The European Journal of Physics N (EPJ-N)
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A new MV bus transfer scheme for nuclear power plants
No full textFast bus transfer method is the most popular and residual voltage transfer method that is used as a backup in medium voltage buses in general. The use of the advanced technology like open circuit voltage prediction and digital signal processing algorithms can improve the reliability of fast transfer scheme. However, according to the survey results of the recent operation records in nuclear power plants, there were many instances where the fast transfer scheme has failed. To assure bus transfer in any conditions and circumstances, uninterruptible bus transfer scheme utilizing the state of the art medium voltage UPS is discussed and elaborated
Possible in-vessel corium progression way in the Unit 1 of Fukushima Dai-ichi nuclear power plant using a phenomenological analysis
No full textIn the field of severe accident, the description of corium progression events is mainly carried out by using integral calculation codes. However, these tools are usually based on bounding assumptions because of high complexity of phenomena. The limitations associated with bounding situations ([J.M. Seiler, B. Tourniaire, A phenomenological analysis of melt progression in the lower head of a pressurized water reactor, Nucl. Eng. Des. 268, 87 (2014)] e.g. steady state situations and instantaneous whole core relocation in the lower head) led CEA to develop an alternative approach in order to improve the phenomenological description of melt progression. The methodology used to describe the corium progression was designed to cover the accidental situations from the core meltdown to the molten core concrete interaction. This phenomenological approach is based on available data (including learnings from TMI2), on physical models and knowledge about the corium behavior. It provides emerging trends and best estimated intermediate situations. As different phenomena are unknown, but strongly coupled, uncertainties at large scale for the reactor application must be taken into account. Furthermore, the analysis is complicated by the fact that these configurations are most probably three dimensional, all the more so because 3D effects are expected to have significant consequences for the corium progression and the resulting vessel failure. Such an analysis of the in-vessel melt progression was carried out for the Unit 1 of the Fukushima Dai-ichi nuclear power plant. The core uncovering kinetics governs the core degradation and impacts the appearance of the first molten corium inside the core. The initial conditions used to carry out this analysis are based on available results derived from codes like MELCOR calculation code [R. Ganntt, D. Kalinich, J. Cardoni, J. Phillips, A. Goldmann, S. Pickering, M. Francis, K. Robb, L. Ott, D. Wang, C. Smith, S. St. Germain, D. Schwieder, S. Phelan, Fukushima Daiichi Accident Study (Status as of April 2012), Sandia Report Sand 2012-6173, Unlimited Release Printed August, 2012]. The core degradation could then follow different ways: axial progression of the debris and the molten fuel through the lower support plate; lateral progression of the molten fuel through the shroud. On the basis of the Bali program results [J.M. Bonnet, An integral model for the calculation of heat flux distribution in a pool with internal heat generation, in Nureth7 530 Conference Saratoga Springs, NY, USA, September 10–15, 1995 (1995)] and the TMI-2 accident observations [D.W. Ackers, J.R. Wolf, Relocation of Fuel Debris to the Lower Head of the TMI2 Reactor Vessel-A possible scenario, TMI 2 pressure vessel investigation project, in Proceedings of the Open forum OECD/NEA and USNRCm, Boston, USA, 20–22 October 1993 (1993)], this work is focused on the consequences of a lateral melt progression (not excluding an axial progression through the support plate). Analysis of the events and the associated time sequence will be detailed. Besides, this analysis identifies a number of issues. Random calculations and statistical analysis of the results could be performed with calculation codes such as LEONAR–PROCOR codes [R. Le Tellier, L. Saas, F. Payot, Phenomenological analyses of corium propagation in LWRs: the PROCOR software platform, in ERMSAR 2015, Marseille, France, 24–26 March, 2015 (2015)]
Thermal hydraulic simulations of the Angra 2 PWR
No full textAngra 2, the second Brazilian nuclear power plant, began the commercial operation in 2001. The plant is a pressurized water reactor (PWR) type with electrical output of about 1350 MW. In the present work, the thermal hydraulic RELAP5-3D code was used to develop a model of this reactor. The model was performed using geometrical and material data from the Angra 2 final safety analysis report (FSAR). Simulations of the reactor behavior during steady state and loss of coolant accident were performed. Results of temperature distribution within the core, inlet and outlet coolant temperatures, coolant mass flow, and other parameters have been compared with the reference data and demonstrated to be in good agreement with each other. This study demonstrates that the developed RELAP5-3D model is capable of reproducing the thermal hydraulic behavior of the Angra 2 PWR and it can contribute to the process of the plant safety analysis
Beyond designed functional margins in CANDU type NPP. Radioactive nuclei assessment in an LOCA type accident
No full textEuropean Union's energy roadmap up to year 2050 states that in order to have an efficient and sustainable economy, with minimum or decreasing greenhouse gas emissions, along with use of renewable resources, each constituent state has the option for nuclear energy production as one desirable option. Every scenario considered for tackling climate change issues, along with security of supply positions the nuclear energy as a recommended option, an option that is highly competitive with respect to others. Nuclear energy, along with other renewable power sources are considered to be the main pillars in the energy sector for greenhouse gas emission mitigation at European level. European Union considers that nuclear energy must be treated as a highly recommended option since it can contribute to security of energy supply. Romania showed excellent track-records in operating in a safe and economically sound manner of Cernavoda NPP Units 1&2. Both Units are in top 10 worldwide in terms of capacity factor. Due to Romania's need to ensure the security of electricity supply, to meet the environmental targets and to move to low carbon generation technologies, Cernavoda Units 3&4 Project appears as a must. This Project was started in 2010 and it is expected to have the Units running by 2025. Cost effective and safety operation of a Nuclear Power Plant is made taking into consideration functional limits of its equipment. As common practice, every nuclear reactor type (technology used) is tested according to the worse credible accident or equipment failure that can occur. For CANDU type reactor, this is a Loss of Cooling Accident (LOCA). In a LOCA type