[en] A calculate methodology is presented to execute the analysis trips gives power place by reactivity inserts caused by explosion or uncontrollable extraction the nucleonic groups control the VVER-440 reactor

[en] Presently work intends to validate the models and programs used in the Nuclear Technology Center for calculating the critical position of control rods by means of the analysis of the measurements performed at the critical facility IPEN/MB-01. The lattice calculations were carried out with the WIMS/D4 code and for the global calculations the diffusion code SNAP-3D was used

[en] The present work has as objective to make a working paper where the approaches settle down Security in the design and the requirements for the sure explotation the nuclear facilities

[en] This work presents results of the research done in INR Pitesti on the new concept of SEU 43 fuel cluster designed for burnups as high as 25 Mw·day/kgU using slightly enriched uranium (up to 1.1% U235). The study of behaviour of fuel elements in accident conditions aims to find their capacity to preserve their physical integrity as well as how their failure can affect both the nuclear safety of nuclear power plant and the environment as well. By using ELOKA-MK6 and CAREB codes, the behaviour of two SEU43 fuel elements in LOCA conditions (RIH 20%, 35%, ROH 80% and 100%) was analyzed. An analysis of the main performance parameters of SEU43 fuel as compared with the results obtained for standard CANDU fuel is included. The characteristic LOCA thermohydraulic conditions were provided by FIREBIRD code and used as input data for CAREB and ELOCA. ELESIM code was used to establish the state of fuel prior to the accident. The temperature, internal pressure and total deformation of the can are plotted. The peak values of SEU43 parameters are lower than the values obtained for standard CANDU fuel. The can rupture due to excessive increase of internal pressure caused by fission gas release is one of the mechanisms characterizing the fuel integrity in LOCA type accidents. The peak total deformation of the can obtained with CAREB is -0.02% in case of 80%ROH and 0.4% in case of 20%RIH, values significantly lower then the rupture threshold. The peak pressure of fission gases in the can's interior is 6.93 MPa after 3.6 sec in case of 80%ROH, while in the 20%RIH it reaches 7.97 MPa after 3.8 sec from accident initiation. The peak temperature of the can in 20%RIH case was 1176.6 K and 1004.7 K for 80%ROH. The results obtained show the capability of SEU 43 cluster of functioning at power levels per cluster higher as compared with the standard fuel with consequences of an LOCA type accident not increased

[en] The results of reactor tests of 18-element fuel assembly with the WWER-type fuel elements in the regime characteristic of the small leakage accident are presented. The state of fuel elements after overheating during reactor tests is determined and evaluation of their efficiency is carried out

[en] The results of measuring the release of gaseous fission products and fuel temperature by irradiation of the uranium-erbium and standard fuel are presented. The results of the above measurements from ventilated fuel elements with different fuel show that the dependences of the release on temperature, decay constant and irradiation time are similar. The release of the gaseous fission products by the fuel elements irradiation increases

[en] he device for in-pile studies on the creep of fuel materials in the dispersion fuel elements at temperatures of 250-600 deg C and flux densities of thermal neutrons (2-3) x 10¹³ cm^-2s^-1 is described. It includes loading system, provision of assigned temperature, automated loading regime control and also deformation, loading and temperature measurements

[en] The problem of using burnable poisons (gadolinium and erbium oxides) integrated with fuel pellets for suppression of the excess reactivity in the LWR reactor cores at fuel cycle begin when the fuel with maximum enrichment is loaded in the core is discussed. It is shown that application of the fuel elements with such pellets ensures sufficient burnup growth for fuel with increased enrichment, increase in the fuel cycle duration and decrease in neutron fluence on reactor vessel in the cases of optimized layouts of fresh and irradiated fuel assemblies in the reactor core. Basing on the analysis of studying into (U, Gd)O₂ pellet heating and thermal conductivity under high burnups it is proved that the fuel with enrichment of 4.4 % of ²³⁵U may be used if the Gd₂O₃ content amounts to 2 %. Application of erbium absorber is recommended in uranium and plutonium fuel in inertial (nonfissible) matrix designed for burnups greater than 100 GeV · days/t

[en] This paper describes the crud induced axial power shift phenomenon, commonly known as axial offset anomaly (AOA), and the Westinghouse Nuclear Fuel methodology developed to assess the risk potential during Loading Pattern (LP) development

[en] The SIMULATE-3K (S-3K) reactor analysis code has been applied to a variety of pressurized water reactor (PWR) and boiling water reactor (BWR) transients since 1993. Over the years, many changes have occurred in the S-3K channel hydraulics and ex-core component modeling. This paper summarizes those changes and outlines the status of existing vessel and steam line models. Examples are given for BWR transients that can be analyzed with S-3K