Study of Heat Mass and Mass Transfers in a Spray for Containment Application: Analysis of the Influence of the Spray Mass Flow Rate

During the course of a hypothetical accident in a nuclear power plant, spray might be activated in order to reduce static pressure in the containment. To have a better understanding of the heat and mass transfers between a spray and the surrounding confined gas, the IRSN has developed the TOSQAN experiment. This article is devoted to analyse the influence of the injected spray mass flow rate on the thermal-hydraulic of spray tests. In order to perform this analysis, two other tests are performed with exactly the same boundary and initial conditions, except the spray mass flow rate that ranges from 10 to 50 g.s−1. First, the scenario of these three tests and the associated results used for this analysis are presented. Then, we focus our analysis on the inter-comparison of the thermal-hydraulic behaviour induced by spray mass flow rates variations. This inter-comparison is divided into two parts: a global and a local one.

Investigation of Operation of VVER Steam Generator in Condensation Mode at the Large-Scale Test Rig

In new Russian NPP with VVER-1200 reactor (V-392M reactor plant) in the event of an accident being due to the rupture of the reactor primary circuit and accompanied by the loss of a.c. sources, provision is made for the use of passive safety systems for necessary core cooling. Among these is passive heat removal system (PHRS). In the case of leakage in the primary circuit this system assures the transition of steam generators (SG) to operation in the mode of condensation of the primary circuit steam coming to SG piping from the reactor. As a result, the condensate from steam generators arrives at the core providing its additional cooling. To experimental investigation of the condensation mode of operation of VVER steam generator, a large scale HA2M-SG test rig was constructed. The test rig incorporates: tank-accumulator, equipped by steam supply system; SG model with volumetric-power scale is 1:46; PHRS heat exchanger simulator, cooling by process water. The rig main equipment connected by pipelines and equipped by valves. The elevations of the main equipment correspond to those of reactor project. The rig maximum operating parameters: steam pressure – 1.6 MPa, temperature – 200 Celsius degrees. Experiments at the HA2M-SG test rig have been performed to investigate condensation mode of operation of SG model at the pressure 0.4 MPa, correspond to VVER reactor pressure at the last stage of the beyond basis accident. The report presents the test procedure and the basic obtained test results.

Near Term Deployment, Long Term Impact: Uranium Price Over the Lifetime of New Capacity

While reasonable short-term resource price projections can be obtained by taking a bottom-up approach — constructing a supply curve based upon current production capacities and costs — this approach breaks down as the time horizon of the analysis lengthens. One approach to long-term price forecasting is to calibrate a simple model of a commodity market against past data. To that end, an analogy was drawn between the behavior of the uranium market and that of some three dozen materials for which the United States Geologic Survey (USGS) maintains data. This work adds to previously published results showing that the USGS-reported prices of minerals similar to uranium have consistently declined over the past century. In this paper, the extent to which uranium geology and extraction technologies are indeed analogous to other minerals is quantitatively addressed. A study of crustal abundances, ore grades being economically mined, concentration factors, market share of extraction techniques, years of proven reserve and other factors indicates that uranium is not at all exceptional with respect to the average of the USGS minerals. This suggests that, on the supply side, the analogy between the USGS minerals and uranium may indeed offer valuable insights into medium and long term uranium price behavior.

Performance of Core Exit Thermocouple for PWR Accident Management Action in Vessel Top Break LOCA Simulation Experiment at OECD/NEA ROSA Project

Presented are experiment results of the Large Scale Test Facility (LSTF) conducted at the Japan Atomic Energy Agency (JAEA) with a focus on core exit thermocouple (CET) performance to detect core overheat during a vessel top break loss-of-coolant accident (LOCA) simulation experiment. The CET temperatures are used to start accident management (AM) action to quickly depressurize steam generator (SG) secondary sides in case of core temperature excursion. Test 6-1 is the first test of the OECD/NEA ROSA Project started in 2005, simulating withdraw of a control rod drive mechanism penetration nozzle at the vessel top head. The break size is equivalent to 1.9% cold leg break. The AM action was initiated when CET temperature rose up to 623K. There was no reflux water fallback onto the CETs during the core heat-up period. The core overheat, however, was detected with a time delay of about 230s. In addition, a large temperature discrepancy was observed between the CETs and the hottest core region. This paper clarifies the reasons of time delay and temperature discrepancy between the CETs and heated core during boil-off including three-dimensional steam flows in the core and core exit. The paper discusses applicability of the LSTF CET performance to pressurized water reactor (PWR) conditions and a possibility of alternative indicators for earlier AM action than in Test 6-1 is studied by using symptom-based plant parameters such as a reactor vessel water level detection.

