Stand-Alone Containment Analysis of the PHÉBUS FPT-1 Test With the ASTEC and the MELCOR Codes

The estimation of Fission Products (FPs) release from the containment system of a nuclear plant to the external environment during a Severe Accident (SA) is a quite complex task. In the last 30–40 years several efforts were made to understand and to investigate the different phenomena occurring in such a kind of accidents in the primary coolant system and in the containment. These researches moved along two tracks: understanding of involved phenomenologies through the execution of different experiments, and creation of numerical codes capable to simulate such phenomena. These codes are continuously developed to reflect the actual SA state-of-the-art, but it is necessary to continuously check that modifications and improvements are able to increase the quality of the obtained results. For this purpose, a continuous verification and validation work should be carried out. Therefore, the aim of the present work is to re-analyze the Phébus FPT-1 test employing the ASTEC (F) and MELCOR (USA) codes. The analysis focuses on the stand-alone containment aspects of the test, and three different modellisations of the containment vessel have been developed showing that at least 15/20 Control Volumes (CVs) are necessary for the spatial schematization to correctly predict thermal-hydraulics and the aerosol behavior. Furthermore, the paper summarizes the main thermal-hydraulic results, and presents different sensitivity analyses carried out on the aerosols and FPs behavior.

A High Effective Parallel Method for the Coupling Between Neutronics and Thermal-Hydraulic

Multi-physics coupling analysis is one of the most important fields among the analysis of nuclear power plant. The basis of multi-physics coupling is the coupling between neutronics and thermal-hydraulic because it plays a decisive role in the computation of reactor power, outlet temperature of the reactor core and pressure of vessel, which determines the economy and security of the nuclear power plant. This paper develops a coupling method which uses OPENFOAM and the REMARK code. OPENFOAM is a 3-dimension CFD open-source code for thermal-hydraulic, and the REMARK code (produced by GSE Systems) is a real-time simulation multi-group core model for neutronics while it solves diffusion equations. Additionally, a coupled computation using these two codes is new and has not been done. The method is tested and verified using data of the QINSHAN Phase II typical nuclear reactor which will have 16 × 121 elements. The coupled code has been modified to adapt unlimited CPUs after parallelization. With the further development and additional testing, this coupling method has the potential to extend to a more large-scale and accurate computation.

Jet Breakup and Droplet Formation in Immiscible Liquid-Liquid System

Mitigative measures against the event of a core disruptive accident (CDA) are of the importance from the viewpoint of safety of a sodium-cooled fast reactor (SFR). If the CDA occurs, the so-called post-accident heat removal must be surely achieved. The present study focuses on the scenario that the molten materials are injected into the lower plenum as jets. The jet breakup behavior during the CDA will be very complicated. Therefore, a specialized study on the fundamental process during the jet breakup is believed to be an effective approach. The aim of this paper is to understand the fundamental process of hydrodynamic interaction of jet breakup and droplet formation Using the immiscible liquid-liquid system, water and silicon oil as the test fluids, visualization via high-speed videography was performed. From the visualization results, the breakup length and droplet diameter were measured by image processing. The experimental data were scaled with ambient Weber number. When the Weber number was smaller than 1, the droplet diameter was close to the nozzle diameter, and distribution of droplet size was not observed. When the Weber number exceeded 1, the breakup length became longer and the generated droplet diameter possessed a distribution with two peaks due to satellite droplet formation. In both cases, the droplet formed at the leading edge of jet. In case that Weber number is around 100, the droplets were formed by entrainment of interfacial wave at jet side. From the mass median diameter data, we can see that the increase of the Weber number caused the decrease of median diameter and the increase of the width of the distribution.

