Abstract: Under the new wave of global artificial intelligence, the nuclear energy industry has gradually started the process of integrating with the development of artificial intelligence. This paper discusses some problems arising from the combined application of artificial intelligence and nuclear energy. ...
Abstract: The fuel element burnup history becomes extremely complex as a result of the multi-pass of the fuel pebble through the reactor core in fuel management of the pebble-bed high-temperature gas-cooled reactor (PB-HTGR). The core physical design program VSOP of PB-HTGR can provide a fine burnup history o...
Abstract: In order to improve the applicability of deterministic whole-core neutron transport code, a general-purpose neutron transport code VITAS is developed. The verification results of TAKEDA3 benchmark problem (rectangular assembly), the TAKEDA4 benchmark problem (hexagonal assembly), the Dodds benchmark...
Abstract: The adjoint neutron flux is of great significance for nuclear safety and detector calculation in pressurized water reactor (PWR). However, existing nodal methods would cause a big error due to heterogeneous nodes, including heterogeneous cross sections and discontinuity factors, which will appear fr...
Abstract: The refueling cycle calculation function for pressurized water reactor (PWR) is developed on the basis of the self-developed numerical nuclear reactor physics calculation code NECP-X. Startup physics experiments are conducted for the first, second and third cycles of an M310 reactor, and fine modeli...
Abstract: The positioning and wrapping wire design is widely applied in the core design of metal cooled fast reactor and gas cooled fast reactor. In this paper, the effects of pitch, number and shape of positioning and wrapping wires on the flow and heat transfer of supercritical carbon dioxide in rod bundle ...
Abstract: Narrow rectangular channel is widely used in various fields because of its compact structure and large heat transfer area. The safety and economy of reactor can be improved by improving the prediction method of critical heat flux (CHF) in the narrow rectangular channel and establishing a CHF mechani...
Abstract: After the emergency shutdown of the lead-bismuth fast reactor, the thermal stratification in the upper plenum has an important impact on the integrity of the reactor structure and the residual heat removal capacity of the natural circulation, which requires special attention. In order to overcome th...
Abstract: High temperature gas cooled reactor combined with magnetohydrodynamic (MHD) power generation is an efficient space power system. It can meet the requirements in space tasks for high power and high efficiency and thus has broad application prospects. In this paper, a core scheme composed of 217 fuel ...
Abstract: In order to resolve the drawbacks of flow regime map used in the one-dimensional two-fluid model based nuclear power plant system analysis code and improve the accuracy of the system analysis code, this paper explores the application of the interfacial area transport equation (IATE) in the one-dimen...
Abstract: Bubble growth in rectangular narrow channel can directly change the phase interfacial area concentration and thus affect the flow channel's heat and mass transfer performance. In order to obtain a model for different types of bubble growth in a narrow flow channel, wall boiling flow heat transfer ex...
Abstract: In order to explore the influence of bionic guide vane on the overall performance of the main pump, a new bionic structural design of guide vane is proposed in this paper by taking the scale model (1:2.5) of CAP1400 main pump as the study object, and an optimized model (the optimal solution of the b...
Abstract: In order to explore the characteristics of steam critical flow leakage from a small break in pipeline during the loss of coolant accident of pressurized water reactor (PWR) nuclear power plant, small-break leakage experiments of pipelines are carried out in this paper to explore the characteristics ...
Abstract: Chromium (Cr) coated zirconium alloy cladding is considered as one of the most promising cladding materials for accident tolerant fuel (ATF). The degree of influence of the surface state of this material on the heat transfer performance will greatly affect the process optimization direction of coate...
Abstract: In the engineering field, the RANS turbulence models are often used for thermal and hydraulic numerical simulation. However, the liquid lead-bismuth eutectic (LBE) has unique thermophysical properties, and the applicability of conventional turbulent Prandtl number models and RANS turbulence models t...
Abstract: In order to ascertain the effect of ion irradiation dose and temperature on the hardness, Young's modulus and microstructure of nuclear-grade graphite, 0.02 dpa, 0.2 dpa and 2 dpa C4+ are used in this paper to irradiate nuclear-grade graphite at room temperature and 180℃, respectively. The propertie...
Abstract: The seismic behavior of fuel assembly, as Class I seismic item, is directly related to the operation safety of nuclear power plants. It is usually necessary to verify the reasonableness of the seismic analysis method for reactor fuel assembly through seismic test. By simulating the actual reactor co...
