• Transactions of the Korean Society of Mechanical Engineers
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• v.13 no.5
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• pp.1032-1043
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• 1989

The heat transfer characteristics of the drag reducing polymer solutions are investigated experimentally in the thermal entrance region of circular tube flows. Fluids used in experiments are the aqueous solutions of high molecular polymer, polyacrylamide Separan AP-273 and the range of polymer concentrations is from 20 to 1000 wppm. Two stainless steel tubes with inside diameter 8.5mm(L/D=712) and 10.3mm(L/D=1160) are used for the heat transfer flow loops. The flow loop is set up to measure friction factors and heat transfer coefficients of test sections in two different modes; the recirculating flow system and once-through flow system. The test tubes are heated directly by electricity to apply the constant heat flux boundary conditions to the wall. Three different types of adaptors are used to observe the effects of the upstream flow conditions of the heat transfer test sections. The viscosity and characteristic relaxation time of the test fluids circulating in the flow system are measured by the capillary tube viscometer and falling ball viscometer at regular time intervals. The installed adaptors exhibit slight effect on the entrance heat transfer of Newtonian fluid. However, no noticeable effects are observed for the entrance heat transfer of the drag reducing fluids. The order of magnitude of the thermal entrance lengths of the drag reducing fluids which follow the minimum friction asymptote is much longer than that of Newtonian fluids in turbulent flows. A new dimensionless parameter, the viscoelastic Graetz number, is defined and all the experimental data are recasted in terms of the viscoelastic Graetz number. The local Nusselt number of the viscoelastic fluids is represented as a function of flow behavior index n and the viscoelastic Graetz number. As degradation continues the viscosity and the characteristic relaxation time of the testing fluids decrease. Weissenberg number defined by the relaxation time and D/V appears to be a proper dimensionless parameter in describing degradation effects on heat transfer of the viscoelastic fluids.

• Nuclear Engineering and Technology
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• v.53 no.8
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• pp.2534-2546
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• 2021

Nuclear emergency preparedness and response is an essential part to ensure the safety of nuclear power plant (NPP). Key support technologies of nuclear emergency decision-making usually consist of accident diagnosis, source term estimation, accident consequence assessment, and protective action recommendation. Source term estimation is almost the most difficult part among them. For example, bad communication, incomplete information, as well as complicated accident scenario make it hard to determine the reactor status and estimate the source term timely in the Fukushima accident. Subsequently, it leads to the hard decision on how to take appropriate emergency response actions. Hence, this paper aims to develop a method for rapid source term estimation to support nuclear emergency decision making in pressurized water reactor NPP. The method aims to make our knowledge on NPP provide better support nuclear emergency. Firstly, this paper studies how to build a Bayesian network model for the NPP based on professional knowledge and engineering knowledge. This paper presents a method transforming the PRA model (event trees and fault trees) into a corresponding Bayesian network model. To solve the problem that some physical phenomena which are modeled as pivotal events in level 2 PRA, cannot find sensors associated directly with their occurrence, a weighted assignment approach based on expert assessment is proposed in this paper. Secondly, the monitoring data of NPP are provided to the Bayesian network model, the real-time status of pivotal events and initiating events can be determined based on the junction tree algorithm. Thirdly, since PRA knowledge can link the accident sequences to the possible release categories, the proposed method is capable to find the most likely release category for the candidate accidents scenarios, namely the source term. The probabilities of possible accident sequences and the source term are calculated. Finally, the prototype software is checked against several sets of accident scenario data which are generated by the simulator of AP1000-NPP, including large loss of coolant accident, loss of main feedwater, main steam line break, and steam generator tube rupture. The results show that the proposed method for rapid source term estimation under nuclear emergency decision making is promising.