Guidance for Human Action Evaluations for External Flood Risk Assessment

For many years external flooding hazards have been recognized as significant contributors to plant risk. However, it was not until the events at Fukushima that there was a concerted effort on the part of the utilities to reassess the plant external flood design basis, identify external flood vulnerabilities and take actions to address them. For many plants, resolution of low probability high consequence floods will likely be addressed by a combination of actions involving enhancements to flood protection and hazard mitigation strategies. Over time, as plants decide on which strategies to apply there is an expectation that the most effective way to develop and justify these strategies will involve probabilistic risk assessment (PRA) concepts. The PRA framework is well suited for performing a human reliability analysis (HRA). Within that framework, HRA evaluations focus on operator and plant staff actions taken in response to plant initiating events (e.g., loss of offsite power, etc.). For many external floods, advance warning of an impending external flood event provides the trigger for pre-emptive manual actions to potentially reconfigure the plant through temporary installation of flood barriers. Unlike the post-initiator actions which tend to be more narrowly focused, these pre-emptive actions are taken in a less controlled environment, may be ad hoc, and may potentially be in competition with site investment protection activities, site evacuation, etc. The purpose of this paper is to define the challenges in defining an approach for treating external flood actions, identifying external flood timelines, identifying the manual actions/organizational environment during external flooding scenarios and proposing an integrated strategy for quantifying those actions. The proposed quantification process is rooted in management science concepts for evaluating project reliability. The overall methodology identifies flood significant performance shaping factors, and identifies three (3) factors, namely time available for flood mitigation, proper access to plant site following flood and environmental factors, as having an overarching impact on the performance shaping factors affecting each of the flood mitigation tasks.

Measurement of Flow Rate of Droplets and Liquid Film in Venturi Scrubber

As one of filtered venting systems which should be installed in light water reactors from viewpoint of protecting a containment vessel and suppressing the diffusion of radioactive materials, there is a system composed of venturi scrubbers. The radioactive materials in the contaminated gas are collected into liquid. By dispersed flow formed in the venturi scrubber, large interfacial area between liquid and gas was obtained, and large decontamination factor is realized. In evaluation for the decontamination performance of the venturi scrubber, interface of droplets and liquid film are important. However, there is a little knowledge about the interfacial area in the venturi scrubber for filtered venting. In this study, to obtain the interfacial area data, amount of the droplets and the liquid film in the venturi scrubber is evaluated by visualizing observation and sampling the liquid film at the outlet of the venturi scrubber. In the venturi scrubber, a pressure drop occurs in the throat part by the inflow of air from the compressor. Water flows from the tank by a pressure difference between a suctioned hole with head pressure and a throat part. An annular spray flow is then formed in the venturi scrubber. Therefore, the liquid flow rate changes according to the gas phase flow rate. To discharge separately the droplets and the liquid film, a rectangular separator is installed at the venturi scrubber outlet. The superficial gas phase flow rate is 25.2–292.3 m/s in the throat. As a result, the liquid film and the droplets through the wall were confirmed to be discharged separately by the separator. The ratio of the liquid film to the total amount of liquid is 80 to 95% and that of the droplets was estimated as 5 to 20%. However, the change of the liquid film thickness caused by the increase of gas phase flow rates was observed. When the liquid film thickness is large, it is possible that some liquid film flowing into the droplet side.

Numerical Analysis of Hydrogen Risk Related Severe Accident Scenario in Chinese PWR

Based on the fact that the pressure loads generated in hydrogen combustion process may jeopardize the integrity of the containment during severe accident, and the changing rate as well as the maximum value of the pressure loads are governed by the flame propagation process, it is important to simulate the hydrogen combustion process with proper methodology. Due to the insufficiency understanding of the turbulent combustion and the difficulties of hydrogen combustion simulation in large scale and complex geometry, explosion safety applications are always based on simplified combustion model, for which the validation work and specified conservative parameter is required. In this study, an methodology combining CFD analysis and model validation based on large scale combustion experiments (HDR E12 and HYCOM01/02) is built up. And domestic hydrogen combustion process in the containment during severe accident is simulated. This study provides solid basis for structure design and integrity analysis of the containment.

Adaptive LMR Simulation of DDT in Obstructed Channel

Deflagration to detonation transition (DDT) is a challenging subject in computational fluid dynamics both from a standpoint of the phenomenon nature understanding and from extremely demanding computational efforts. In recent years, as the development of computer technology and improvement of numerical schemes was achieved, some more direct methods have been found to reproduce the DDT mechanistically without additional numerical or physical models. In the current work, highly resolved DDT simulations of hydrogen-air and of hydrogen-oxygen mixtures in 2D channel with regular repeating obstacles are present. The technique of local mesh refinement (ALMR) is implemented in the simulations to minimize the computational efforts. The criteria for the ALMR are examined and optimized in simulations.

