Loss-of-coolant accident | EPFL Graph Search

A loss-of-coolant accident (LOCA) is a mode of failure for a nuclear reactor; if not managed effectively, the results of a LOCA could result in reactor core damage. Each nuclear plant's emergency core cooling system (ECCS) exists specifically to deal with a LOCA. Nuclear reactors generate heat internally; to remove this heat and convert it into useful electrical power, a coolant system is used. If this coolant flow is reduced, or lost altogether, the nuclear reactor's emergency shutdown system is designed to stop the fission chain reaction. However, due to radioactive decay, the nuclear fuel will continue to generate a significant amount of heat. The decay heat produced by a reactor shutdown from full power is initially equivalent to about 5 to 6% of the thermal rating of the reactor. If all of the independent cooling trains of the ECCS fail to operate as designed, this heat can increase the fuel temperature to the point of damaging the reactor.
*If water is present, it may

Modelling of fuel fragmentation, relocation and dispersal during Loss-of-Coolant Accident in Light Water Reactor

Vladimir Vladimirov Brankov

Recent LOCA tests with high burnup fuel at the OECD Halden Reactor Project and at Studsvik demonstrated the susceptibility of the fuel to fragment to small pieces, to relocate and possibly cause a hot-spot effect and to be dispersed in the event of cladding rupture. However, the LOCA safety criteria defined by the US NRC are still based on fuel tests with fresh and low burnup fuel and therefore require revision for high burnup fuel. In this context a PhD project with the goal of developing new models for high burnup fuel fragmentation, relocation and dispersal during Loss of Coolant Accident in Light Water Reactors was launched at Paul Scherrer Institute in June 2013 with a financial support from swissnuclear. The work continued with base irradiation simulation with a closer look at the fission gas release measurements of high burnup BWR fuel rods.The data showed significant scatter for fuel rods at the same average burnup, originating from the same fuel assembly and having almost the same enrichment. The scatter was explained with a BWR-specific mechanism for fission gas trapping. It was motivated with Scanning Electron Microscopy images at the pellet-cladding interface that showed very strong fuel-cladding bonding layer on one side of the fuel rod. Base irradiation with the EPRI's FALCON code coupled with an in-house advanced fission gas release and gaseous swelling model GRSW-A was done for selected high burnup BWR fuel rods. The calculated fission gas release was overestimated in all cases. This was expected, because at the time the modelling did not account for fission gas trapping. From the available modelling and experimental data, a model for fission gas trapping was proposed. After calibration, the agreement between calculated and measured fission gas release was significantly improved. Fuel fragmentation modelling was focused on fuel pulverization. This is a high burnup fuel-specific phenomenon, in which the fuel pellet periphery may fragment to sizes less than 100 ÎŒm. Such fragments are very mobile and can be easily relocated and dispersed in the event of cladding rupture regardless of the rupture opening size. Therefore, the fuel fragmentation model addresses the potentially most important mode of fragmentation - fuel pulverization. Fuel relocation inside fuel rod is simulated during cladding ballooning and until cladding failure. The model takes as input the time-dependent cladding deformation supplied by FALCON and the fragment size distribution either provided directly by experimental data or by the fuel fragmentation model. Fuel relocation was modelled under the specific assumption that outermost fragments (e.g. pulverized fuel) relocate first, resulting in a large packing factor and local cladding temperature increase at the balloon (i.e. hot-spot effect) and, as a consequence, in enhanced cladding oxidation at the balloon. Fuel dispersal is modelled by solving the mass, energy and momentum conservation equations for two phases - the gas inside the fuel rod and the fraction of fuel which is considered "moveable". Although the model uses simplified geometrical representation of the fuel rod and some other simplifying assumptions, the underlying reason for fuel dispersal (namely the interfacial friction between the gas outflow and the solid) is explicitly simulated. The model for fuel dispersal is calibrated using Halden and Studsvik LOCA tests.

EPFL2017

Improvement of the Decay Heat Removal Characteristics of the Generation IV Gas-cooled Fast Reactor

