The experimental tests of tensile for lead-flee solder Sn-3.5Ag were performed for the general work temperatures range from 11 to 90 ℃ and strain rate range from 5 × 10^-5 to 2 × 10^-2s^-1, and its stress--...The experimental tests of tensile for lead-flee solder Sn-3.5Ag were performed for the general work temperatures range from 11 to 90 ℃ and strain rate range from 5 × 10^-5 to 2 × 10^-2s^-1, and its stress--strain curves were compared to those of solder Sn-37Pb. The parameters in Anand model for solder Sn-3.5Ag were fitted based on experimental data and nonlinear fitting method, and its validity was checked by means of experimental data. Furthermore, the Anand model was used in the FEM analysis to evaluate solder joint thermal cycle reliability. The results show that solder Sn-3.5Ag has a better creep resistance than solder Sn-37Pb. The maximum stress is located at the upper right comer of the outmost solder joint from the symmetric center, and thermal fatigue life is predicted to be 3.796 × 10^4 cycles under the calculated conditions.展开更多
Small reactors have become a new hotspot of international nuclear energy research.The nuclear heating reactor(NHR)technology developed by Tsinghua University is an important multipurpose small reactor solution with fe...Small reactors have become a new hotspot of international nuclear energy research.The nuclear heating reactor(NHR)technology developed by Tsinghua University is an important multipurpose small reactor solution with features such as high integration,modular design and full power natural circulation.A new small reactor based on the existing NHR-200 reactor was developed by the Institute of Nuclear and New Energy Technology of Tsinghua University.A full-scale natural circulation test loop with the same operating parameters as the actual reactor was built in order to experimentally validate the natural circulation ability of the reactor primary loop and heat-transfer ability of fuel assemblies and heat exchangers.Corresponding results are given in detail,including parameter validation of the reactor primary loop,flow rules of the natural circulation and heat-transfer coefficients of heaters and heat exchangers,which can be directly used in the actual reactor as a reference for optimization design.Finally,a characteristic parameter k is proposed to represent the natural circulation ability of a system.By using the new data arrangement method in the form of parameter k,comprehensive experimental results of the natural circulation can be represented by a simple integrated expression.The work in this paper is of importance in broadening application fields and pushing forward commercialization of the NHR type reactors.展开更多
基金Project(50376076) supported by the National Natural Science Foundation of China
文摘The experimental tests of tensile for lead-flee solder Sn-3.5Ag were performed for the general work temperatures range from 11 to 90 ℃ and strain rate range from 5 × 10^-5 to 2 × 10^-2s^-1, and its stress--strain curves were compared to those of solder Sn-37Pb. The parameters in Anand model for solder Sn-3.5Ag were fitted based on experimental data and nonlinear fitting method, and its validity was checked by means of experimental data. Furthermore, the Anand model was used in the FEM analysis to evaluate solder joint thermal cycle reliability. The results show that solder Sn-3.5Ag has a better creep resistance than solder Sn-37Pb. The maximum stress is located at the upper right comer of the outmost solder joint from the symmetric center, and thermal fatigue life is predicted to be 3.796 × 10^4 cycles under the calculated conditions.
基金supported by the National S&T Major Project(Grant No.ZX06901)the National Natural Science Foundation of China(Grant No.11072131)
文摘Small reactors have become a new hotspot of international nuclear energy research.The nuclear heating reactor(NHR)technology developed by Tsinghua University is an important multipurpose small reactor solution with features such as high integration,modular design and full power natural circulation.A new small reactor based on the existing NHR-200 reactor was developed by the Institute of Nuclear and New Energy Technology of Tsinghua University.A full-scale natural circulation test loop with the same operating parameters as the actual reactor was built in order to experimentally validate the natural circulation ability of the reactor primary loop and heat-transfer ability of fuel assemblies and heat exchangers.Corresponding results are given in detail,including parameter validation of the reactor primary loop,flow rules of the natural circulation and heat-transfer coefficients of heaters and heat exchangers,which can be directly used in the actual reactor as a reference for optimization design.Finally,a characteristic parameter k is proposed to represent the natural circulation ability of a system.By using the new data arrangement method in the form of parameter k,comprehensive experimental results of the natural circulation can be represented by a simple integrated expression.The work in this paper is of importance in broadening application fields and pushing forward commercialization of the NHR type reactors.
