[en] The 2011 International Congress on Advances in Nuclear Power Plants depicts the recent evolution in reactor physics, thermal-hydraulics, materials, operation and maintenance, safety and licensing of NPP. The fuel cycle and waste management strategies are intimately part of the overall picture raising near term deployment stakes such as environment, public acceptance and non-proliferation issues. Also, the long term sustainability challenges are emphasized through the many alternative designs including innovative fast reactors concepts and dedicated non-electrical systems. The papers are organized into 10 issues: -) water-cooled reactor programs, -) high temperature gas-cooled reactors, -) LMFR and innovative reactor programs, -) operation, performance and reliability management, -) plant safety assessment and regulatory issues, -) reactor physics and analysis, -) thermal hydraulics analysis and testing, -) fuel cycle and waste management, -) materials and structural issues, -) nuclear energy and global environment, -) deployment and cross-cutting issues, -) plant licensing and international regulatory issues, and -) 5. European nuclear education network PhD event

[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] Both in classical and nuclear power plants the corrosion processes taking place at the ferrous alloys/aqueous medium interfaces result in coating the tubes with iron oxides or oxi-hydroxides. Some of these layers with a given thickness or chemical composition such as magnetite may have a protecting role for the basis material, while in other cases these oxides or hydroxides have a harmful effect on thermal transfer which proceeds through tubes and consequently the energy demand increases to compensate heat spurious losses. The solutions mainly acid used for descaling these depositions must fulfil the following requirements: - solubilization of the coating film should be as complete as possible; - the base material must remain unaffected; - chemical residues resulting from descaling should not cause further corrosion if they were not completely removed; - these solutions must be easy to handle, neither explosive or toxic. For the case of CANDU NPPs the descaling solutions contain one or a mixture of organic acids as for instance lactic, oxalic, citric acids, a complexing agent such as H-EDTA, Na₄EDTA, etc, and one or more inhibitors to protect the base material against the solutions' a tack. Diverse inhibitors or inhibitors of different concentrations were compared by means of the cathodic portions of the potentio-dynamic curves. The inhibitors ordered by decreasing efficiency, as found in this work, appear to be: trietilamina, etilen diamine, triethanolamine, N-phenylthiourea and thiourea

[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