyzed by the code. Sodium heated once-through steam generator (OTSG) is a critical component of Sodium cooled Fast Reactor (SFR). It is a counter current heat exchanger. It transfers heat from hot sodium to water. Therefore, the steam generator is a necessity for the operation of the plant. However, due to the phase change exists in water side, it becomes challenging to design a steam generator for SFR. After the successful construction of CEFR (China Experiment Fast Reactor), a power-up demonstration SFR is being developed in China. For the design and analysis of the steam generator of the demonstration SFR, a thermal hydraulic Design Code of Once-through Steam generator for Sodium cooled fast reactor (DeCOSS) has been developed based on the movable mesh method. The DeCOSS code has been benchmarked based on the design data of OTSGs reported. The tube lengths and temperature distributions calculated by DeCOSS showed general agreements with published data. The influences of tube diameters on heat transfer capacity have been analyzed. The results indicated that the heat transfer capacity is determined by the thermal resistance of tube wall under the condition of fixed heat transfer coefficient of water side. In addition, under the same flow area of water and sodium, the heat transfer capacity could be enhanced by decreasing the tube diameters with constant tube wall thickness. Hence, the code can be utilized to do the sensitivity analyses and optimum thermal design of OTSGs. Moreover, it can support the design work of once-through steam generator of SFR. [ABSTRACT FROM AUTHOR]

Based on the analysis of energy structure and the development trends of advanced nuclear energy technology, combining the complementary advantages of nuclear energy and other energy sources in China, the Fluoride-Salt-cooled high-Temperature Advanced Reactor (FuSTAR) system has been proposed. FuSTAR system is designed to address production capacity issues in areas with energy shortages or resource concentration and can be combined with high-temperature process heat to achieve high-value-added processes such as hydrogen production. In this paper, the Tri-structural isotropic fuel combined with helical cruciform elements has been studied in detail on the neutron physics and thermal-hydraulic properties, and the flow and heat transfer experiment is carried out for the first time. The multilayer nonlinear programming method has been applied to the design of thermal transport systems. Then the optimal parameters and configuration of the thermoelectric conversion system are obtained by the superstructure optimization method. The inherent safety of FuSTAR is demonstrated through deterministic safety analysis. In addition, the iodine-sulfur cycle technology and the characteristics of tritium production are also analyzed. This study highlights the significant potential of integrating nuclear energy with other energy sources in an optimized energy-production system. • Multi-purpose reactor energy system for power - hydrogen - tritium production. • Over 50 % maximum thermal efficiency. • Simultaneous analysis of design, optimization, and safety. [ABSTRACT FROM AUTHOR]

• A conceptual design of FuSTAR is developed, and validated by neutronics and thermal-hydraulics analysis. • Detailed core design and neutronics analysis with near-critical configuration and burnup calculation are carried out. • Detailed thermal-hydraulics analysis with turbulent flow and heat transfer is carried out. The fluoride-salt-cooled high-temperature advanced reactor (FuSTAR) is an advanced small modular reactor, which aims to the electric supplement and clean heat production to western China. In order to have good economy and safety, FuSTAR needs to have the characteristics of small size, light weight, deep burnup, and inherent safety. A conceptual design of FuSTAR is proposed in this study with neutronics and thermal-hydraulics assessments. In this study, its concept is established with two-region fuel to flatten core power, and the neutron spectrum of the core is in the thermal spectrum to improve the utilization of thermal neutrons and reduce fuel inventory. Neutronics and thermal-hydraulics analyses are also carried out, and established the core layout and material ratio under the near-critical condition. Burnup depth of the fuel, power distribution, neutron flux distribution, flow and heat transfer in rod-bundles channel, and 1/6 core porous media analysis are also investigated. The parameters of the designed core are evaluated quantitatively from various aspects. The results show that the core design in this study has sufficient shutdown depth, the equilibrium cycle is 3.24 years, and the discharge burnup depth is 106 GWd/MtU. The helical cruciform fuel (HCF) adopted by this reactor has a better flow and heat transfer capacity than cylindrical-type fuel at the inner channel. The analysis of 1/6 core porous media shows that the peak fuel temperature under normal operating condition is 1009.3 K, which is far below the limit of 1573 K. Relevant design and analysis results can provide the reference for the subsequent small modular FHR design and construction. [ABSTRACT FROM AUTHOR]

