期刊文献+
共找到6篇文章
< 1 >
每页显示 20 50 100
Indirect neutron radiography experiment on dummy nuclear fuel rods for pressurized water reactors at CMRR
1
作者 Yong Sun Qi-Biao Wang +11 位作者 Peng-Cheng Li Ming Xia Bin Liu He-Yong Huo Wei Yin Yang Wu Sheng Wang Chao Cao Xin Yang Run-Dong Li Hang Li Bin Tang 《Nuclear Science and Techniques》 SCIE EI CAS CSCD 2024年第11期13-27,共15页
Nuclear energy is a vital source of clean energy that will continue to play an essential role in global energy production for future generations.Nuclear fuel rods are core components of nuclear power plants,and their ... Nuclear energy is a vital source of clean energy that will continue to play an essential role in global energy production for future generations.Nuclear fuel rods are core components of nuclear power plants,and their safe utilization is paramount.Due to its inherent high radioactivity,indirect neutron radiography(INR)is currently the only viable technology for irradiated nuclear fuel rods in the field of energy production.This study explores the experimental technique of indirect neutron computed tomography(INCT)for radioactive samples.This project includes the development of indium and dysprosium conversion screens of different thicknesses and conducts resolution tests to assess their performance.Moreover,pressurized water reactor(PWR)dummy nuclear fuel rods have been fabricated by self-developing substitute materials for cores and outsourcing of mechanical processing.Experimental research on the INR is performed using the developed dummy nuclear fuel rods.The sparse reconstruction technique is used to reconstruct the INR results of 120 pairs of dummy nuclear fuel rods at different angles,achieving a resolution of 0.8 mm for defect detection using INCT. 展开更多
关键词 Conversion screen DYSPROSIUM Indirect neutron computed tomography Dummy nuclear fuel rods
下载PDF
Multiphysics simulation of VVER-1200 fuel performance during normal operating conditions 被引量:2
2
作者 Khaled M.Yassin Mohamed H.Hassan +3 位作者 Mohammad M.Ghoneim Mostafa S.Elkolil Adel Alyan Said A.Agamy 《Nuclear Science and Techniques》 SCIE EI CAS CSCD 2023年第2期139-152,共14页
Nuclear fuel performance modeling and simulation are critical tasks for nuclear fuel design optimization and safety analysis under normal and transient conditions.Fuel performance is a complicated phenomenon that invo... Nuclear fuel performance modeling and simulation are critical tasks for nuclear fuel design optimization and safety analysis under normal and transient conditions.Fuel performance is a complicated phenomenon that involves thermal,mechanical,and irradiation mechanisms and requires special multiphysics modules.In this study,a fuel performance model was developed using the COMSOL Multiphysics platform.The modeling was performed for a 2D axis-symmetric geometry of a UO2fuel pellet in the E110 clad for VVER-1200 fuel.The modeling considers all relevant phenomena,including heat generation and conduction,gap heat transfer,elastic strain,mechanical contact,thermal expansion,grain growth,densification,fission gas generation and release,fission product swelling,gap/plenum pressure,and cladding thermal and irradiation creep.The model was validated using a code-to-code evaluation of the fuel pellet centerline and surface temperatures in the case of constant power,in addition to validation of fission gas release(FGR)predictions.This prediction proved that the model could perform according to previously published VVER nuclear fuel performance parameters.A sensitivity study was also conducted to assess the effects of uncertainty on some of the model parameters.The model was then used to predict the VVER-1200 fuel performance parameters as a function of burnup,including the temperature profiles,gap width,fission gas release,and plenum pressure.A compilation of related material and thermomechanical models was conducted and included in the modeling to allow the user to investigate different material/performance models.Although the model was developed for normal operating conditions,it can be modified to include off-normal operating conditions. 展开更多
关键词 VVER-1200 fuel performance COMSOL code Zr-1%Nb cladding UO2 fuel rod
下载PDF
Seismic and stress qualification of LMFR fuel rod and simple method for the determination of LBE added mass effect 被引量:1
3
作者 M.Khizer Jian-Wei Chen +3 位作者 Guo-Wei Yang Qing-Sheng Wu Yong Song Yong Zhang 《Nuclear Science and Techniques》 SCIE CAS CSCD 2020年第1期42-56,共15页
In this study, two different designs of liquid metal fast reactor(LMFR) fuel rods wire-wrapped and nonwire-wrapped(bare) are compared with respect to different parameters as a means of considering the optimum fuel des... In this study, two different designs of liquid metal fast reactor(LMFR) fuel rods wire-wrapped and nonwire-wrapped(bare) are compared with respect to different parameters as a means of considering the optimum fuel design. Nuclear seismic rules require that systems and components that are important for safety must be capable of bearing earthquake effects, and that their integrity and functionality should be guaranteed. Mode shapes, natural frequencies, stresses on cladding, and seismic aspects are considered for comparison using ANSYS. Modal analysis is compared in a vacuum and in lead–bismuth eutectic(LBE) using potential flow theory by considering the added mass effect. A simple and accurate approach is suggested for the determination of the LBE added mass effect and is verified by a manually calculated added mass, which further proved the usefulness of potential flow theory for the accurate estimation of the added mass effect. The verification of the hydrodynamic function(τ) over the entire frequency range further validated the finite element method(FEM) modal analysis results. Stresses