Measurement of Three-Dimensional Time-Averaged Void Fraction Distribution in Rod Bundle in Air-Water System by X-Ray CT Technique

We have been developing a void fraction distribution measurement technique using the three-dimensional (3D) time-averaged X-ray CT (computed tomography) system to understand two-phase flow behavior inside a fuel assembly for BWR (boiling water reactor) thermal hydraulic conditions of 7.2 MPa and 288 °C. Unlike CT images of a normal standstill object, we can obtain 3D CT images that are reconstructed from time-averaged X-ray projection data of the intermittent two-phase flow. We measured the 3D void fraction distribution in a vertical square (5 × 5) rod array that simulated a BWR fuel assembly in the air-water test. From the 3D time-averaged CT images, we confirmed that the void fraction at the center part of the channel box was higher than that near the channel box wall, and the local void fraction at the central region of a subchannel was higher than that at the gap region of the subchannel. A comparison of the volume-averaged void fractions evaluated by the developed X-ray CT system with those evaluated by a differential pressure transducer in a void fraction range from 0.05 to 0.40 showed satisfactory agreement within a difference of 0.03.

The Severe Accident Analysis of CPR1000 Based on MELCOR Code

The severe accident of CPR1000 caused by station blackout with the SG safety valve failure is simulated and analyzed using MELCOR code in this work. The CPR1000 power plant severe accident response process and the results with three different assumptions, which are no the seal leakage nor the auxiliary feed water, the seal leakage and auxiliary feed water exist, the seal leakage exist but no auxiliary feed water separately, are analyzed. According to the calculation results, without the seal leakage and auxiliary feed water, pressure vessel would fail at 9576 s. When auxiliary feed water was supplied, pressure vessel’s failure time would delay nearly 30000s. When the seal leakage exists, pressure vessel’s failure time would delay about 50 s. The results are meaningful and significant for comprehending the detailed process of severe accident for CPR1000 nuclear power plant, which is the basic standard for establishing the severe accident management guideline.

Thermal–Hydraulic Performance Analysis for AP1000 Passive Containment Cooling System

In order to improve the safety of new generation nuclear power plant, passive containment cooling system is innovatively used in AP1000 reactor design. However, since the system operation is based on natural circulation, physical process failure — natural circulation cannot establish or be maintained — becomes one of the important failure modes. Uncertainties in the physical parameters such as heat and cold source temperature and in the structure parameters have important effect on the system reliability. In this paper, thermal–hydraulic model is established for passive containment cooling system in AP1000 and the thermal–hydraulic performance is studied, the effect of factors such as air temperature and reactor power on the system reliability are analyzed.

An Experimental Study on Liquid Film Dynamics and Interfacial Wave of Air-Water Two-Phase Flow in a Horizontal Channel

This study is concerned with liquid film dynamics and stability of annular flow, which plays an important role in understanding film rupture and dryout in boiling heat transfer. The research work starts from designing and making a test facility which enables the visualization and measurement of liquid film dynamics. A confocal optical sensor is applied to track the evolution of film thickness. A horizontal rectangular channel made of glass is used as the test section. Deionized water and air are supplied into that channel in such a way that an initial stratified flow forms, with the liquid film on the bottom wall. The present study is focused on characterization of liquid film profile and dynamics in term of interfacial wave and shear force induced film rupture under adiabatic condition. Based on the experimental data and analysis, it is found that given a constant water flowrate, the average thickness of water film decreases with increasing air flowrate, while the interfacial wave of the two-phase flow is intensified. As the air flowrate reaches a critical value, a localized rupture of the water film occurs.

A New Model Towards Thermal Mixing and Stratification in Passive Containment

Thermal mixing and stratification often appears in passive containment cooling system (PCCS), which is an important part of passive safety system. So, it is important to accurately predict the temperature and density distributions both for design optimization and accident analysis. However, current major reactor system analysis codes only provide lumped parameter models which can only get very approximate results. The traditional 2-D or 3-D CFD methods require very long simulation time, and it’s not easy to get result. This paper adopts a new simulation code, which can be used to calculate heat transfer problems in large enclosures. The new code simulates the ambient fluid and jets with different models. For the ambient fluid, it uses a one-dimensional model, which is based on the thermal stratification and derived from three conservation equations. While for different jets, the new code contains several jet models to fully simulate the different break types in containment. Now, the new code can only simulate rectangular enclosures, not the cylinder enclosure. So it is meaningful for us to modify the code to simulate the actual containment, then it can be applied to solve the heat transfer problem in PCCS accurately.

