• Energy Research
• Nuclear Engineering and Design close

Abstract AP1000® Generation III+ reactor bases its safety concept on passive systems, differently from the previous Generation II reactors. This fact has led the approximations and methodologies previously used for modeling active safety systems to be reviewed and adapted to simulate the physics of passive systems. Diverse studies about the AP1000 containment have demonstrated the difficulty to correctly model the occurring phenomenology. In this paper, an integral AP1000 3D containment GOTHIC model is presented, including the Passive Containment Cooling System (PCCS). The model includes the compartments inside and outside the metallic containment liner that influence the thermal–hydraulic behavior. The model is tested against a Large Break Loss of Coolant Accident (LBLOCA) to assess its thermal–hydraulic performance, assuming a PCS tank malfunction, what is a conservative hypothesis. The pressure and temperature evolution predicted by the 3D containment model is analyzed and compared with a single node Lumped Parameters model, allowing to evaluate some preliminary benefits of 3D modeling for containment safety analysis. The 3D containment model allows to predict the thermal evolution in each containment compartment capturing the heterogeneity of this phenomenon, with higher resolution than the lumped parameters models traditionally used in this kind of analyses. It allows to observe the thermohydraulic conditions locally at any time during the transient.

Abstract A novel multi-scale domain overlapping coupling methodology designed to couple a computational fluid dynamics (CFD) code with a system thermal hydraulic (STH) code was developed and its performance was investigated. The methodology has been implemented in the coupling infrastructure code Janus, developed at the University of Michigan, providing methods for the on-the-fly data transfer through memory between the commercial CFD code STAR-CCM+ and the US NRC best-estimate thermal hydraulic system code TRACE. Coupling between these two software packages is motivated by the desire to extend the range of applicability of TRACE to scenarios in which local momentum and energy transfer are important, such as three-dimensional mixing of localized slugs of deborated or cold water in the downcomer and lower plenum of a reactor pressure vessel. The intra-fluid shear forces necessary to correctly capture these effects are neglected in the TRACE equations of motion, but are readily calculated from CFD solutions. CFD/STH coupling implementations therefore have applications in reactor transients such as boron dilution scenarios, Anticipated Transient Without SCRAM (ATWS) and Main Steam Line Break (MSLB). The proposed method is based on aliasing all spatial sources and sinks of momentum in the CFD domain as frictional losses in the system code domain. The internal velocity fields and, consequently, the inertial component of the pressure field are maintained consistent between the CFD and STH domains through a complementary velocity-matching interface. In this paper, coupled simulations are performed on Cartesian and cylindrical geometry with emphasis on consistency, convergence, and stability during transient scenarios. Results show that the presented domain overlapping coupling method is capable of adjusting pressure and velocity profiles of multi-dimensional system code solutions to match CFD solutions accurately. Important characteristics of transient simulations were found to include the background flow rate, specifically the stabilizing effect of viscous forces, as well as the time derivative of the flow rate. Under certain adverse conditions, the basic coupling method is found to produce unstable behavior. A stabilization method for adjusting CFD data is laid out and found to significantly improve the method’s performance under the most challenging conditions. Recommendations are laid out for further improving the coupling via advanced time stepping methods.

Abstract In the context of light water reactor severe accidents analysis, this paper is focused on one key parameter of in-vessel corium phenomenology: the immiscible phases stratification and its impact on the heat flux distribution at the corium pool lateral boundary with the so-called focusing effect related to a “thin” top metal phase and the potential vessel failure at that point. More particularly, based on the limited knowledge of the stratification transient phenomenon derived from the MASCA-RCW experiment, a basic model is proposed that can be used for corium in lower head sensitivity analyses. It has been implemented in the PROCOR platform developed at CEA Cadarache. A short parametric study on a simple hypothetical transient is presented in order to highlight the different focusing effect “modes” that can be encountered based on this in-vessel corium pool model. An early mode may occur during the formation of the top metal layer while two other modes may appear later during the thinning of this top metal layer because of thermochemically induced mass transfers. Some associated relevant parameters (model or scenario-dependent) and modelling issues are mentioned and illustrated with some results of a Monte-Carlo based sensitivity calculation on the transient behaviour of the corium in the lower head of a 1650 MWe GenIII PWR. Within the limiting modelling hypotheses, the thermal modelling of the steel layer for small (centimetre) heights and the mass diffusivity (limited in this case to the uranium diffusivity in the oxidic layer) are main sensitive parameters.

The starting event of the massive air ingress into the core of the HTR module reactor, classified as hypothetical incident, is the very fast depressurization of the primary circuit. Provided that the integrity of the reactor pressure vessel is not in question, a rupture of the connecting pressure vessel between reactor pressure vessel and steam generator vessel is the maximum possible leak cross-section. In this work it is investigated whether the components of the reactor pressure vessel are exposed by the depressurization process to mechanical loads which exceed the load limits. These loads are caused by two different events, the strong momentum change of the fluid and the local pressure differences, respectively. Due to the momentum change the bottom reflector receives the maximum load, whereby only 2% of the compressive strength of the graphite quality used there are reached. However, the load by local pressure differences is between passed volumes and in normal operation, not-passed volumes lead to high load values. A maximum pressure difference of 44.5 bar was calculated at the thermal top shield.

