State Nuclear Power Technology nter

Beijing, China

State Nuclear Power Technology nter

Beijing, China

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Chang H.,State Nuclear Power Technology nter | Chang H.,Tsinghua University | Hu T.,State Nuclear Power Technology nter | Lu W.,State Nuclear Power Technology nter | And 2 more authors.
Experimental Thermal and Fluid Science | Year: 2017

In vessel retention (IVR) by passive external reactor vessel cooling (ERVC) under severe accidents is a feasible approach to retain radioactive core melt within the reactor vessel. The effectiveness of IVR by external reactor vessel cooling strongly depends on the critical heat flux (CHF). As long as the local heat flux does not exceed the local CHF in the vessel, the lower head can be cooled sufficiently to prevent its failure. In this study, the FIRM subcooled flow boiling facility conducted by State Nuclear Power Technology Research & Development Center (SNPTRD) was built to simulate the IVR-ERVC condition for Chinese Advanced Passive 1400 MWe PWR, as well as to verify and validate the safety margin of CHF during severe accidents. To investigate the effects of the heater surface material, coolant additives and other thermal hydraulic parameters such as flow rate and subcooling on CHF, flow boiling CHF experiments using a full scale 2-D curved test section under atmospheric pressure with SA508 Gr3 carbon steel heater as well as coolant additives of trisodium phosphate (TSP, Na3PO4) and boric acid (BA, H3BO5) with different concentrations were performed. The results showed that SA508 material displayed quite different CHF behavior in comparison with other materials such as stainless steel, copper and aluminum. It showed that higher CHF value and the test heater surfaces were changed significantly. The test heater surface change was due to the corrosion of SA508, and the rate of corrosion increased with boiling time. CHF values showed a little reduction with increasing the concentrations of BA. In case of TSP, CHF values were enhanced with lower concentrations of 500 ppm and 1000 ppm due to the increased wettability of coolant, while it was reduced with 3500 ppm probably due to the instability of two phase flow and the preventing effect on SA508 corrosion. CHF values for the mixed solutions of BA and TSP showed the similarity with the case of TSP, which indicated that TSP played more important effect on CHF behavior. © 2017


Chen L.,State Nuclear Power Technology nter | Fang F.-F.,State Nuclear Power Technology nter | Deng C.-C.,Tsinghua University | Cui C.-X.,State Nuclear Power Technology nter
Yuanzineng Kexue Jishu/Atomic Energy Science and Technology | Year: 2015

The uncertainty evaluation is a key step of best estimate safety analysis for nuclear power plant. The origins of the uncertainty and the uncertainty evaluation methods were described. The methods could be categorized into statistical type and deterministic type, and the general procedure of statistical evaluation method was summarized. Various uncertainty evaluation methods were analyzed and compared from the aspect of computational cost and accuracy. The analysis result shows that the method coupling the nonparametric sampling with a sophisticated high-fidelity thermal-hydraulic model may be the best choice for the uncertainty evaluation in best estimate analysis at present. Under the condition that the “95/95criteria” is satisfied, such a method is easy to implement and the cost is relatively low. ©, 2015, Atomic Energy Press. All right reserved.


Chen L.,State Nuclear Power Technology nter | Hu X.,State Nuclear Power Technology nter | Deng C.-C.,Huazhong University of Science and Technology | Huang T.,State Nuclear Power Technology nter
Yuanzineng Kexue Jishu/Atomic Energy Science and Technology | Year: 2016

The application of best estimate plus the uncertainty (BEPU) analysis method becomes trend in the nuclear power plant accident analysis and safety review. The procedure of the statistical BEPU method based on input parameters' propagation was summarized and the key steps were analyzed. The evaluation process can be divided into five major steps: Defining the target parameter(s), identifying important input parameters and their probability distribution, sampling, model analysis and target parameter analysis. It's believed that phenomenon identification and ranking table (PIRT) is a suitable method to identify the important input parameters. The distribution of input parameters is usually obtained from the experimental data or expert judgment. The parametric or nonparametric sampling method can be used to determine the required sampling number, and the later greatly decreases sampling number. Its calculation model should be verified sufficiently to demonstrate its applicability. The statistical results of objective parameter provide uncertainty range and the sensitivity of the input parameters. © 2016, Editorial Board of Atomic Energy Science and Technology. All right reserved.


