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Katchadjian P.,Comision Nacional de la Energia Atomica | Desimone C.,Comision Nacional de la Energia Atomica | Garcia A.,Comision Nacional de la Energia Atomica | Schroeter F.,Nucleoelectrica Argentina SA
AIP Conference Proceedings | Year: 2012

In this paper the applications, detailed in previous works, of ultrasonic transducers with the addition of axicon lenses are extended. Axicon lenses were manufactured to generate an angular refracted beam in order to study defectology in welds and other components. To achieve greater depth of focus while maintaining a relationship between focus depth and near field (F/N) less than 0.4, larger diameter transducers were used. Furthermore, its effect on the focus diameter (dF) was also analyzed. For different combinations of lens-transducer, diagrams of axial and transverse sound pressure distribution were obtained. At last, several practical applications are shown where it is possible to exploit the advantages that these transducers offer; for example: sizing of shallow cracks, high resolution corrosion mapping simulation, etc. © 2012 American Institute of Physics.

Katchadjian P.,Comision Nacional de la Energia Atomica | Desimone C.,Comision Nacional de la Energia Atomica | Garcia A.,Comision Nacional de la Energia Atomica | Antonaccio C.,Nucleoelectrica Argentina SA | And 2 more authors.
AIP Conference Proceedings | Year: 2011

The present work refers to the welding inspection of the brackets of Atucha I Nuclear Power Plant's Pressure Vessel (RPV) using the wet fluorescent magnetic particles technique (MT). Due to limited access and high radiation levels in the inspection area, it was necessary to automate the testing and use non conventional magnetization techniques. This paper describes the design and implementation of an automated inspection device and the tests carried out on the mock-up to set up the system. Also, magnetization techniques used are described, explaining in detail the non conventional technique of magnetization by current plates and the use of magnetic field concentrators to increase the field values in the area of interest. Finally, the device mounted on the RPV, used to inspect the bracket's weld, and the results achieved from the inspection are shown. © 2011 American Institute of Physics.

Dauria F.,University of Pisa | Mazzantini O.,Nucleoelectrica Argentina S.A.
Science and Technology of Nuclear Installations | Year: 2011

Within the licensing process of the KWU Atucha II PHWR (Pressurized Heavy Water Reactor), the BEPU (Best Estimate Plus Uncertainty) approach has been selected for issuing of the Chapter 15 on FSAR (Final Safety Analysis Report). The key steps of the entire process are basically two: (a) the selection of PIE (Postulated Initiating Events) and (b) the analysis by best estimate models supported by uncertainty evaluation. Otherwise, key elements of the approach are (1) availability of qualified computational tools including suitable uncertainty method, (2) demonstration of quality, and (3) acceptability and endorsement by the licensing authority. The effort of issuing Chapter 15 is terminated at the time of issuing of the present paper, and the safety margins available for the operation of the concerned NPP (Nuclear Power Plant) have been quantified. © 2011 Francesco DAuria and Oscar Mazzantini.

Adorni M.,University of Pisa | Del Nevo A.,University of Pisa | D'Auria F.,University of Pisa | Mazzantini O.,Nucleoelectrica Argentina S.A.
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2010

Depending on the specific event scenario and on the purpose of the analysis, it might be required the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes. This may imply the use of a dedicated fuel rod thermo-mechanical computer code. This paper provides an outline of the methodology for the analysis of the 2A LB-LOCA accident in Atucha-2 NPP and describes the procedure adopted for the use of the fuel rod thermomechanical code. The methodology implies the application of best estimate thermal-hydraulic, neutron physics and fuel pin performance computer codes, with the objective to verify the compliance with the specific acceptance criteria. The fuel pin performance code is applied with the main objective to evaluate the extent of cladding failures during the transient. The procedure consists of a deterministic calculation by the fuel performance code of each individual fuel rod during its lifetime and in the subsequent LB-LOCA transient calculations. The boundary and initial conditions (e.g. pin power axial profiles) are provided by core physics and three dimensional neutron kinetic coupled thermal-hydraulic system codes (RELAP5-3D©) calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, with the aim of a better understanding of the uncertainties involved and their technological consequences on the behavior of the fuel rods, not addressed in the current paper. Copyright © 2010 by ASME.

Idiart M.I.,National University of La Plata | Idiart M.I.,CONICET | Ramos Nervi J.E.,National University of La Plata | Ramos Nervi J.E.,Nucleoelectrica Argentina S.A.
Comptes Rendus - Mecanique | Year: 2014

A linear-comparison homogenization technique and its relaxed version are used to compute bounds of the Hashin-Shtrikman and the self-consistent types for the hydrostatic strength of ideally plastic voided polycrystals. Closed-form analytical results are derived for isotropic aggregates of various cubic symmetries (fcc, bcc, ionic). The impact of the variational relaxation on the bounds is found to be significantly larger than that previously observed in fully dense polycrystals. So much so that, quite surprisingly, relaxed self-consistent bounds are found to be weaker than non-relaxed Hashin-Shtrikman bounds in some of the material systems considered. © 2013 Académie des sciences.