Post-Fukushima Research and Development Strategies and Priorities for Water Cooled Reactor Technology Development

The Great East Japan Earthquake and the subsequent tsunami on 11 March 2011 initiated accident conditions at several nuclear power plants (NPPs) on the north-east coast of Japan and developed into a severe accident at the Fukushima Daiichi NPP, which highlighted a number of nuclear safety issues. After the Fukushima Daiichi accident, new research and development (R&D) activities have been undertaken by many countries and international organizations relating to severe accidents at NPPs. The IAEA held, in cooperation with the OECD/NEA, the International Experts’ Meeting (IEM) on “Strengthening Research and Development Effectiveness in the Light of the Accident at the Fukushima Daiichi Nuclear Power Plant” at IAEA Headquarters in Vienna, Austria, 16–20 February 2015. The objective of the IEM was to facilitate the exchange of information on these R&D activities and to further strengthen international collaboration among Member States and international organizations. One of the main conclusions of the IEM was that the Fukushima Daiichi accident had not identified completely new phenomena to be addressed, but that the existing strategies and priorities for R&D should be reconsidered. Significant R&D activities had been already performed regarding severe accidents of water cooled reactors (WCRs) before the accident, and the information was very useful for predicting and understanding the accident progression. However, the Fukushima Daiichi accident highlighted several challenges that should be addressed by reconsidering R&D strategies and priorities. Following this IEM, the IAEA invited several consultants to IAEA Headquarters, Vienna, Austria, 12–14 May 2015, and held a meeting in order to discuss proposals on possible IAEA activities to facilitate international R&D collaboration in relation to severe accidents and how to effectively disseminate the information obtained at the IEM. The IAEA also held Technical Meeting (TM) on “Post-Fukushima Research and Development Strategies and Priorities” at IAEA Headquarters, Vienna, Austria, 15–18 December 2015. The objective of this meeting was to provide a platform for experts from Member States and international organizations to exchange perspectives and information on strategies and priorities for R&D regarding the Fukushima Daiichi accident and severe accidents in general. The experts discussed R&D topic areas that need further attention and the benefits of possible international cooperation. This paper discusses lessons learned from the Fukushima Daiichi accident based on the presentations and discussions at the meetings mentioned above, and identifies the needs for further R&D activities to develop WCR technologies to cope with Fukushima Daiichi-type accidents.

CFD Analysis of Cross Flow in Rod Bundles Under Rolling Motion

A CFD analysis of cross flow in rod bundles in rolling motion was performed to investigate the effect of rolling motion on the flow behavior between the subchannels. The rolling motion was assumed as a sinusoid. The additional forces due to the rolling motion including azimuthal force, centrifugal force and coriolis force were added into the source term in the momentum equation. A transient three dimensional simulation of square rod bundles model was performed in various rolling conditions. In order to precisely predict secondary flow patterns in rod bundles, Reynolds Stress Model was selected as the turbulent model. Effect of various rolling parameters such as rolling velocity and amplitude on the cross mixing was investigated. The results show that cross flow is strongly affected by the rolling motion. The local cross flow field in rolling motion was showed in detail. Also, the mixing coefficients based on CFD results were calculated. Empirical correlations of turbulent mixing were modified to consider the effect of rolling conditions, which can be used in the traditional subchannel thermal hydraulic code.

Analysis of Flashing and Slow Descending Phenomenon in Ring Coelom of Integrated Reactor During LOCA

Integrated reactor has an integrated pressure vessel with all components of the main coolant loop inside. Ring coelom is formed between the pressure vessel and the heat exchanger. Several small-bore pipes are plugged into the ring coelom to attain the coolant. Because of the existence of the isolated ring coelom, slow descending phenomenon of liquid level in the ring coelom takes place during LOCA with a coolant diversion pipe breaking. The phenomenon is analyzed by modelling the reactor during LOCA in Relap5 and one mitigation measure is acquired. As drastic flashing seriously influences the slow descending phenomenon, which enhance the pressure in the ring coelom, we divide the ring coelom into several parts and use Fluent to gain the flashing details. The results of Relap5 and Fluent show good agreement, which proves the flashing and slow descending phenomenon in the ring coelom of integrated reactor is reasonable during LOCA.