Modeling of Hydrogen-Air Detonations in the TONUS CFD Code and Its Application to the FLAME F-19 Test

In severe accident management, the ability to predict pressure and thermal loads resulting from hydrogen combustion is important since they may threaten containment integrity. In computational modeling, different combustion regimes have to be accounted for and state-of-the-art techniques developed for reliable analysis. In the present study, the focus is on computational fluid dynamics code validation for reactive flows in the detonation regime. The FLAME hydrogen combustion test F-19 performed at the Sandia National Laboratories has been simulated by using the gas detonation model implemented in the TONUS CFD code which is developed by CEA and IRSN (France). In this model the reactive Euler equations are solved and the reaction rate is obtained by the Arrhenius global rate equation. Several simulations were run in order to examine the effect of modifying the parameters of the chemistry model. A mesh convergence study was performed for the purpose of finding out the necessary mesh resolution which could capture the detonation propagation with adequate accuracy. In addition, Chapman-Jouguet post-shock equilibrium conditions and the ZND detonation structure for the present gas mixture were examined by chemical kinetics calculations. The CFD simulation results were compared to the test results and the Chapman-Jouguet post-shock conditions. It was observed that the computational results differ from the C-J results with the C-J velocity being slightly exceeded. The model parameter study showed that it is not possible to significantly affect the flame propagation by adjusting the model parameters.

Dynamic Sensitivity Analysis of Thermal-Mechanical Deformation of a CANDU Fuel Channel

In a limiting critical break loss of coolant accident in a CANDU reactor significant degradation of heat transfer from the fuel can occur. As a result of the subsequent increase in fuel temperature it is possible that the pressure tube undergoes heat up at intermediate pressure during blowdown. This can result in ballooning deformation of the pressure tube into contact with its calandria tube. It is required that fuel channels not fail as a consequence of the thermal mechanical deformation of the pressure tube and calandria tube in such events. Dynamic sensitivity functions are derived as analytical partial differential equations derived from the equations used to model the time-dependent behavior of physical systems. The dynamic sensitivity functions can be used to propagate uncertainties using a time-dependent perturbation approach in which the variations in a set of output variables, with respect to perturbations of the input parameters, are evaluated about reference response trajectories of the input parameters and associated output variables. The dynamic sensitivity method is described in this paper and results are presented for the pressure tube heatup phase of a LOCA. These results show the importance of all key parameters with respect to specified safety evaluation criteria. The dynamic sensitivity method is applied in a probabilistic uncertainty analysis to evaluate the probability of a pressure tube experiencing creep strain deformation to contact its calandria tube during the early stages of a LOCA.

Assessment of the Airborne Release Coefficient During Cutting Operations in the Frame Dismantling Studies

In the frame of the decommissioning of nuclear facilities, many cutting techniques have been studied in different configurations, in order to qualify the performance of the tools implemented and also the quantity and characteristics of the wastes and secondary emissions produced. The present paper, calling the operational conditions of each tool, presents a synthesis of the results obtained from the suspension or airborne release point of view. The knowledge of the suspension fraction of radioactive aerosol contamination is needed for the safety analysis of nuclear installation and for the radioprotection of workers as well. This synthesis shows that the suspension fraction can vary over several orders of magnitude according to the configurations studied and provides relevant information for the dismantlers.

TRU-Fueled VHTRs for Applications Requiring an Extended Operation With Minimized Control and No Refueling

This paper presents an analysis of TRU-fueled VHTR systems focusing on applications requiring an extended operation with minimized control and no refueling (single-batch mode). As an example of such applications, international deployment opportunities for grid-appropriate VHTR systems could be mentioned addressing demands for electricity, industrial heat and co-generation in those regions where minimized servicing is desirable for various reasons. The study is performed for the hexagonal block core concept within the framework of the ongoing U.S. DOE NERI Project on utilization of higher actinides (TRUs and partitioned MAs) as a fuel component for extended-life VHTRs. The up-to-date analysis has shown reasonable reactivity swings, core life limits with respect to fast fluences and criticality.

Piping Supports Design by Analysis and Computer Program Development for NPPs

There are over ten thousand piping supports in a NPP. Piping supports are generally divided into two groups, one is standard and the other is non-standard. No matter to which group the support belongs, should it be proved to withstand the loads acting on it with sufficient design margin following the requirements of ASME or AISC code. With the development of the computer soft ware and hardware, we use 3D mathematic model to do stress analysis for the support. The model is a combination of linear element and nonlinear element, elastic material properties and inelastic material properties. For this composite model, we use plastic limit load analysis to determine assessment section, and use linearizing method to evaluate the section, and use simplified method to model nonlinear boundary conditions. At the end, a computer program is accomplished to perform analysis and evaluation for the supports efficiently and rapidly. The program not only can be used to establish standard support manual, but also can be used to perform analysis and evaluation for the non-standard support in the field.

Investigation on Header Manifold Model Effect at CANDU6 Large Loss of Cooling Accident

In a CANDU 6 nuclear power reactor fuel bundles are supported in horizontal Zircaloy pressure tubes tube through which the heavy-water coolant flows. 95 pressure tubes are connected by individual feeders to a common header. For CANDU 6 safety analyses, even when multiple channels model is employed, only one node is used for header. In this approach, all the channels are subjected to the same boundary condition. However, site inlet and outlet header pressure measurements and ultrasonic feeder flow data, confirm the existence of axial pressure gradients along the inlet and outlet headers. These axial pressure gradients would give rise to individual header-to-header pressure drops for each channel and also to flow distribution throughout both the inlet and outlet headers. In this paper, the header manifold model effect on the large break loss of coolant accident analyses of CANDU reactors has been performed by RELAP5/ mod 3.4 code. The 35% reactor inlet header break was selected for this study, as this break size produce the highest fuel clad temperature among all postulated breaks size. The results obtained considering the header manifold model, show that location of fuel channel upon break location has a strong impact on peak clad temperature calculation.