Abstract: This paper introduces the basic situation and the latest progress of nuclear power units in China, as well as the related measures of improving the safety level of nuclear facilities. Under the requirement of the "national energy administration’s 13 th five-year plan for energy technology innovation...
Abstract: For the heat pipe cooled space reactor, region-dependent homogenized cross sections in the predefined 26 group structure were generated with the OpenMC code based on the R-Z geometric model of the reactor core. The neutron transport calculation was performed with SARAX, which was a deterministic neu...
Abstract: sm-TMSR is a small modular multi-purpose thorium-based molten salt demonstration reactor designed by Center for Thorium Molten Salt Reactor System (TMSR), CAS. In this paper, the physical analysis of the control rod arrangement was carried out. Firstly, we analyzed the unique reactivity variation in...
Abstract: A numerical simulation program for the dynamic distribution of xenon with flow and on-line removal function was established for the primary loop system of molten salt reactor（MSR） based on Mathematica7.0. The dynamic characteristics of xenon concentration with time under different flow rates, differ...
Abstract: This paper investigates the load-following characteristics of an AMTEC converted and heat pipe cooled space reactor power system. The effects of the external load and AMTEC temperature on module AMTEC performance were analyzed, and the response of the AMTEC space reactor power system to a load deman...
Abstract: CENTER fuel assembly irradiation test should be carried out before the fuel assembly is formally designed, and the CENTER fuel assembly would be irradiated in HFETR to complete the fuel burn-up test. To ensure the success of the irradiation test and meet the test requirements, the accuracy of power ...
Abstract: In order to study the application of thorium uranium fuel in the CANDU6 reactor, the code DRAGON/DONJON is adopted to study the time-averaged equilibrium core of CANDU 6 with 37-element bundle assembly of discrete thorium uranium fuel rods. The results show that, when the assembly is the uranium rod...
Abstract: In order to predict the bubble departure diameter in a horizontal rectangle heating mini-channel to study its heat transfer characteristics, a bubble departure diameter predicting model based on force balance is proposed. A visualization experiment is used to verify it, and the results show that the...
Abstract: Reactor with 177-fuel-assembly core usually operates in G mode and the core boron concentration requires to be adjusted during the load following. Limited by the capacity of the boron recycle system, the load follow capability is available only for 85% of the lifetime. In order to improve the load-f...
Abstract: Based on the test results of the domestic tin Zr-4 C zirconium alloy in low alloy autoclave in the pure water and LiOH aqueous corrosion, using the transmission electron microscopy（TEM） observation matrix and oxide film microstructure, through the analysis of oxidation weight data, the corrosion mec...
Abstract: A preliminary study is conducted for the preparation technology of UO2-Er2O3 burnable poisonous fuel pellets with Er2O3 mass fraction of 4.32%. We obtain the preparation technology by comparing the performance（integrity, density, and grain size） of the pellets under different conditions（mixing, pres...
Abstract: The response characteristics of the fuel rods in the sub-critical reactor of China Initiative Accelerator Driven System（CiADS） under beam trip were simulated by the reactor system analysis program RELAP5 mod4.0. The fatigue life of the fuel cladding in CiADS under beam trip is calculated by ANSYS 17...
Abstract: SiC/Zr composites of different composition were prepared by vacuum sintering, using SiC powder and ZrH2 powder as the raw material. The effect of SiC addition on the microstructure and properties of SiC/Zr composite was studied. The results show that the compact and single Zr metal can be obtained w...
Abstract: The electrochemical behaviors of 316 NG stainless steel in the boron-lithium solutions（pH=5～8） at 300℃ were studied by using electrochemical polarization curves and the electrochemical impedance spectroscopy（EIS）, and the experimental E-pH diagram at the same environment was graphed. The results ind...
Abstract: A simplified methodology for assessing the shock and vibration effects of the aircraft impact on nuclear power plant（NPP） is discussed in this paper. Both the force time-history method（FTHM） and the missile-target interaction method（MTIM） are used to assess NPP shock response and its propagation. Th...
Abstract: Aiming at the weld overlay repair（WOR） for the upper Ω seal of the control rod drive mechanism（CRDM）, numerical simulation is applied in the integrity analysis of the overlay repaired structure. 2-D axisymmetric Gauss heat source equivalent input was built according to the welding parameter, and the...
Abstract: In order to obtain the response spectrum for the specific nuclear power plant（NPP）, considering the specific site condition and ground motion parameters, the stochastic simulation and probabilistic hazard analysis is combined to obtain the UHS which the probability of exceedance is 10-4. In order to...
Abstract: There were different defects possibly existing in the processing and installation of nuclear piping. In addition, there may be a small amount of defects such as cracks existing in the piping due to the effect of operation conditions in the nuclear power plants. The residual life of nuclear piping wi...
Abstract: This paper adopts the method of vacuum exhaust in the primary loop of PWR nuclear power unit. Primary loop of PWR Ling’ao Nuclear Power Plant is vacuum exhausted after refueling overhaul, and the dynamic exhaust process is cancelled. The results show that the requirements of the air content of the p...