Aaron Simon Epiney

Gas cooling in nuclear power plants (NPPs) has a long history, the corresponding reactor types developed in France, the UK and the US having been thermal neutron-spectrum systems using graphite as the moderator. The majority of NPPs worldwide, however, are currently light water reactors, using ordinary water as both coolant and moderator. These NPPs – of the so-called second generation – will soon need replacement, and a third generation is now being made available, offering increased safety while still based on light water technology. For the longer-term future, viz. beyond the year 2030, R&D is currently ongoing on Generation IV NPPs, aimed at achieving closure of the nuclear fuel cycle, and hence both drastically improved utilization of fuel resources and minimization of long-lived radioactive wastes. Since the very beginning of the international cooperation on Generation IV, viz. the year 2000, the main research interest in Europe as regards the advanced fast-spectrum systems needed for achieving complete fuel cycle closure, has been for the Sodium-cooled Fast Reactor (SFR). However, the Gas-cooled Fast Reactor (GFR) is currently considered as the main back-up solution. Like the SFR, the GFR is an efficient breeder, also able to work as iso-breeder using simply natural uranium as feed and producing waste which is predominantly in the form of fission products. The main drawback of the GFR is the difficulty to evacuate decay heat following a loss-of-coolant accident (LOCA) due to the low thermal inertia of the core, as well as to the low coolant density. The present doctoral research focuses on the improvement of decay heat removal (DHR) for the Generation-IV GFR. The reference GFR system design considered in the thesis is the 2006 CEA concept, with a power of 2400 MWth. The CEA 2006 DHR strategy foresees, in all accidental cases (independent of the system pressure), that the reactor is shut down. For high-pressure events, dedicated DHR loops with blowers and heat exchangers are designed to operate when the power conversion system cannot be used to provide acceptable core temperatures under natural convection conditions. For depressurized events, the strategy relies on a dedicated small containment (called the guard containment) providing an intermediate back-up pressure. The DHR blowers, designed to work under these pressure conditions, need to be powered either by the power grid or by batteries for at least 24 hours. The specific contributions of the present research – aimed at achieving enhanced passivity of the DHR system for the GFR – are design and analysis related to (1) the injection of heavy gas into the primary circuit after a LOCA, to enable natural convection cooling at an intermediate-pressure level, and (2) an autonomous Brayton loop to evacuate decay heat at low primary pressure in case of a loss of the guard-containment pressure. Both these developments reduce the dependence on blower power availability considerably. First, the thermal-hydraulic codes used in the study – TRACE and CATHARE – are validated for gas cooling. The validation includes benchmark comparisons between the codes, serving to identify the sensitivity of the results to the different modeling assumptions. The parameters found to be the most sensitive in this analysis, such as heat transfer and friction models, are then validated via a detailed re-analysis of earlier PSI (EIR, at the time) gas-loop experiments conducted in the 1970s. Conclusions and recommendations on the models to be used for transient analysis are derived. In general, it has been shown that the agreement, between experiments and the correlations for heat transfer and friction used in TRACE and CATHARE, is quite satisfactory. The thus validated codes are then used in the two detailed, DHR improvement studies carried out. The first improvement of the reference DHR strategy is the heavy gas injection. Assuming a DHR blower failure after a LOCA, the helium pressure in the guard containment is not high enough to evacuate the decay heat by natural convection. To improve the natural convection, the effects of injecting different heavy gases (N2, CO2, Ar and a N2/He mixture) into the primary circuit were analyzed, in order to address the possibility of dealing with DHR-blower failure while accepting an intermediate back-up pressure in the guard containment. Furthermore, different injection locations and injection mass flows were considered, and the sensitivity to the number of available DHR loops and LOCA break-sizes was also addressed. It has been found that injecting the heavy gas in the vicinity of the core could lead to overcooling problems. For an injection point sufficiently far from the core, however, both CO2 and N2 are found to be able to cool the core satisfactorily in natural convection. N2 is proposed as the reference, due to possible chemical problems with CO2. The second proposition for DHR improvement is related to the possibility of a simultaneous guard-containment failure, i.e. a loss-of-back-up-pressure (LOBP) combined with a blower failure after a LOCA. In this case the natural convection, even with heavy gas injection, is no longer strong enough to evacuate the decay heat. To address this issue, the possibility of decay heat removal via use of a dedicated autonomous Brayton cycle – as a standalone DHR loop – has been investigated. First, an analytical Brayton cycle model has been set up, so as to identify convenient machine design points and to study the machine's off-design behavior. Two machine designs have then been drawn up: one for helium in order to provide a reference for understanding the Brayton loop behavior in a generic sense, and the other for nitrogen which is the envisaged gas to be injected after a LOCA. Both, the design of the proposed devices and their validation are discussed. Finally, a detailed transient analysis, involving usage of both heavy-gas injection and the Brayton device (i.e. of the complete, proposed DHR system), is presented. This serves to illustrate the effectiveness of the new strategy for the highly hypothetical worst-case scenario of sequential failures following a LOCA.

EPFL2010

Heavy-gas injection in the Generation IV gas-cooled fast reactor for improved decay heat removal under depressurized conditions

Rakesh Chawla, Aaron Simon Epiney, Konstantin Mikityuk

Decay heat removal is a key safety and design issue for the Generation IV gas (helium)-cooled fast reactor. This paper investigates the natural convection capability of the dedicated DHR loops under depressurized conditions while injecting a heavy gas into the system. Investigated is a loss-of-coolant accident using the TRACE code. The goal of the study is to improve fuel/cladding temperature behavior during LOCA transients with the enhancement of passive safety by operation in natural convection only, while accepting 10 bar back-up pressure in the guard containment. The paper investigates the cooling capabilities of different heavy gases. Furthermore, different injection locations and mass flow rates have been tested, in order to address possible core-overcooling problems resulting from rapid depressurization of the gas reservoir. It has been shown that, among the gases investigated, CO2 is the best choice from the thermal-hydraulics viewpoint, being able to cool the core satisfactorily for a broad range of injection rates. N-2 can be envisaged as an alternative solution in case of chemical problems with CO2. Supplementary studies carried out for the CO2 and N-2 injection cases include that of the sensitivity to the number of available DHR loops and to the LOCA break-size. The effect of the resulting neutron spectrum changes on the shutdown-reactivity margin has also been investigated. (C) 2010 Elsevier B.V. All rights reserved.