Sobol′ Sensitivity Analysis Using a Neural Network Model of a LB-LOCA in the ZION Nuclear Power Plant With CATHARE-2 V2.5 Code

In nuclear safety, the Best-Estimate (BE) codes may be used in safety demonstration and licensing, provided that uncertainties are added to the relevant output parameters before comparing them with the acceptance criteria. The uncertainty of output parameters, which comes mainly from the lack of knowledge of the input parameters, is evaluated by estimating the 95% percentile with a high degree of confidence. IRSN, technical support of the French Safety Authority, developed a method of uncertainty propagation. This method has been tested with the BE code used is CATHARE-2 V2.5 in order to evaluate the Peak Cladding Temperature (PCT) of the fuel during a Large Break Loss Of Coolant Accident (LB-LOCA) event, starting from a large number of input parameters. A sensitivity analysis is needed in order to limit the number of input parameters and to quantify the influence of each one on the response variability of the numerical model. Generally, the Global Sensitivity Analysis (GSA) is done with linear correlation coefficients. This paper presents a new approach to perform a more accurate GSA to determine and to classify the main uncertain parameters: the Sobol′ methodology. The GSA requires simulating many sets of parameters to propagate uncertainties correctly, which makes of it a time-consuming approach. Therefore, it is natural to replace the complex computer code by an approximate mathematical model, called response surface or surrogate model. We have tested Artificial Neural Network (ANN) methodology for its construction and the Sobol′ methodology for the GSA. The paper presents a numerical application of the previously described methodology on the ZION reactor, a Westinghouse 4-loop PWR, which has been retained for the BEMUSE international problem [8]. The output is the first maximum PCT of the fuel which depends on 54 input parameters. This application outlined that the methodology could be applied to high-dimensional complex problems.

Transient Study on Sodium Heat Pipe in Passive Heat Removal System of Molten Salt Reactor

High temperature heat pipes are effective devices for heat transfer, which are characterized by remarkable advantages in conductivity, isothermality and passivity. It is of significance to apply heat pipes on new concept passive residual heat removal system (PRHRS) of molten salt reactor (MSR). In this paper, the transient performance of high temperature sodium heat pipe is simulated with numerical method in the case of MSR accident. The model of the heat pipe is composed of three conjugate heat transfers, i.e. the vapor space, wick structure and wall. Based on finite element method, the governing equations and boundary conditions are solved by using FORTRAN code to acquire the profiles of the temperature, velocity and pressure for the heat pipe transient operation. The results indicated that high temperature sodium heat pipe had a good operating characteristic and removed the residual heat of fuel salt rapidly under the accident of MSR.

Theoretical Research on Flow Instability in Parallel Channels Under Motion Conditions

In motion conditions, in addition to gravitational acceleration, a new acceleration was developed and it was added to the thermal hydraulics characteristics in flow channels. The additional acceleration leads to the different thermal hydraulic characteristics and will trigger the flow oscillation and even flow instability in parallel channels. In order to study the effect of the additional acceleration on the flow oscillation, the corresponding physical models are established in this work. Through the deduction of the mathematical model, the code for flow instability under motion conditions with Gear algorithm is developed. The flow oscillation curves, critical power, marginal stability boundary (MSB) are obtained. After comparison and analysis, it is found that some motion conditions lead to flow periodic oscillation. Different flow passage position results in different oscillation amplitudes. The marginal stability boundaries (MSB) under different motion conditions fit well, that is, the effect of motion conditions on MSB is small. Number of channels has little effect; however, channel arrangement influences the flow in every channel. These conclusions are of great significance in marine reactor design.

Verification of Accumulator Model in COSINE

The objective of COSINE (COre and System INtegrated Engine for design and analysis) project is to develop a software platform which is used for nuclear power plant design and safety analysis. The system code (SYST) is a part of COSINE code. In this paper, the mathematic model of accumulator is established. The model is based on following assumptions: nitrogen above liquid level in accumulator is represented in idea gas equation, water is modeled as an isothermal system. The model for liquid flow include inertia, wall friction, form loss, and gravity effects. Several cases were calculated under different conditions, and the results were compared with RELAP5. It shows that COSINE results agree well with RELAP5.

A Research on Eliminating the Deposited Water After Water-Ingress Accident in HTR

Water-ingress accident, caused by steam generator heating tube rupture of a high temperature gas-cooled reactor (HTGR) is an important accident to consider because it will introduce positive reactivity leading the nuclear power increase rapidly, as well as the chemical reaction of graphite fuel elements and reflector structure material with steam. Researches and simulations (Zuoyi Zhang et al. 1995; Zheng Yanhua et al. 2009) have been carried out for calculating the total amount of water ingress and to validate the safety and security of HTR. The water ingress amount, which is our mainly concerned, ranges from a few hundred kilograms to thousands of kilograms, because of the different reactors and different computing methods. The place, where the water deposits, is most likely to be the bottom of the steam generator.. Such liquid water removal, with the approach of providing a forced circulation in primary loop and accelerating the evaporation, is analyzed in this paper. Many experimental data have been got on water evaporation rate (Dalton et al.; Willis Carrie et al. 1914; Yoshida, Hyodo et al. 1970; Sweer et al. 1976; Pauken et al. 1995). All these formulas have a common form, ṁ = hc(Pw − Pa)/hw, which shows the relationship between evaporation rate and velocity over water face, water temperature and the relative partial pressure of the water vapor. This formula has been used widely in chemical industry and other fields and shows good agreement. FLUENT CFD code (ANSYS Fluent 14) is used for the calculation of the distribution of the flow and temperature fields. The evaporation rate is estimate combined thermal fluid data with empirical formula. As the working condition of empirical formula and that of actual reactor don’t match very well, sensitivity analysis is necessary in this report.