Abstract: In order to quantitatively evaluate operator’s response implementation reliability in digital main control rooms（MCRs） of nuclear power plants（NPPs）, main performance shaping factors（PSF） are identified by contextual environment analysis. Weight of the PSF are identified by analytic hierarchy proces...
Abstract: Current designs of medium voltage mobile power supply, as the improvement item after Fukushima nuclear accident, vary in various nuclear power projects due to different reactor types and site conditions. Based on the engineering practice experience obtained in the design of medium voltage mobile pow...
Abstract: Referring to the water quality standard of the pressurized water reactor（PWR） second circuit and the thermal power direct current furnace, combined with the structure and material characteristics of the second circuit of the high temperature reactor, this paper updated the original intake and outlet...
Abstract: Reactor core melt down accident may result in lower head failure, and molten material may relocate into the cavity, which endangers the safety of personnel and hull. The severe accident integration calculation program called MAAP4 code is adopted to study the effects of low pressure injection system...
Abstract: The design and structure specificities of loose part measurement system（LPMS） of VVER in Tianwan Nuclear Power Plant, as well as the differences of LPMS with the item NRC RG1.133, are introduced. The difficulties and the negative influence are analyzed based on these differences and the specificitie...
Abstract: In order to provide a seismic design foundation to the Equipment and Instrument installed in the pool which are related to nuclear safety, in connection with the rectangular pool related to nuclear safety, study the change regulation of the response spectrum on the pool wall according to a liquid-st...
Abstract: Taking a 1000 MW PWR as an example, a two-dimensional polar coordinate thermal model was used to analyze the coupling heat transfer among the wall surface of RPV, the two-layer melting core pool and the outer water chamber. The transient 2 D temperature and ablation of the bottom head wall surface w...
Abstract: In this paper, an analytical verification method for valve discharge under saturated steam is studied by taking the pressurizing system stop valve as an example. Comparison of the values of valve discharge test under small opening and low pressure difference with the theoretical calculation value sh...
Abstract: The stepping motion of the control rod drive mechanism is a complicated dynamic process combined with electromagnetic field, flow field, and motion. It is not accurate to analyze the process by static method or partial technique. Magnet software, Fluent software, and Adams software were combined to ...
Abstract: Nuclear vessel hydrostatic test is one of the important in-service inspection methods in nuclear power plants, to verify vessels under sustained pressure integrity and sealing. According to the installation requirements of the temporary special devices, in the design stage of vessel and pipeline of ...
Abstract: Containment vessel of AP1000 is designed to be built by modules. Based on the new lifting method called rigging lifting, further improvement to change the containment vessel from four rings to three rings was proposed, and the best location of lifting lugs was studied. The load ratio of the crane an...
Abstract: In order to improve the reliability of the fire damper in nuclear power plants, the application status and failure history of fire damper in several nuclear power plants were analyzed in this paper. The failure mechanism and maintenance strategy optimization of different fire dampers are studied by ...
Abstract: HPR1000 is the third generation NPP, and its reactor and primary loop system need to fulfil higher requirements for inherent safety. For the core cooling monitoring system（CCMS） of the second generation nuclear power plant, the bottom of the reactor shall be bored to measure the water level. This de...
Abstract: According to the environmental characteristics of nuclear power ships, a passive sampler with a volume ratio of ethylene glycol to tritium-free water of 1:1 is used to monitor the concentration of tritiated water in a closed nuclear power plant. After continuous monitoring for 1 month in different w...
Abstract: Based on the principle of human factors engineering HFE, this paper takes the secondary feed water deaerator system of a nuclear power plant as an example to analyze the performance requirements, obtain different levels of static function database, and determine the basic information flow and its pr...
Abstract: Dynamical analysis of the reactor structure under seismic load is a key procedure during the safety design of the entire reactor system. Due to stochastic and other uncontrollable errors during computation, manufacturing and installation, structural parameters usually subject to a certain amount of ...
Abstract: In this study, both of the minimum DNBR point and the BO point methods are applied to develop a new CHF correlation named the ACC correlation coupled with sub-channel code ATHAS. The statistics and predication rate are analyzed. The Owen criterion is used to determine the DNBR limit. The data analys...
Abstract: Severe accident release categories of a nuclear power plant of third generation were introduced. The release characteristics to the environment were calculated using MAAP code for those release categories and its severe accident sequences which can result in the large release of radioactive material...
Abstract: Based on the researches of the event alarm logic of loose parts and worldwide regulations and standards of loose parts event alarm and emergency response, combined with the loose parts event alarm and emergency response in a domestic nuclear power plant during hot state function test, the process of...
Abstract: When the results of the pump flow test of the low head safety injection system（LHSI） of CPR1000 do not satisfy the acceptance criteria, the throttle elements in the pipeline system need to be adjusted. The adjustment method of the lower limit flow orifice plate can be obtained through the calculatio...
Abstract: During the outage of a nuclear power plant, a loose parts impact test was carried out on the steam generator, and the main defect modes and treatment methods of the loose component monitoring system（LPMS） in long-term operation are proposed. The blind area in the function of fault self-checking in L...
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