State Nuclear Power Technology Corporation Ltd., Shanghai Nuclear Engineering Research and Design Institute | Date: 2014-01-08
A flow distribution device (3) for a reactor, a lower internals (100) of a reactor and a reactor are provided. The lower internals (100) includes: a lower core support plate (2) defining a coolant hole therethrough; a flow distribution device (3) mounted on the lower core support plate (2) and including a distribution annular plate (8) and a distribution bottom plate (9); a vortex suppression plate (7) disposed below the distribution bottom plate (9); a support column (4) defining an upper end connected with the lower core support plate (2) and a lower end passing through the distribution bottom plate (9) to connect with the vortex suppression plate (7); an energy absorption device (5) defining an upper end connected with the vortex suppression plate (7); and an anti-break bottom plate (6) disposed on the lower end of the energy absorption device (5).
Qi Z.,Shanghai Nuclear Engineering Research and Design Institute
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2015
In the large advanced passive PWR nuclear power plant, the long term core cooling (LTCC) following loss of coolant accidents (LOCA) is provided by passive safety system. The purpose of this paper is to study whether the factors that could affect LTCC challenge the safety margin. According to the preliminary PIRT (Phenomena Identification and Ranking Table) of the large advanced passive PWR nuclear power plant, the following factors have been selected: the timing of the recirculation initiation ahead, the swing check valves in the safety injection lines partially opened, core inlet blockage due to debris, the resistances of automatic depressurization system ADS valves, the containment water flooding inventory, the containment pressure and the recirculation temperature. The cases of LTCC following a double-ended direct vessel injection (DEDVI) line break are analyzed. The effects of the increase of ADS valves resistances, the decrease of containment pressure and the increase of recirculation temperature are calculated respectively. It's found that the containment pressure and the containment sump temperature may also play important roles during LTCC. A limiting case combined with all of these factors is performed finally, which the results demonstrates that the passive systems of the large advanced passive PWR nuclear power plant will provide adequate core cooling performance during LTCC. Copyright © 2015 by JSME.
Shi G.,Shanghai Nuclear Engineering Research and Design Institute
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2015
The "Nuclear Safety Planning" has been published in Oct. 2012 in China, which stipulates the safety goals for the NPPs which will be built in the future. As for the nuclear power plants (NPPs) which will be built in China's Thirteenth Five-Year (2016-2020) and later, the high level safety goal is described as that "the possibility of the large radioactive release should be practically eliminated by design". A thorough investigation has been performed at SNERDI to explore the technical insight of this high level safety goal by using Multinational Design Evaluation Project (MDEP) hierarchical safety goal approach. The definition of large release is proposed accordingly, DID requirements and probabilistic requirements are derived from this high level safety goal. Copyright © 2015 by JSME.
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.
Pan J.,Shanghai Nuclear Engineering Research and Design Institute
IEEE International Professional Communication Conference | Year: 2015
This paper introduces a concept of process-oriented wiki system; verifies its effectiveness by means of applying it to daily work, evaluating human performance and working quality; as well as, comparing it to a conventional training method. The result shows the time required to master a skill will be greatly cut down and the working quality can be steadily increasing. © 2014 IEEE.
Huang G.,Shanghai Nuclear Engineering Research and Design Institute |
Fang L.,Shanghai Nuclear Engineering Research and Design Institute
Annals of Nuclear Energy | Year: 2013
After Fukushima accident, hydrogen control system reliability evaluation is implemented for most of nuclear power plants around the world. For early CANDU6 design, there is no hydrogen control system used for severe accidents. For CANDU6 hydrogen control in severe accidents, accept criteria is different with PWR. Consideration of deuterium is discussed. CL4, LLOCA and FBS are selected as typical severe accident sequences. Hydrogen concentration and hydrogen risk without mitigation are investigated. With the analysis base of hydrogen behavior without mitigation, 18 passive autocatalytic recombiner (PAR) is located into containment compartments. According to analysis results, hydrogen concentration is controlled below 10% during accidents, hydrogen risk is eliminated. © 2013 Elsevier Ltd. All rights reserved.
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.
Shanghai Nuclear Engineering Research and Design Institute | Date: 2012-06-20
A heat pipe based passive residual heat removal system for a spent fuel pool has a plurality of partitions (6) arranged around the inside of a spent fuel pool (3), wherein the heights of the partitions (6) are all lower than the height of the spent fuel pool (3); a plurality of evaporation-end heat pipes (4) are arranged between the outside of the partitions (6) and an inner wall of the spent fuel pool (3), and these evaporation-end heat pipes (4) are divided into several groups. Top outlets of each group of evaporation-end heat pipes (4) are extended beyond the spent fuel pool (3) and are connected to an inlet of an ascending pipe (10). An outlet of the ascending pipe (10) is connected to top inlets of a group of condensation-end heat pipes (7) comprising a plurality of condensation-end heat pipes. Bottom outlets of the group of condensation-end heat pipes (7) are connected to an inlet of a descending pipe (5). An outlet of the descending pipe (5) is extended downwardly into the spent fuel pool (3) and is connected to bottom inlets of a group of evaporation-end heat pipes (4). The present invention employs the heat pipes for cooling the spent fuel pool, so that a heat exchange by phase change or working medium in the heat pipe leads to heat exchange of low temperature difference and high efficiency, relying on density difference for natural circulation and driving, thus fundamentally eliminating reliance on a power source and personnel, and thereby implementing long-term passive heat exchange for cooling of the spent fuel pool efficiently.
Shanghai Nuclear Engineering Research and Design Institute | Date: 2013-10-25
A gray control rod, a neutron absorber thereof and a gray control rod assembly. The neutron absorber comprises at least one first component and at least one second component, the reactivity worth of the first component increases as the service time of the neutron absorber increases, the reactivity worth of the second component decreases as the service time of the neutron absorber increases; the reactivity worth of the neutron absorber varying no more than 15% within the service time of the neutron absorber. By using the first component and the second component to form the neutron absorber, and adjusting the proportion of the respective components in the neutron absorber, the neutron absorber having a substantially planar reactivity worth loss characteristic can be obtained. The gray control rod and the assembly having required reactivity controlling ability can be obtained by increasing or decreasing the material dosage of the neutron absorber.
Shanghai Nuclear Engineering Research and Design Institute | Date: 2013-12-31
A passive cooling system for a reactor core of a large-scale pressurized water reactor nuclear power plant includes a shield building having an outer wall and a through air inlet arranged on an upper part of the outer wall, a water tank arranged at an upper part of the shield building, a cooling water distribution plate arranged above a top of a containment within the shield building, a spray pipe arranged at an inside of the top of the shield building and having a water inlet end and a water outlet end, wherein the water inlet end is connected to a bottom of the water tank and the water outlet end is extended to be above the cooling water distribution plate, and an air deflector arranged between the shield building and the containment and having an upper end connected to an inside of the top of the shield building.