Abstract: In order to study the influence of neutron irradiation on the mechanical properties of Cr-coated zirconium alloy, a neutron irradiation test is carried out in this paper for the Cr-coated zirconium alloy prepared by multi-arc ion plating technology. The mechanical properties were conducted by in sit...
Abstract: The Fukushima nuclear accident in Japan in 2011 exposed the inherent safety problems of traditional zirconium alloy fuel cladding under LOCA conditions. To investigate the performance of a new Cr-coated zirconium alloy cladding under LOCA conditions, high temperature steam oxidation and quenching ex...
Abstract: In order to accurately explore the influence of fluid-structure interaction behavior between reactor coolant and fuel assembly on the vibration characteristics of fuel assembly, this paper takes the pressurized water reactor (PWR) fuel assembly as the research object, applies the computational fluid...
Abstract: In order to determine the accident tolerant fuel (ATF) element design schemes for future high performance pressurized water reactor (PWR), this study comprehensively analyzes several potential ATF design schemes from the perspectives of safety, economical efficiency and fuel performance by using the...
Abstract: Fuel storage canister is a key equipment in the fuel supply system for high temperature reactor (HTR). In order to explore the optimal design scheme, a structure-performance-cost integration multi-objective optimization design method for fuel storage canister is proposed as follows: select the struc...
Abstract: The predicted value of the opening displacement of the circumferential through-wall crack (CTWC) in piping under different load levels is a critical parameter for the application of the leak-before-break (LBB) technology. In this paper, both numerical analysis and comparative verification are adopte...
Abstract: As an important feature of digital instrument control system, online self-diagnostics plays an important role in the reliability analysis of shutdown function of reactor protection system (RPS) in nuclear power plant. By analyzing the influence of self-diagnostics on human factors and periodic testi...
Abstract: In the design of nuclear power plant, the safety classification of items is carried out to ensure that the design, manufacturing and construction of the items meet appropriate requirements and that the reliability consistent with their functions can be achieved. This paper briefly describes the meth...
Abstract: The friction properties and service life of the secondary seal are the key factors affecting the service life of the hydrodynamic shaft seal of the reactor coolant pump. In order to study the long-term service life of the secondary seal, a high-frequency reciprocating test device is established to s...
Abstract: In order to explore the hydrodynamic characteristics of transient process of the reactor coolant pump shaft stuck accident, a full three-dimensional simplified model of the reactor primary circuit system was established by dynamically matching the hydraulic characteristics of the reactor coolant pum...
Abstract: The analog control system of Daya Bay Nuclear Power Plant consists of discrete electrical components integrated by hard wiring, and it is planned to undergo digital transformation during the 30a overhaul. Due to the functional limitation of the analog platform, the single failure of equipment has a ...
Abstract: In view of a number of titanium tube cracking events caused by steam flow induced vibration of a certain type of nuclear power condenser under half-side condenser operation conditions, this paper uses the computational fluid dynamics (CFD) method based on porous medium model to conduct a full three-...
Abstract: The cumulative effective dose to fuel assembly repair workers, the total radioactive activity of gaseous effluent released to the environment and the impact on gaseous effluent discharge monitoring are calculated and evaluated by taking the radioactive substances released from single fuel rod damage...
Abstract: P280GH carbon steel pipe is widely used in the main feedwater system of Hualong One unit, and octadecylamine (ODA) is applied to form a corrosion inhibition film by adsorption on the inner wall of the pipe. ODA can effectively inhibit corrosion and avoid pipe failure and serious scaling of steam gen...
Abstract: In order to realize the reasonable and compact general layout of the main plant building for 2 MW thorium based molten salt experimental reactor (TMSR) with liquid fuel, the overall design characteristics of the main plant building are determined in this paper according to the type characteristics, ...
Abstract: This paper studies the complex two-phase thermal hydraulic process after the opening of the pressurizer safety valve so as to determine the two-phase discharge characteristics of the pressurizer safety valve under different initial events. In this paper, the autonomous system analysis program ARSAC ...
Abstract: To meet the development need of advanced nuclear systems, an ultra-high flux reactor (UFR) core design concept is proposed in this paper. In this concept, plate-type fuel and square fuel assembly design is adopted, and a wide flow channel is provided to ensure a high volume share of the core coolant...
Abstract: Based on the annular fuel elements, a conceptual design of ultra-high flux reactor (UFR) is proposed. The fuel assembly design adopts a hexagonal assembly composed of 61 fuel elements. The core is designed with 52 boxes of fuel assemblies, 9 boxes of control rod assemblies and a thick reflecting lay...
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