accident in a CANDU NPP, using RELAP/SCDAP code for fuel bundle damage assessment the radioactive nuclei are to be quantified. Recently, CANDU type NPP accidents are studied using the RELAP/SCDAP code only. The code formerly developed for PWR type reactors was adapted for the CANDU geometry and can assess the accident progression consequences up to a certain point. The code assesses the fuel bundle damage progression, but cannot assess further core damage for a CANDU type core, and starting from these data the amount of damaged fuel can be calculated. The radio nuclei present in the damaged fuel are supposed to be released into the main heat transport system and after that into the containment building in the worst case scenario. Assessing the radioactive nuclei maximum release is the purpose of the present paper. The radioactive nuclei release is needed for the accident management plan, limiting the environmental and population impact of the supposed accident, and furthermore for a later site remediation plan that can be put in action after the complete mitigation of the accident consequences. The maximum quantity of radio nuclei released during the accident calculated in this paper is a worst case scenario evaluation that can lead to better preparedness in an accident scenario
Evaluation of relevant information for optimal reflector modeling through data assimilation procedures
No full textThe goal of this study is to look after the amount of information that is mandatory to get a relevant parameters optimisation by data assimilation for physical models in neutronic diffusion calculations, and to determine what is the best information to reach the optimum of accuracy at the cheapest cost. To evaluate the quality of the optimisation, we study the covariance matrix that represents the accuracy of the optimised parameter. This matrix is a classical output of the data assimilation procedure, and it is the main information about accuracy and sensitivity of the parameter optimal determination. From these studies, we present some results collected from the neutronic simulation of nuclear power plants. On the basis of the configuration studies, it has been shown that with data assimilation we can determine a global strategy to optimise the quality of the result with respect to the amount of information provided. The consequence of this is a cost reduction in terms of measurement and/or computing time with respect to the basic approach
Cooling the intact loop of primary heat transport system using Shutdown Cooling System in case of LOCA events
No full textThe purpose of this paper is to model the operation of the Shutdown Cooling System (SDCS) for CANDU 6 nuclear power plants in case of LOCA accidents, using Flowmaster calculation code, by delimiting models and setting calculation assumptions, input data for hydraulic analysis and input data for calculating thermal performance check for heat exchangers that are part of this system
Experimental facility for development of high-temperature reactor technology: instrumentation needs and challenges
No full textA high-temperature, multi-fluid, multi-loop test facility is under development at the Idaho National Laboratory for support of thermal hydraulic materials, and system integration research for high-temperature reactors. The experimental facility includes a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions. Each loop will also include an interchangeable high-temperature test section that can be customized to address specific research issues associated with each working fluid. This paper also discusses needs and challenges associated with advanced instrumentation for the multi-loop facility, which could be further applied to advanced high-temperature reactors. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. A preliminary design configuration of the ARTIST facility will be presented with the required design and operating characteristics of the various components. The initial configuration will include a high-temperature (750 °C), high-pressure (7 MPa) helium loop thermally integrated with a molten fluoride salt (KF-ZrF4) flow loop operating at low pressure (0.2 MPa), at a temperature of ∼450 °C. The salt loop will be thermally integrated with the steam/water loop operating at PWR conditions. Experiment design challenges include identifying suitable materials and components that will withstand the required loop operating conditions. The instrumentation needs to be highly accurate (negligible drift) in measuring operational data for extended periods of times, as data collected will be used for code and model verification and validation, one of the key purposes for the loop. The experimental facility will provide a much-needed database for successful development of advanced reactors and provide insight into the needs and challenges in instrumentation for advanced high-temperature reactors
Safety operation of chromatography column system with discharging hydrogen radiolytically generated
No full textIn the extraction chromatography system, accumulation of hydrogen gas in the chromatography column is suspected to lead to fire or explosion. In order to prevent the hazardous accidents, it is necessary to evaluate behaviors of gas radiolytically generated inside the column. In this study, behaviors of gas inside the extraction chromatography column were investigated through experiments and Computation Fluid Dynamics (CFD) simulation. N2 gas once accumulated as bubbles in the packed bed was hardly discharged by the flow of mobile phase. However, the CFD simulation and X-ray imaging on γ-ray irradiated column revealed that during operation the hydrogen gas generated in the column was dissolved into the mobile phase without accumulation and discharged
Helium behaviour in implanted boron carbide
No full textWhen boron carbide is used as a neutron absorber in nuclear power plants, large quantities of helium are produced. To simulate the gas behaviour, helium implantations were carried out in boron carbide. The samples were then annealed up to 1500 °C in order to observe the influence of temperature and duration of annealing. The determination of the helium diffusion coefficient was carried out using the 3He(d,p)4He nuclear reaction (NRA method). From the evolution of the width of implanted 3He helium profiles (fluence 1 × 1015/cm2, 3 MeV corresponding to a maximum helium concentration of about 1020/cm3) as a function of annealing temperatures, an Arrhenius diagram was plotted and an apparent diffusion coefficient was deduced (Ea = 0.52 ± 0.11 eV/atom). The dynamic of helium clusters was observed by transmission electron microscopy (TEM) of samples implanted with 1.5 × 1016/cm2, 2.8 to 3 MeV 4He ions, leading to an implanted slab about 1 μm wide with a maximum helium concentration of about 1021/cm3. After annealing at 900 °C and 1100 °C, small (5–20 nm) flat oriented bubbles appeared in the grain, then at the grain boundaries. At 1500 °C, due to long-range diffusion, intra-granular bubbles were no longer observed; helium segregates at the grain boundaries, either as bubbles or inducing grain boundaries opening