obtained for fuel rods against different loading combinations revealed that they were within the allowable limits with maximum stress ratios of 0.25(bare) and 0.74(wire-wrapped). In order to verify the structural integrity of cladding tubes, stresses along the cladding length were determined during different transients and were also calculated manually for static pressure. The manual calculations could be roughly compared with the ANSYS results, and the two showed a close agreement. Contact analysis methodology was selected,and the most appropriate analysis options were suggested for establishing contact between the wire and cladding for the wire-wrapped design grid independence analysis,which proved the accuracy of the results, confirmed the selection of the appropriate procedure, and validated the use of the ANSYS mechanical APDL code for LMFR fuel rod analysis. The results provided detailed insight into the structural design of LMFR fuel rods by considering different structural configurations(i.e., bare and wire-wrapped) in the seismic loading;this not only provides a FEM procedure for LMFR fuel with complex configuration, but also guides the reference design of LMFR fuel rods. 展开更多
关键词 LMFR fuel rod Added mass Seismic analysis Contact analysis
下载PDF
Development and validation of a new oxide fuel rod performance analysis code for the liquid metal fast reactor
4
作者 Guang-Liang Yang Hai-Long Liao +1 位作者 Tao Ding Hong-Li Chen 《Nuclear Science and Techniques》 SCIE EI CAS CSCD 2022年第5期167-177,共11页
The integrity and reliability of fuel rods under both normal and accidental operating conditions are of great importance for nuclear reactors.In this study,considering various irradiation behaviors,a fuel rod performa... The integrity and reliability of fuel rods under both normal and accidental operating conditions are of great importance for nuclear reactors.In this study,considering various irradiation behaviors,a fuel rod performance analysis code,named KMC-Fueltra,was developed to evaluate the thermal–mechanical performance of oxide fuel rods under both normal and transient conditions in the LMFR.The accuracy and reliability of the KMC-Fueltra were validated by analytical solutions,as well as the results obtained from codes and experiments.The results indicated that KMC-Fueltra can predict the performance of oxide fuel rods under both normal and transient conditions in the LMFR. 展开更多
关键词 fuel rod analysis code Thermal-mechanical performance Irradiation behaviors Pellet-cladding mechanical interaction Liquid metal fast reactor
下载PDF
Irradiation Testing of Coated Particle Fuel at HANARO
5
作者 Bong Goo Kim Moon Sung Cho Yong Wan Kim 《Journal of Energy and Power Engineering》 2014年第10期1740-1747,共8页
TRISO (Tri-structural iso-tropic)-coated particle fuel is being developed to support the development of a VHTR (very high temperature reactor) in Korea. From August 2013, the first irradiation testing of coated pa... TRISO (Tri-structural iso-tropic)-coated particle fuel is being developed to support the development of a VHTR (very high temperature reactor) in Korea. From August 2013, the first irradiation testing of coated particle fuel was begun to demonstrate and qualify TRISO fuel for use in the VHTR in HANARO (high-flux advanced neutron application reactor) at KAERI (Korea Atomic Energy Research Institute). This experiment is currently undergoing under an atmosphere of a mixed inert gas without on-line temperature monitoring and control combined with on-line fission product monitoring of the sweep gas. The irradiation device contains two test rods, one contains nine fuel compacts and the other five compacts and eight graphite specimens. Each compact has 263 coated particles. After a peak bum-up of about 4% and a peak fast neutron fluence of about 1.7 × 1021 n/cm2, PIE (post irradiation examination) will be carried out at KAERI's irradiated material examination facility. This paper describes the characteristics of coated particle fuels, and the design of the test rod and irradiation device for the coated particle fuels, and discusses the technical results of irradiation testing at HANARO. 展开更多
关键词 TRISO coated particle fuel fuel compact test fuel rod IRRADIATION irradiation device HANARO.
下载PDF
Revisiting Stainless Steel as PWR Fuel Rod Cladding after Fukushima Daiichi Accident
6
作者 Alfredo Abe Claudia Giovedi +1 位作者 Daniel de Souza Gomes Antonio Teixeira e Silva 《Journal of Energy and Power Engineering》 2014年第6期973-980,共8页
In the past, stainless steel was utilized as cladding in many PWRs (pressurized water reactors), and its performance under irradiation was excellent. However, stainless steel was replaced by zirconium-based alloy as... In the past, stainless steel was utilized as cladding in many PWRs (pressurized water reactors), and its performance under irradiation was excellent. However, stainless steel was replaced by zirconium-based alloy as cladding material mainly due to its lower neutron absorption cross section. Now, stainless steel cladding appears as a possible solution for safety problems related to hydrogen production and explosion as occurred in Fukushima Daiichi accident. The aim of this paper is to discuss the steady-state irradiation performance using stainless steel as cladding. The results show that stainless steel rods display higher fuel temperatures and wider pellet-cladding gaps than Zircaloy rods and no gap closure. The thermal performance of the two rods is very similar and the neutron absorption penalty due to stainless steel use could be compensating by combining small increase in U-235 enrichment and pitch size changes. 展开更多
关键词 Austenitic stainless steel cladding Zircaloy cladding PWR fuel rod steady-state fuel performance codes.
下载PDF
上一页 1 下一页 到第
使用帮助 返回顶部