He X.,Shanghai Nuclear Engineering Research and Design Institute
International Journal of Advanced Manufacturing Technology | Year: 2011
In heavy forging, a manipulator is indispensable to help the precision of the forging process. The manipulator holds forging workpiece during the forming process, but whether manipulator compliant movements have effects on forging quality of workpiece is still unknown. A method to evaluate the effects of manipulator compliant movements on the forging quality qualitatively is presented in this paper. Forging deformation FEM simulation and manipulator dynamics analysis are coupled. A forging quality evaluation function considering strain distribution uniformity is established. The case study of an eight-square elongation process indicates that manipulator active vertical compliant movement and positive horizontal compliant movement can improve forging quality to some extent in certain conditions. The results show that manipulator compliant movements do have effects on forging quality and can be utilized potentially in designing a forging process. © 2011 Springer-Verlag London Limited.
Ke X.,Shanghai Nuclear Engineering Research and Design Institute |
Ke X.,Shanghai JiaoTong University
Yuanzineng Kexue Jishu/Atomic Energy Science and Technology | Year: 2014
The steam generator overfilling of CAP1000 nuclear power plant in steam generator tube rupture (SGTR) accident was analyzed. The power range for this analysis included typical part-power, zero power and hot standby. The sensitivity analyses were compared at each power level. And the conservative case at every power level was selected to contrast with the result of full power. Based on the transient characteristics, the accident processes were analyzed, and then the limiting case and critical parameters were evaluated. It turns out that the steam generator of CAP1000 will not be overfilled in SGTR accident which is assumed to occur in entire power range, and the most limiting case appears in full power condition.
Zhan W.,Shanghai Nuclear Engineering Research and Design Institute
He Jishu/Nuclear Techniques | Year: 2010
The Steam Generator Tube Rupture (SGTR) accident has been analyzed with RELAP5/MOD3 thermal hydraulic system analysis code under the conditions that the safety valve of affected SG is stuck open. The AFW is unavailable, and only one HPSI pump is available. It can be mitigated successfully if the operators take actions to cooldown and depressurize RCS within 150 min after the safety injection signal. The results show that the active core is covered by coolant before the break flow is terminated with the balance between primary and secondary side of affected SG at atmospheric pressure.
Wang X.-J.,Shanghai Nuclear Engineering Research and Design Institute
Yuanzineng Kexue Jishu/Atomic Energy Science and Technology | Year: 2012
The parametric uncertainties of DNBR and exit quality were calculated using sampling statistical method based on Wilks formula and VIPRE-W code. Then the DNBR design limit and exit quality limit were got by combining with the uncertainties of models and DNB correlation. This method can gain the more DNBR margin than RTDP methodology which is developed by Westinghouse by comparison of these two methods.
Zhang L.,Shanghai JiaoTong University |
Bao Y.,Shanghai Nuclear Engineering Research and Design Institute |
Tang R.,National Key Laboratory For Nuclear Fuel And Materials
Nuclear Engineering and Design | Year: 2012
Supercritical water cooled reactor (SCWR) is a promising Gen IV high performance reactor which can be developed for future large capacity electric power plants. However, the material selection for fuel cladding still remains one of the key issues. For a typical prototype SCWR design with outlet coolant temperature of 510°C and pressure of 25 MPa, the hot spot temperature on fuel cladding exceeds 600°C at normal operation conditions, and will be much higher during transients. Materials for fuel cladding should have good mechanical properties to meet the harsh working conditions in order to keep the integrity of fuel rod under normal and abnormal operational conditions, while corrosion in supercritical water and neutron irradiation damage will not lead to significant loss of strength and ductility, or lead to unacceptable deformation during service life. Materials for ultra-supercritical fossil fire plants, fast breeder reactor and jet air engines etc., are proposed as the candidate materials for SCWR fuel cladding. This paper reports the results based on corrosion screening tests of candidate materials exposed in supercritical water up to 650°C. Results show that their corrosion rates increase significantly with the increase of temperature, and the protective oxide films of most candidate materials turn to be unstable above temperature of 600°C. According to the present knowledge available, austenitic stainless steels with high Cr concentration show better performance and are more potential to be the references for developing the SCWR fuel cladding material. © 2012 Elsevier B.V.