Summary: Closed type assembly leads to complicated core thermohydraulics which are significant for design and safety analysis in sodium‐cooled fast reactor (SFR). Most of the major phenomena can be dealt with by general parallel channel network and two‐dimensional model for inter‐wrapper flow (IWF) in system code transient thermal‐hydraulic code for analysis of SFR (THACS). A main deficiency left is the overestimation of inter‐assembly heat transfer effect due to lump temperature message that one‐dimensional assembly model can only provide. A methodology was thus developed to compute wall‐subchannel temperature for simulating energy transportation through duct wall. Intra‐assembly thermohydraulics were classified as two mechanisms and non‐dimensional parameter was introduced for characterizing each of them. Ratio between linear heat flux density through hexagonal wrapper surface and pin surface qw/p was employed to account for radial heat transfer. Flow redistribution effect arouse in mixed and natural convection was concretized by a importance symbol of buoyancy force, Richardson number Ri. Two criteria were also proposed for dividing flow state into three parts and specific model for each section was defined. Two validation cases against 3‐D computational fluid dynamics (CFD) simulation of China experimental fast reactor whole core and test data from plant dynamics test loop with direct heat exchanger experiment were performed with THACS. Noticeable improvement after incorporation of this model demonstrated its value in system analysis. Although the procedure for parameter computations is geometry related, this method can be reproduced to any type of assembly directly. [ABSTRACT FROM AUTHOR]

Summary: To evaluate residual heat removal capacity of next generation China sodium‐cooled fast reactor (SFR) and provide data for code validation, an integral natural circulation experiment was performed on a scaled water platform which had same structure with reactor primary system. A modular code THACS focusing on safety analysis of SFR was validated by this test. Some necessary development and verification for physical property, frictional loss, and heat transfer of water were carried out first. Whole simulation of an unprotected loss‐of‐flow (ULOF) case was divided into two phases and computation lasted until 5000 s of test time during second stage. Agreement in light of system temperature distribution in first stage provided a suitable initial condition for subsequent transient. Satisfied flow rate curve was reproduced with decaying pressure head and valve characteristic as input when pump was stopping. All natural circulation paths were predicted, including original flow in primary loop through main pipe and reversed flow in reactor vessel cooling system (RVCS). Transition to operative mode of direct heat exchanger (DHX) in direct reactor auxiliary cooling system (DRACS) and its cooling power were also forecasted successfully. This work means a significant advance for THACS in the field of engineering application of China SFR. [ABSTRACT FROM AUTHOR]

Thermal characteristics and mechanical analysis are particularly essential in the performance and safety analysis of metallic fuel for Liquid Metal Fast Reactor (LMFR). Therefore, the development of advanced thermo-mechanical analysis code is significant for the safety design of metallic fuels. In order to meet the future demand for metallic fuel application in China, the FRAC code was developed to analyze the metallic fuel performance during steady-state conditions. The code was benchmarked against the open-literature EBR-II experimental database. In comparison with the experimental data given, the code calculated Zr mole fraction, fuel temperature distribution, fission gas release and clad strain distribution are in good agreement with the data, which indicates that the code is reliable and can be applied into safety analysis of metal fuel design in China. The overall code assessment shows that the prediction error is within acceptable range. [ABSTRACT FROM AUTHOR]

Summary: Waste heat recovery helps reduce energy consumption, decreases carbon emissions, and enhances sustainable energy development. In China, energy‐intensive industries dominate the industrial sector and have significant potential for waste heat recovery. We propose a novel waste heat recovery system assisted by a heat pipe and thermoelectric generator (TEG) namely, heat pipe TEG (HPTEG)，to simultaneously recover waste heat and achieve electricity generation. Moreover, the HPTEG provides a good approach to bridging the mismatch between energy supply and demand. Based on the technical reserve on high‐temperature heat pipe manufacturing and TEG device integration, a laboratory‐scale HPTEG prototype was established to investigate the coupling performances of the heat pipes and TEGs. Static energy conversion and passive thermal transport were achieved with the assistance of skutterudite TEGs and potassium heat pipes. Based on the HPTEG prototype, the heat transfer and the thermoelectric conversion performances were investigated. Potassium heat pipes exhibited excellent heat transfer performance with 95% thermal efficiency. The isothermality of such a heat pipe was excellent, and the heat pipe temperature gradient was within 15°C. The TEG's thermoelectric conversion efficiency of 7.5% and HPTEG's prototype system thermoelectric conversion efficiency of 6.2% were achieved. When the TEG hot surface temperature reached 625°C, the maximum electrical output power of the TEG peaked at 183.2 W, and the open‐circuit voltage reached 42.2 V. The high performances of the HPTEG prototype demonstrated the potential of the HPTEG for use in engineering applications. [ABSTRACT FROM AUTHOR]

