The structural-phase state and residual lifetime of the mechanical properties in the regions where VVER fuel element cladding has become unsealed after tests on the PARAMETR stand under loss-of-coolant accident conditions are investigated. It is determined that deformation engenders breaks in the oxide film which promote oxidation. The structural and strength characteristics in the region of unsealing exhibit strong azimuthal nonuniformity, which is due to the nature of the deformation, thinning of the cladding wall, and the degree of damage to the oxide as result of deformation. The results presented show that it is very important to examine the processes leading to the deformation of fuel element cladding and the subsequent behavior of the regions of unsealing in a corrosive medium from the standpoint of the formation of the most vulnerable zone, which is responsible for the residual lifetime of the entire fuel element.The main requirements for assuring the safety of a VVER reactor facility are that the core can be cooled and disassembled at any operating stage [1]. The following conditions are necessary to satisfy these requirements:• the degree of swelling of the cladding (plastic strain and unsealing) must be limited;• the nominal limit on the damage to fuel elements must be satisfied;• there must be no fragmentation (multiple fracture) of fuel elements or melting of the fuel. It has been shown experimentally [2, 3] that during a loss-of-coolant accident with temperature rising to 1100°C unsealing of fuel element cladding as result of plastic strain is unavoidable. This has a substantial effect on the oxidation of cladding and on the temperature regime of the core because of the following:• the gap between the fuel and the cladding increases and heat removal from the fuel pellets is degraded;• the cladding becomes thinner and the total surface area of the fuel elements which is open to oxidation increases; • cladding becomes unsealed and the inner surface becomes oxidized. Such changes will accelerate the degradation of the structure and the mechanical properties of the cladding. Our objectives in the present work are to investigate the structure-phase state and the remaining life of the mechanical properties of regions where fuel element cladding has become unsealed after tests under loss-of-coolant accident conditions.Tests of fuel element simulators in superheated steam flow at 500°C and flow 0.2 g/sec on a fuel element were performed on the PARAMETR bench [4]. The test model of a VVER-1000 fuel assembly contained seven fuel elements with interior pressure 2 ± 0.25 MPa and 12 active fuel elements with internal pressures 0.1-0.2 MPa with a 1275 mm long heating part. The tungsten heaters were located inside the active fuel elements and were separated from the cladding by UO 2 pellets. The rate of heating was 0.5-1°C/sec. The maximum temperature of the cladding reached 1280-1300°C. The fuel assemblies were cooled by reserve water poured from below (simulation of repeated flooding of VVER), and the cladding w...
Investigations of the behavior of model VVER-1000 fuel assemblies have been performed on the PARAMETR bench under conditions of an unanticipated accident with reserve water poured from above. The results of materials engineering investigations of a model fuel assembly, using optical and electron microscopy, as well as x-ray microspectral and x-ray structural analysis are presented. The results of these investigations are necessary for developing and verifying methods of computing improved evaluations for the purpose of modeling accidents with substantial core damage.Ensuring safety is one of the main problems in designing, building, operating, and decommissioning nuclear power plants. Substantiating safety includes analysis of anticipated and unanticipated accidents, including accidents with serious damage to and melting of the core.An unanticipated accident is an accident caused by initial events which are not taken into account for anticipated accidents or an accident accompanied by additional, as compared with anticipated accidents, failures of safety systems above a single failure and implementation of incorrect decisions by personnel. According to the normative documents [1], the consequences of an unanticipated accident must be decreased by taking actions that are directed toward preventing the development of an anticipated accident into an unanticipated accident and toward decreasing the consequences of an unanticipated accident. According to existing ideas about limiting the development of an unanticipated accident, the core is cooled by pouring reserve water from above and below. Investigations of the behavior of the structural components of a reactor core under conditions of an unanticipated accident with reserve water poured from below have been performed for a long time (CORA and QUENCH [2] experiments), but the effect of pouring water from above on the behavior of overheated fuel has still not been studied experimentally. In this connection, the objective of the present work is to investigate the effect of cooling by pouring water from above on melting and fracture of a model VVER fuel assembly heated above 2000°C.The tests of a 19 fuel element VVER-1000 model fuel assembly with a 1275 mm long heated part and total length 3300 mm were performed on the PARAMETR bench [3]. The model assembly was made of standard structural materials. Tantalum heaters were placed inside fuel elements and isolated by UO 2 pellets from cladding made of E-110 alloy.The experiment consists of four main stages:• preparatory stage -stabilization of a prescribed argon flow rate of 2 g/sec and a steam flow rate of 3 g/sec with fuel assembly temperature 500°C and heating the fuel assembly up to 1200°C in the hottest zone; • pre-oxidation -holding a fuel assembly at 1200°C in the hottest zone for 3450 sec; • fast heating -the temperature in the hottest section of the fuel assembly is increased up to 1800-2000°C and higher; • water poured from above -water with flow rate to 40 g/sec is poured for 500 sec on the fuel assembly from ...
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