Band 83, Heft 1

A double-ended guillotine break at a direct vessel injection line of the ATALS facility has been evaluated with the MARS-KS thermal hydraulic system analysis code. The results are validated against the ATLAS experimental data provided under the First Domestic Standard Problem for Code Assessment (DSP-01) project, indicating that the calculation can generally explain the important phenomena observed in the experiment. The major difference is that the simulation predicts a 2 nd peak for the peak cladding temperature, which was not measured during the experiment. Further investigation for the core and downcomer liquid levels reveals that the 2 nd peak happens because of incorrect prediction of the downcomer wall temperature caused by the lack of information for heat loss through the downcomer wall in the experiment. This is confirmed by a calculation with the downcomer wall temperature measured at the experiment modeled as a boundary condition. The results with modified downcomer wall boundary conditions show a good prediction of the PCT behavior, indicating that additional heat loss measurements at the downcomer are required for a better understanding of the complex phenomena occurring in the ATLAS downcomer.

After the events of 9/11, the impact of fast flying commercial aircraft is considered as major hazard threatening the Nuclear Power Plant's (NPP) safety. The study of fuel spillage phenomenon and fireball formation is important to understand fire hazards due to burning of dispersed aviation fuel. The detailed analysis of fuel dispersion is very difficult to deliberate because both, large NPP structures and the large size of commercial aircrafts. Sandia National Laboratories, USA conducted impact tests using cylindrical projectiles filled with water to measure the associated parameters. Due to combustion properties and volatile nature of hydrocarbon fuels, the obtained parameters from impact studies using water are incomplete in fire analysis of flammable droplet clouds. A mathematical model is developed for prediction of droplet sizes and distribution associated with the impact of a liquid-containing projectile. The model can predict the transient behavior of droplet cloud. It is validated with experimental data available in literature. In the present study, the analysis has been performed using water and kerosene. The data obtained can be utilized as boundary and initial condition for CFD analysis. This information is useful for fire hazard analysis of aircraft impacts on NPP structures.

Many upcoming new generation reactors employ natural circulation for heat transfer in normal mode of operation. Natural circulation systems are simpler and safer than their forced circulation counterparts. However, these systems are prone to flow instability which are undesirable due to several reasons. In the present work, a rectangular glass loop, wherein cooler is just above the heater, has been considered for experimental and numerical investigation at atmospheric pressure. Heat addition from room conditions has been studied to understand the natural circulation loop dynamics, checking the possibility of occurrence of instability with the new orientation of the heater and the cooler. Experiments were performed at different power levels and coolant flow rates. CFD analyses were performed for all the cases investigated experimentally using the commercial CFD code ANSYS FLUENT 14.0. No instability was observed during the experiments and none during the simulations done for the duration of the experiments.

Numerical analysis on rewetting of nuclear fuel rod bundle by injecting coolant water in radial jet direction has been performed using Computational Fluid Dynamic (CFD). CFD-CFX results are compared with experimental data and an investigation is carried out for three ranges of numerical operating conditions (central water flow: 150 lpm, 225 lpm and 300 lpm; sub-cooled coolant: 288 K, 298 K and 308 K; initial wall temperature: 430 K, 500 K and 600 K). It was observed that the rewetting behavior shows an irregular pattern of rewetting progress. Rewetting velocity has got no significance under any operating conditions. The effective numerical response is observed for rewetting temperature and wetting delay. Results show an increase in rewetting temperature, and reduction in wetting delay under higher flow rate, low sub-cooling and lowering the initial wall temperature.

The theoretical prediction of the cross-sections in the determination of the structure materials used in the fission and fusion technology is very important in terms of reactor safety. Also, theoretical cross-sections are necessary to evaluate the performance of structure materials, especially, in the lack or absence of experimental data. Nickel base alloys are one of the wide ranges of structural materials used in fission and fusion reactors. Therefore, the excitation functions of the (n, 2n) reactions on the structure material nickel with neutrons producing via fusion reactions are investigated in this study. Thus, the excitation functions for 58,60 – 62,64 Ni(n, 2n) reactions are calculated using some pre-equilibrium models which are the hybrid, the geometry depended hybrid and the two-component excitation models. The calculations are performed in the ALICE/ASH and the Talys 1.8 codes. Also, the equilibrium and pre-equilibrium effects on these reactions are investigated in this study. The obtained cross-sections are discussed and compared with the experimental data and the theoretical predicted data.

For the simulation of centrifugal pump in flow network system, in general three approaches can be used, the fitting model, the numerical method and the internal characteristics model. The fitting model is simple and rapid thus widely used. The numerical method can provide more detailed information in comparison with the fitting model, but increases implementation complexity and computational cost. In real-time simulations of flow networks, to simulate the condition out of the rated condition, especially for the volume flow rate, which the accuracy of fitting model is incredible, a new method for simulating centrifugal pumps was proposed in this research. The method based on the theory head and hydraulic loss in centrifugal pumps, and cavitation is also to be considered. The simulation results are verified with experimental benchmark data from an actual pump. The comparison confirms that the proposed method could fit the flow-head curves well, and the responses of main parameters in dynamic-state operations are consistent with theoretical analyses.

The method for the evaluation of the control room habitability is presented in this paper with focus on Maanshan PWR nuclear power plant (NPP) using the codes RADTRAD, HABIT, and ALOHA. Therefore, this paper is divided into two parts: The first part is the evaluation of the cumulative dose at the control room, the exclusion area boundary (EAB) and the low population zone (LPZ) in case of an design basis loss of coolant accident (DBA/LOCA). For this first part, the Maanshan NPP models of the code RADTRAD/SNAP were used for the analysis. The second part is the evaluation of the control room habitability under the assumption of CO 2 storage burst. For this part, the HABIT and ALOHA codes were used. As result it was seen that the RADTRAD calculation results are below the failure criteria of standard review plan (SRP) and 10 CFR 100.11. The HABIT and ALOHA results are below the R.G. 1.78 failure criteria. These results indicate that Maanshan NPP’ habitability can be maintained under the above conditions.

Safety analysis is a major part during licensing a nuclear power plant. This paper presents coupled neutronic thermal-hydraulic calculations for a MOX/UOX mixed core loading of the Westinghouse AP1000 pressurized water reactor concept. Originally, the initial core loading of the reactor has three types of UO 2 fuel assemblies with U 235 enrichments are: 2.35 w/o, 3.40 w/o and 4.45 w/o. In this paper, it is assumed that one-third of fuel assemblies is replaced by MOX assemblies. The steady state conditions are calculated with the code QUARK. QUARK is a 3D neutronic kinetics code coupled to core thermal hydraulics code. The required cross-section data for QUARK were generated using WIMSD5. The results of the steady state analysis show that both core loading schemes can be used in AP1000 as the safety criteria mentioned in design control document are met.