The Integrated Pressurized Water Reactor (IPWR) designed with series of passive safety system has been regarded as a very promising reactor concept. In this paper, in order to assess the accident mitigation ability of passive safety systems in China 100 MW IPWR, the complete system models, including the reactor vessel, passive safety injection system and passive residual heat removal system, were established using RELAP5 software. The comprehensive and quantitative comparative analysis of auxiliary feed water system and passive safety system under typical accident scenarios, including loss of flow accident and Station Black-out (SBO), were investigated, respectively. Results show that the passive residual heat removal system has the similar accident mitigation capability compared with the auxiliary feed water system, which is adopted in most of traditional PWR design. And the IPWR could realize the reactor safe shutdown under typical accident scenarios with passive safety systems. Finally, the Small break loss of coolant accident (SBLOCA) induced by pressurizer surge line rupture was calculated to verify the function of passive safety systems for IPWR. Results demonstrate that the passive safety system could provide emergency coolant to the reactor and remove the reactor decay heat successfully. This work is meaningful for the quantitative evaluation of passive safety systems in IPWR. [ABSTRACT FROM AUTHOR]

An innovative Direct Residual Heat Removal System (DRHRS) is proposed for Pressurized Water Reactor (PWR) in this paper. The new designed parallel DRHRS is different from traditional Passive Residual Heat Removal System (PRHRS), which is connected to steam generation. The thermal hydraulic transient analysis of the new designed DRHRS for CPR1000 has been carried out using the widely accepted safety analysis software RELAP5. The new designed DRHRS is directly connected to the primary loop, which consists of three independent parallel loops, three intermediate cooling circuits and an air loop. The transient behaviors of passive safety system are studied, and design parameter sensitivity analysis is carried out. Results show that during Station Black_Out (SBO) accident, natural circulations are established stably in passive safety system so that core decay is continuously removed from primary loop. And the new designed DRHRS has the capability of removing residual heat to the atmosphere without any external energy input at different surrounding environmental temperature. In emergency, the DRHRS directly remove core decay heat from reactor outlet, and efficiency of residual heat removal is improved. Moreover, reactor power plant maintains safe even if double-ended rupture of a single tube during SBO accident occurs. Thus, the designed DRHRS has great significance for increasing the degree of inherent safety features of CPR1000. [ABSTRACT FROM AUTHOR]

Interest in evaluation of reactor response and off-site consequences following beyond design basis accidents has significantly increased after Fukushima. MELCOR is an advanced computational aid that has been widely used and adapted to various reactor designs for severe core damage analyses. This paper summarizes the core thermal hydraulic response for a hypothetical severe accident caused by station blackout with failure of the steam generator safety relief valve at a Chinese pressurized reactor 1000-MW power plant – CPR1000 using MELCOR. Analytical results for a) temperature distribution of the fuel pellets, b) the fuel cladding, c) flow rate of the coolant, and, d) hydrogen mass changing over time are presented. The analyses are focused on safety assessment of the reactor core for severe accidents and are part of the overall evaluation of safety features of the CPR1000 reactor for residual risk posed by severe accidents. [ABSTRACT FROM AUTHOR]

Abstract: The concept of passive safety is put forward to improve the safety and reliability of the nuclear power plant. The secondary emergency passive residual heat removal system (EPRHRS) is an innovative design for China Pressurized Reactor (CPR1000), which can remove the core residual heat effectively and improve the inherent safety by passive methods when the station blackout (SBO) occurs. In this work, the EPRHRS, which is connected to the steam generator (SG) and consists of a heat exchanger (HX), a water tank and corresponding pipes and valves, was designed for CPR1000. The primary loop and EPRHRS were modeled using RELAP5/MOD3.4 to investigate the residual heat removal capability of the EPRHRS and the transient characteristics of the primary loop system in the event of the SBO accident. In addition, sensitivity studies were also performed to investigate effects of the main parameters of the EPRHRS on the transient behavior of the primary loop and the EPRHRS. The effect of isolation valve opening time on the primary loop and the EPRHRS was also studied in the paper. The results of the study indicate that the optimized HX area is 200m2 when the EPRHRS water tank height is 10m and the optimized isolation valve opening time is 30s. The EPRHRS can remove the core decay heat from the primary loop effectively through the stable natural circulation established in the primary loop and the EPRHRS loop based on the calculation results. Therefore, it illustrates that the EPRHRS designed in the paper is effective and meaningful to CPR1000. [Copyright &y& Elsevier]

