Time filter

Source Type

Paffumi E.,European Commission | Radu V.,Institute for Nuclear Research of romenia | Nilsson K.-F.,European Commission
International Journal of Fatigue | Year: 2013

Thermal fatigue is an important degradation mechanism for the life time assessment of nuclear reactor components. A reliable life-assessment of components is difficult because usually only the nominal temperature differences are known and the thermal surface loadings are not known. This paper outlines a multi-level procedure for assessment of pipe components subjected to thermal fatigue. The different levels are: (a) simple screening criterion, (b) the thermal spectrum replaced by a sinusoidal load (SIN-method) with constant amplitude and frequency and assessment of crack initiation and crack propagation in relation to a critical frequency, and (c) spectrum loading applied to crack initiation and propagation. The different levels are described together with the underlying assumptions. The different levels in the procedure are applied to assess the life of the Civaux case, where a pipe failed due to thermal fatigue. The different levels of the procedure give conservative estimates of the thermal fatigue life but where the conservatism is reduced with the more complex higher level assessments. The influence of important factors such as boundary conditions and primary loads are illustrated. It is also shown that the SIN-method can be used to determine a threshold below which there is no thermal fatigue failure. © 2012 Elsevier Ltd. All rights reserved. Source

Tudora A.,University of Bucharest | Hambsch F.-J.,EC JRC Institute for Reference Materials and Measurements IRMM | Visan I.,University of Bucharest | Visan I.,Institute for Nuclear Research of romenia | Giubega G.,University of Bucharest
Nuclear Physics A | Year: 2015

Different methods to partition the total excitation energy (TXE) of fully accelerated fragments, presently used in prompt emission calculations include different assumptions about what is happening at scission. In fact the energy partition takes place at scission or even before scission, depending on the physical assumptions supporting the models used in different methods of TXE partition. The paper discusses two TXE partition methods in which the amount of energy to be shared (at scission and before scission, respectively) is very different. These methods (based on different principles and physical considerations) are: A. The method used in the Point-by-Point (PbP) treatment of prompt emission in which the available excitation energy at scission is shared between complementary nascent fragments. The amount of energy to be shared is sufficiently high to consider the nascent fragments in the Fermi-gas regime of the level density. B. The method used in the GEF code, in which the intrinsic energy before scission is shared between pre-nascent fragments according to the "energy sorting mechanism". This sorting mechanism is based on the assumption of level densities in the constant temperature regime, only. This is supported by the low amount of the shared intrinsic energy in the case of thermal and low energy neutron induced fission. Taking into account that the principles and physical considerations of any TXE partition method are independent on the way to treat the prompt emission (i.e. deterministically as in the PbP model or probabilistically by Monte-Carlo as in the code GEF) the methods A and B are applied to the same fission fragment range (built as in the PbP treatment). Extreme hypotheses are made for the fragment level densities on which the partitions are based (only in the Fermi-gas regime or only in the constant temperature regime). The results are compared with the energy partition obtained with fragment level densities described by the composite Gilbert-Cameron formula. Different assumptions for the deformation energies of fragments (absolute deformation at scission, or extra-deformation at scission with respect to the full acceleration) impacting the sawtooth-like shape of the excitation energy E*(A) at full acceleration are discussed, too. Limitations and advantages of these methods are also mentioned. Both TXE partition methods applied in the PbP model lead to prompt emission results (e.g. ν(A) and Eγ(A)) describing well the experimental data. © 2015 Elsevier B.V. Source

Lucan D.,Institute for Nuclear Research of romenia
Nuclear Engineering and Design | Year: 2011

Steam generators are crucial components of pressurized water reactors. The failure of the steam generator as a result of tube degradation by corrosion has been a major cause of Pressurized Heavy Water Reactor (PHWR) plant unavailability. Steam generator problems have caused major economic losses in terms of lost electricity production through forced unit outages and, in cases of extreme damage, as additional direct cost for large-scale repair or replacement of steam generators. The excellent performance to date of CANDU steam generators can be attributed, in part, to their design and performance characteristics, which typically involve higher recirculation ratios and lower heat fluxes and temperatures. However, the steam generator tubes are susceptible to failure by a variety of mechanisms, the vast majority of which are related to corrosion. The generalized corrosion is an undesirable process because it is accompanied by deposition of the corrosion products which affect the steam generator performances. It is very important to understand the generalized corrosion mechanism with the purpose of evaluating the quantities of corrosion products which exist in the steam generator after a determined period of operation (IAEA, 1997). The purpose of the experimental research consists in the assessment of corrosion behaviour of the tubes material, Incoloy-800, at normal secondary circuit parameters (temperature - 260 °C, pressure - 5.1 MPa). The testing environment was the demineralised water without impurities, at different pH values regulated with morpholine and cyclohexylamine (all volatile treatment - AVT). The results are presented like micrographics and graphics representing weight loss of metal due to corrosion, corrosion rate, total corrosion products formed, the adherent corrosion products, released corrosion products, release rate of corrosion products and release rate of the metal. © 2010 Elsevier B.V. Source

Nitoi M.,Institute for Nuclear Research of romenia
Progress in Nuclear Energy | Year: 2016

The external hazards constitute a significant source of challenges for the safe operation of a NPP. An overview of the available operating experience is presented in this paper, in order to provide a better picture about the recurrence of meteorological hazards and their impact on the safety of nuclear installations. The IAEA International Reporting System was used as a reference database in the analysis. The meteorological related events identified in the selected time window were analyzed in detail, and the contributions of each external hazard that have induced the meteorological related events, together with the lessons learnt are presented and discussed. The obtained results and the conclusions regarding the occurrence of extreme meteorological hazards, together with the distribution of recurrent events for EU and non-EU areas are highlighted. © 2016 Elsevier Ltd. All rights reserved. Source

Radu G.,Institute for Nuclear Research of romenia | Prisecaru I.,Polytechnic University of Bucharest
Nuclear Engineering and Design | Year: 2015

In case of a hypothetical severe accident in a nuclear power plant, the corium consisting of the molten reactor core and internal structures may flow onto the concrete floor of containment building. This would cause an interaction between the molten corium and the concrete (MCCI), in which the heat transfer from the hot melt to the concrete would cause the decomposition and the ablation of the concrete. The potential hazard of this interaction is the loss of integrity of the containment building and the release of fission products into the environment due to the possibility of a concrete foundation melt-through or containment over-pressurization by the gases produced from the decomposition of the concrete or by the inflammation of combustible gases. In the safety assessment of nuclear power plants, it is necessary to know the consequences of such a phenomenon. The paper presents an example of application of the ASTECv2 code to a generic CANDU6 reactor. This concerns the thermal-hydraulic behaviour of the containment during molten core-concrete interaction in the reactor vault. The calculations were carried out with the help of the MEDICIS MCCI module and the CPA containment module of ASTEC code coupled through a specific prediction-correction method, which consists in describing the heat exchanges with the vault walls and partially absorbent gases. Moreover, the heat conduction inside the vault walls is described. Two cases are presented in this paper taking into account two different heat transfer models at the pool/concrete interface and siliceous concrete. The corium pool configuration corresponds to a homogeneous configuration with a detailed description of the upper crust. © 2015 Elsevier B.V. All rights reserved. Source

Discover hidden collaborations