Wang X.-M.,Tsinghua University | Chang H.-J.,Tsinghua University | Chang H.-J.,State Nuclear Power Technology nter | Yang L.,State Nuclear Power Technology nter | Zhao R.-C.,State Nuclear Power Technology nter
Yuanzineng Kexue Jishu/Atomic Energy Science and Technology | Year: 2016

Falling film evaporation plays a key role in the heat removal process of the containment of the CAP1400. In order to estimate the heat removal capability of the containment, it is important to calculate the evaporation rate accurately. In the paper, two different evaporation models were built to simulate film evaporation in a vertical channel based on ANSYS FLUENT code. The prediction results by two evaporation models and experimental data were compared. The results show that both models can predict the evaporation heat transfer coefficient accurately. The results calculated by the model 1 are more reliable with finer mesh. The model 2 can be used with coarser mesh near the wall. However, the model depends greatly on the calculation of the convective heat transfer coefficient. © 2016, Editorial Board of Atomic Energy Science and Technology. All right reserved.


Lu Y.,State Nuclear Power Technology nter | Wang Y.,State Nuclear Power Technology nter | Liu L.,State Nuclear Power Technology nter
Hedongli Gongcheng/Nuclear Power Engineering | Year: 2016

Quasi-steady time characterized the response of the water film to the steady state in the passive containment cooling system. It is a key factor in the heat transfer and design of the nuclear safety system. There are two processes during the falling film covering the containment surface. The first process is water flows down, resulting in a wet area; the second is water laterally wet the dry area due to its semi-stability, and the contact angle becomes smaller and slowly increases the coverage area. Coupling of the two processes makes it difficult to determine the quasi-steady time. On the basis of the results of collection tank level, flow and pictures of water film coverage in CAP1400 water distribution experiment, present study proposed two methods to determine the stability of the water film, which are flow balance method and coverage stable method. It solved the calculation problem of quasi-steady time and provided a new way for determining the stability of the water film, and furthermore, relations of relative quasi-steady time with Reynolds number based on different distribution water structures are obtained. © 2016, Editorial Board of Journal of Nuclear Power Engineering. All right reserved.


Cui C.-X.,State Nuclear Power Technology nter | Chang H.-J.,Tsinghua University | Huang T.,State Nuclear Power Technology nter | Chen L.,State Nuclear Power Technology nter
Yuanzineng Kexue Jishu/Atomic Energy Science and Technology | Year: 2016

Due to the requirements of safety level of nuclear power plant are gradually improving, more and more passive systems are used in the advanced type reactor. However, the reliability evaluation work for these passive systems is still in the primary stage. In this paper, based on the reliability evaluation process of passive system, the physical process of passive system was simulated by RELAP5 thermal-hydraulics program and the reliability of AP1000 external reactor vessel cooling (ERVC) system was evaluated. After the large number of calculations, the cumulative density distribution curve of the temperature in lower head of the reactor pressure vessel and other parameters were obtained. Combined with specific success criteria, the reliability of AP1000 ERVC system was got. This reliability evaluation result of ERVC system can be used in the PSA model to better guide the design of nuclear power plants and improve the safety of nuclear power plant. © 2016, Editorial Board of Atomic Energy Science and Technology. All right reserved.