• The PIRT of SBO for China pool type SFR has been developed by 15 experts. • A math method to evaluate different opinion about PIRT results has been proposed. • The results could be developed V&V matrix and a better understanding for SFR. In order to satisfy analysis of safety requirements for China pool type sodium-cooled fast reactor, a phenomena identification and ranking table (PIRT) is developed for station blackout (SBO) accidents by 15 experts. The SBO accidents could be divided into two phases, i.e. phase 1: force circulation and phase 2: natural circulation, and the most important phenomena but maximum disagreements among experts are the gap conductance model between fuel and clad, the inter wrapper flow, the heat transfer between coolant and inter wrapper. Those phenomena should be carefully assessed and done more theorical and experiment analysis to develop more accurate and realistic model, to reduce uncertainty of safety analysis. The PIRT could be also used for developing verification and validation matrix for thermal hydraulic safety code, guiding researchers and designers to conduct experiments to close technological gap and giving related researcher and designer a comprehensive understanding for pool-type sodium-cooled fast reactor. [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 core residual heat removal system of the pressurized water reactor (PWR) is a vital facility to ensure the reactor safety, and its failure after shutdown will inevitably lead to the meltdown of the core. However, the existing core residual heat removal system of CPR1000 only guarantee the 72-hour grace period, after which the core decay heat can still melt the core within one day. The present study proposes an improved design of passive residual heat removal system (PRHRS) that utilizes natural forces to realize a passive self-start. Meanwhile, an innovative method is also proposed to use the atmosphere as the final heat sink in this paper. Subsequently a numerical study is conducted to verify the performance of the improved design during accident condition. Numerical results show that the improved design of PRHRS maintains a timeless safety cooling capacity during the station blackout (SBO) scenario. Finally, the parameter sensitivity study is performed to investigate a reasonable design value of chimney height. And numerical results prove that the proposed PRHRS is still capable of removing the core residual heat continuously even if the atmosphere temperature is variable. All component functions of new designed PRHRS are completely driven only by the action of natural factors without any external energy power. Therefore, the improved design of PRHRS can actually eliminate the risk of severe accident and lays the necessary technical foundation for the faster and larger-scale development of PWR. [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]

• An aerosol module based on the aerosol dynamic and hygroscopic models was added to the ATHROC code. • Three different types of benchmark experiments were selected for assessment of the improved code. • The comparative results show that the code has capabilities to accurately predict the experiments. As the main carrier of fission products, aerosols will transport in the containment and even leak into the environment during accident conditions, resulting in radioactive pollution. Thus, it is of great significance for radiation protection to study the aerosol distribution mitigation measures in containment after severe accidents. A multi-compartment lumped parameter code named ATHROC (Analysis of Thermal Hydraulic Response of Containment) was developed to analyze the thermal–hydraulic behaviors in the containment after accidents for China advance PWR. In order to satisfy the analysis of aerosol behavior in containment, an aerosol module based on the aerosol dynamic and hygroscopic models was added to the code and coupled with spray removal mechanism models. Three different types of benchmark experiments were selected for assessment of the improved code, including CVTR, LACE and CSE tests. Most of the predicted results of thermal–hydraulic responses and aerosol distribution are in realistic agreement with the experimental data. The comparative results show that the code has capabilities to accurately predict the thermal–hydraulic responses during spray activation, aerosol distribution and removal of aerosols by spray droplets. And the problems identified in the code will be further optimized in the future. Currently, the developed improved code satisfies the analysis of aerosol behavior in containment and provides a design for aerosol removal for spray system arrangement. [ABSTRACT FROM AUTHOR]

Performance of the separation system in the UTSG determines the quality of supplied steam in a nuclear power plant. In the present study, the Eulerian-Lagrangian methodology is applied in a full-scale UTSG separation system of GEN-III China advanced PWR. Detailed flow characteristics and separation performance of the whole system are investigated. Effects of the steam velocity and droplets sizes on the separation process are also discussed. Last but not least, the mal-distribution phenomena in dryers are analyzed. Results illustrate that the separators at the hot side have a better performance than the ones at the cold side, and the separation efficiency falls within the range from 98.16% to 99.39% in the current system. What's more, both the steam velocity and the droplets radii have obvious effects on the separation performance. For dryers, obvious mal-distribution phenomena are estimated. As a result of the lower efficiency of separators at cold side, the #7 dryer suffers the maximum loading of entrained water on unit-area (1.24kg/(m2·s)), and the greatest humidity at entrances of dryers is observed to be 4.90% accordingly. In general, the present work could be expected to be the assessment and basis of UTSG separation system optimization for China advanced PWR. • Full-scope multiphase-flow characteristics in separation system of China advanced PWR. • Overview on the performance of the separation system. • Mal-distribution phenomena of dispersed drops at entrances of dryers. [ABSTRACT FROM AUTHOR]