Abstract Over the past seven years the use of seismic experience data to address seismic concerns has received a great deal of attention, particularly regarding the Nuclear Regulatory Commission Unresolved Safety Issue (USI) A-46. The Seismic Qualification Utility Group (SQUG) was formed in January 1982 to develop a practical alternative to the rigorous seismic qualification of equipment for resolution of USI A-46. The alternative method chosen is seismic experience technology. The purpose of this paper is to explore two additional potential applications of seismic experience technology: • bb Replacement parts and • bb New equipment/design change process. The need for, and benefits of, these applications will be summarized. The available technology and the methodologies proposed will be described. The methodology descriptions include a summary of the requirements for the seismic evaluations, an outline of the method, and the documentation requirements.

Abstract Experiments have been performed with 19- and 61-pin test assemblies in the Thermal-Hydraulic Out-of-Reactor Safety (THORS) Facility at the Oak Ridge National Laboratory (ORNL) since 1971. The THORS Facility is a high-temperature sodium system operated for the US Liquid-Metal Fast Breeder Reactor (LMFBR) Safety Program. The facility is used primarily for testing simulated LMFBR fuel subassemblies (pin (bundles). High-performance, electrically heated fuel pin simulators (FPSs) duplicate the heat generating capabilities and the dimensional characteristics of the nuclear fuel pins. A number of test bundles have been built and operated to obtain base thermal-hydraulic data, inlet and heated zone blockage data, and transient boiling data. Five of these bundles have been operated under two-phase conditions. Sodium boiling for periods up to twelve minutes were sustained in one bundle. (The lengths of the periods were limited only by automatic data recording capability). Clad dryout occurred in several tests. Tests were run at widely varying conditions of flow and power density. Testing with nonuniform power distribution across the bundle was also a part of the program. A 19-pin bundle with 12 peripheral guard heaters and a 6-subchannel blockage around the center pin in the heated zone was tested. The test program for this bundle was designed to determine if local boiling in the wake of the blockage propagates radially or axially during quasi-steady-state conditions. Post-test inspection revealed that significant helical distortion of the FPSs occurred in the vicinity of the blockage plate. This distortion probably influenced the boiling behavior. In the more severe tests, boiling initiated at the outlet of the heated zone and propagated radially into the unblocked subchannels after it had progressed upstream to the blockage. The subchannel analysis codes, SABRE and COBRA, accurately predict the extent of the boiling region. Experimental and analytical studies of sodium boiling behavior in unblocked 19- and 61-pin bundles indicate that cooling can be maintained for a significant period of time beyond boiling inception in a flow-power transient. Quasi-steady-state boiling occurred under natural-convection conditions. Investigations of the temperature data indicate that the thermal-hydraulic behavior during boiling transients is determined by two-dimensional effects, and that one-dimensional models cannot accurately predict the important phenomena associated with sodium boiling in test bundles. The subchannel code SABRE-2P (with a simple two-phase multiplier boiling model) and the two-region equilibrium mixture code THORAX (developed at ORNL) accurately predict the two-dimensional behavior between boiling inception and dryout. Extrapolation of the data from the smaller bundle tests to full-size fuel assemblies shows that the time between boiling inception and dryout would be lower for a 217-pin bundle than for a 61-pin bundle for a comparable transient. However, the time delay would still be significant, especially in a heterogeneous reactor core.

AbstractThe operating experience feedback is important for maintaining and improving safety and availability in nuclear power plants. Detailed investigation of all events is challenging since it requires excessive resources, especially in case of large event databases. This paper presents an event groups ranking method to complement the analysis of individual operating events. The basis for the method is the use of an internationally accepted events characterization scheme that allows different ways of events grouping and ranking. The ranking method itself consists of implementing the analytical hierarchy process (AHP) by means of a custom developed tool which allows events ranking based on ranking indexes pre-determined by expert judgment. Following the development phase, the tool was applied to analyze a complete set of 5 years of real nuclear power plants operating events (1453 events). The paper presents the potential of this ranking method to identify possible patterns throughout the event database and therefore to give additional insights into the events as well as to give quantitative input for the prioritization of further more detailed investigation of selected event groups.

Nuclear power generation is now one of the most important technologies in the energy supply in Japan. The operational experience of nuclear power plants for more than 20 years in Japan has shown that water chemistry is a key technology for safer and more economical operation of nuclear power plants. Extensive efforts have been made in the field of water chemistry to achieve the targets of radiation field control, to keep the integrities of piping and fuel elements and to reduce the radioactive waste generation. This paper briefly describes the present status of Japanese water chemistry technology with the main emphasis placed on BWRs.