Hu X.,State Nuclear Power Technology nter | Huang T.,State Nuclear Power Technology nter | Pei J.,State Nuclear Power Technology nter | Chen L.,State Nuclear Power Technology nter
Yuanzineng Kexue Jishu/Atomic Energy Science and Technology | Year: 2015

The MELCOR1.8.6 code was applied to a severe accident model of a 1 000 MWe PWR which includes primary system, secondary system, passive core cooling system and containment system. For the transient case, a small break LOCA with 2 inch (5.08 cm) break at the cold leg concurrent with failure of gravity injection was selected. After the core was damaged due to the failure of gravity injection, it was assumed that the coolant was injected into the pressure vessel, and then the water reflooding effectiveness was evaluated and analyzed. In this calculation, the coolant injection into reactor core with the small (10 kg/s), medium (50 kg/s) and large (200 kg/s) mass flow rates respectively at 3 different time stages of the severe accident was simulated. The effectiveness of water reflooding was assessed through hydrogen production, radioactive materials released from core, and core temperature. The results show that the mass flow rate above 10 kg/s is believed to be efficient for cooling a 1 000 MWe reactor at the beginning of core damage. However, with the accident developing to core relocation, a large mass flow rate of 200 kg/s is considered to be applicable for core cooling. As a result, the mass flow rate below this value should be carefully considered when injecting water into the core. © 2015, Editorial Board of Atomic Energy Science and Technology. All right reserved.


Deng C.-C.,Huazhong University of Science and Technology | Chang H.-J.,State Nuclear Power Technology nter | Chen L.,State Nuclear Power Technology nter
Yuanzineng Kexue Jishu/Atomic Energy Science and Technology | Year: 2016

The RELAP5 best-estimate code was used to perform small break loss of coolant accident (SBLOCA) for the ACME (advanced core-cooling mechanism experiment) facility built in China. Thereafter, uncertainty quantification evaluation was performed, including the selection of input uncertain parameters, the application of Wilks nonparametric statistics and the calculation of uncertainty propagation based on SNAP interface. Furthermore, the results of uncertainty and sensitivity analysis were discussed. The 95/95 uncertainty bands of key output parameters were obtained, and the lower band of minimum core level is maintained still above the top of the active core section, thus the core uncover does not occur with 95% confidence. Important uncertainty contributors for the minimum core level are identified through sensitivity analysis. © 2016, Editorial Board of Atomic Energy Science and Technology. All right reserved.


Zhang J.-L.,State Nuclear Power Technology nter | Li P.-F.,State Nuclear Power Technology nter | Chen Y.,State Nuclear Power Technology nter
Yuanzineng Kexue Jishu/Atomic Energy Science and Technology | Year: 2015

The applicability of cold-crucible induction heating technology to the experimental study on the heat transfer characteristics of a molten oxide pool was analyzed in this paper. Then improvements on traditional cold-crucible structure were proposed. The intensity and distribution of internal Joule heat and Lorentz force of the improved cold-crucible were analyzed with numerical simulation method. The analysis results show that the distribution of internal Joule heat is more homogeneous with lower electric power frequency, and the Joule heat amount accounts for a smaller share of the total power, which will be benefit to the heat flux measurement of the side wall of the pool. Besides, when compared with the driving force of natural convection in the pool, the Lorentz force can be ignored, which means the Lorentz force has little impact on the heat transfer and the flow behavior of the molten pool. ©, 2015, Atomic Energy Press. All right reserved.


Wang Y.,Tsinghua University | Yang Y.,State Nuclear Power Technology nter | Zhang Y.,Tsinghua University | Zhou Z.,Tsinghua University
Hedongli Gongcheng/Nuclear Power Engineering | Year: 2016

PCCSAP-3D is a dedicated analysis code developed for the Westinghouse AP1000 PCCS. In this paper, the models on the AP1000 reactor system are described. The PCCS performances during the postulated design basis accidents (LOCA and MSLB) simulated and analyzed from PCCSAP-3D are compared with that from the WGOTHIC developed by Westinghouse. The results show good agreements, which indicate that the contain-ment pressure during the postulated design basis accident can be limited effectively under the design value by the PCCS performance and also the capacity of the PCCSAP-3D for the analysis and the evaluation on the PCCS performance is validated preliminarily. © 2016, Yuan Zi Neng Chuban She. All right reserved.

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