Volume 84, Issue 6

The Secondary Circuit Passive Residual Heat Removal System (PRHRS) is designed for advanced nuclear power system. A research on thermal hydraulic characteristics of two-phase natural circulation process for secondary circuit PRHRS has been carried out using the RELAP5 code. For this, the natural circulation steam generator (SG) and the PRHRS are modeled. Five thermal boundary conditions are set to verify the function of PRHRS. The phenomenon of flow instability in the two-phase natural circulation process is of great significance for system operation, as well as the performance of SG and condenser. Therefore, the focus is on the natural circulation flowrate oscillation of steam and condensed water. The mixed water level difference between descending pipe of PRHRS and secondary side of natural circulation SG is found to be the meaningful parameter to characterize the flow instability of natural circulation. The comparative analysis is also performed by considering the effects of height between the SG and condenser, as well as system resistance, which can provide support for PRHRS design and optimization.

This paper provides analysis of the operating experience feedback from two events of incidental fluctuations of the power measurement channels at Egypt Second Research Reactor (ETRR-2). In the first event, the fluctuation was due to improper functioning of the secondary cooling system, while in the second one the fluctuation was associated with a leakage of a gadolinium nitrate solution into chambers surrounding the reactor core (which are part of a diverse second shutdown system of the reactor). The paper describes these events, including observation of the reactor operation performance and operators’ actions, and provides the root causes analysis of these events along with the corrective actions. Lessons learned from these events, which are generically applicable to other research reactors, are also identified and presented. The analysis showed the vital role of continuing training of reactor operators, consideration of human factors in all activities during the reactor lifetime, and quality checks and field verifications during operation and post-maintenance works to the safety and availability of research reactors. The analysis also showed that underperformance of non-safety components could have an effect on those important to safety and could negatively impact the availability and the operational performance of the reactor.

This work is a part of the International Atomic Energy Agency (IAEA) Coordinated Research Project (CRP) titled: \Benchmarks of Computational Tools against Experimental Data on Fuel Burnup and Material Activation for Utilization, Operation and Safety Analysis of Research Reactors". The OPAL research reactor has been modelled using MCNP codes as a benchmark study. Neutron flux profiles, critical configurations, control rod worths, fuel burnup and kinetic parameters have been calculated. Comparisons between the calculated results and the experimental data given in the OPAL benchmark have been performed.

In the probabilistic safety analysis (PSA) of nuclear power plants (NPP), fault tree analysis is often taken as a useful way for reliability evaluation. Logic loops due to mutual dependencies among systems may emerge in fault trees. Logic loops or circular logics can make it complex to get the minimum cut sets. This paper presents a method based on directed graphs and Binary Decision Diagram (BDD) to break logic loops and get the minimum cut sets consistent to previous attempting methods. The paper illustrates this method by giving one linear actual case in nuclear field and non-linear case presented by Vaurio. Utilizing directed graphs and Binary Decision Diagram is effortless for reliability analysts and concerned people in a nuclear field to have an intuitive comprehension for correlations among systems without logic loops. What’s more, directed graphs and BDD can express the minimum cut sets of systems that can be understood intuitively. In addition, this present method is able to solve non-linear logic loops that is hard to operate in fault tree analysis. This proposed non-iterative method is also helpful to design good reliability engineering systems.

The nuclear industry has seen an increased use of Computational Fluid Dynamics (CFD) technology as a high-fidelity tool for design-basis and beyond design-basis accident simulations. For fire hazards, detailed experimental investigation in a realistic environment is often impractical. Therefore, the use of an advanced 3D CFD modeling approach to predict smoke and fire propagation has risen, especially for nuclear power plant fire modeling applications. CFD simulations can include major contributing and mitigating mechanisms regarding the fire propagation and its consequences, in fine detail and under realistic geometry conditions. In addition, accurate simulations of unsteady 3D fire and smoke spread, in conjunction with evacuation analyses, can be used to establish effective emergency evacuation strategies. In the present study, the fire and smoke propagation in a main control room was modelled using Reynolds-Averaged Navier-Stokes Equation based twoequation turbulence models and a Large Eddy Simulation model implemented in a commercial CFD code, Siemens STAR-CCM+. The predictions from STAR-CCM+ were compared with those obtained in a previous study using the NIST Fire Dynamic Simulator (FDS) code. Although higher temperature was predicted at the fire source by STAR-CCM+, the current study predicted trends similar to the FDS results with some minor differences at the operator location. A follow-up study is currently underway to investigate the factors affecting the differences, including further assessment of codes against available benchmark experiments.

IIST (Institute of Nuclear Energy Research Integral System Test) facility is a RHRP (Reduced-Height and Reduced-Pressure) facility for the three-loop PWR (Pressurized Water Reactor) NPP (Nuclear Power Plant) Maanshan. This test facility was established in 1992 for safety studies of the Maanshan NPP. TRACE is an advanced thermal hydraulic code which is still developed and updated by U.S. NRC. So, validation of specific models against experiments is still necessary. In this paper, cooldown experiments in IIST facility are used to validate the TRACE models. Additionally, the results of TRACE calculations are compared to results of RELAP5 calculations. This comparison shows that the calculations of TRACE are in good agreement to the data of IIST experiments and to the calculational results of RELAP5. So, TRACE can be used to simulate cooldown transients.