Nervi J.E.R.,National University of La Plata | Nervi J.E.R.,Nucleoelectrica Argentina S.A. | Idiart M.I.,National University of La Plata | Idiart M.I.,CONICET
Proceedings of the Royal Society A: Mathematical, Physical and Engineering Sciences | Year: 2015

The elastoplastic response of polycrystalline voided solids is idealized here as rigid-perfectly plastic. Bounds on the macroscopic plastic strength for prescribed microstructural statistics and single-crystal strength are computed be means of a linearcomparison homogenization technique developed by Idiart&Ponte Castañeda (2007Proc. R. Soc. A 463, 907-924. (doi:10.1098/rspa.2006.1797)). Hashin-Shtrikman (HS) and Self-Consistent (SC) results in the form of yield surfaces are reported for cubic and hexagonal polycrystals with isotropic texture and varying degrees of crystal anisotropy. In all cases, the surfaces are smooth, closed and convex. Improvements over earlier linear-comparison bounds of up to 40% are found at high-stress triaxialities. New HS results can even be sharper than earlier SC results for some material systems. In the case of deficient crystals, the SC results assert that voided aggregates of crystals with four independent systems can accommodate arbitrary deformations, those with three independent systems can dilate but not distort, and those with fewer than three independent systems cannot deform at all. We report the sharpest bounds available to date for all classes of material systems considered. © 2015 The Author(s) Published by the Royal Society. All rights reserved.

Zhang T.,Engineering Mechanics Corporation Of Columbus | Brust F.W.,Engineering Mechanics Corporation Of Columbus | Wilkowski G.,Engineering Mechanics Corporation Of Columbus | Xu H.,Engineering Mechanics Corporation Of Columbus | And 2 more authors.
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2012

The Atucha II nuclear power plant is a pressurized heavy water reactor being constructed in Argentina. Nuclear power plants must be designed to maintain their integrity and performance of safety functions for a bounding set of normal operational events as well as abnormal events that might occur during the lifetime of the plant. Seismic fracture mechanics evaluations for the Atucha II plant showed that even with a seismic event with the amplitudes corresponding to an event with a probability of 10-6 per year, that a double-ended guillotine break (DEGB) was pragmatically impossible due to the incredibly high leakage rates and total loss of make-up water inventory. The critical circumferential through-wall flaw size for this case is 94-percent of the circumference. These analyses are performed by placing cracked-pipe-elements into a complete model of the primary cooling system including the reactor pressure vessel, pumps, and steam generators as summarized in the paper. This paper summarizes these results and further shows how much higher the applied accelerations would have to be to cause a DEGB for an initial circumferential through-wall crack that was 33 percent (about 120°) around the circumference. This flaw length would also be easily detected by leakage and loss of make-up water inventory. These analyses showed that the applied seismic peak-ground accelerations had to exceed 25 g's for the case of this through-wall-crack to become a DEGB during a single seismic loading event. This is a factor of 80 times higher than the 10-6 seismic event accelerations, or 240 times higher than the SSE accelerations. This suggests there is a huge safety margin for beyond design basis seismic events and Atucha II plant rupture is pragmatically impossible. These surprising results are discussed and could be potentially applicable to other nuclear power plants as well. Copyright © 2012 by ASME.

Uddin M.,Engineering Mechanics Corporation Of Columbus | Brust F.,Engineering Mechanics Corporation Of Columbus | Wilkowski G.,Engineering Mechanics Corporation Of Columbus | Zhang T.,Caterpillar Inc. | And 3 more authors.
American Society of Mechanical Engineers, Pressure Vessels and Piping Division (Publication) PVP | Year: 2014

The Transition Break Size analysis is given in Draft Regulatory Guide 1216 for assessment of ECCS requirements. There are additional analyses required for seismic considerations for beyond design basis seismic loading. The beyond design basis loading was for seismic event with a probability of occurrence of 1e-6 per year, whereas safe-shutdown earthquake loading is typically closer to a probability of 1e-4 events per year. The peak-ground accelerations for US plants are typically in the 0.1 to 0.2 g's range for SSE loading, while the 1e-6 seismic loading may be about 3 times higher (depending on site specific seismic hazard curves). A simplified method was created for the TBS seismic consideration analysis, which is given in Appendix A and B in the Draft Reg Guide. The technical basis is in NUREG-1903. The TBS analysis approach utilizes a simple method for scaling the seismic stresses from the SSE to 1e-6 loading, allowing for a bilinear stress correction (linear to yield of the material and an ultimate point consistent with uncracked seismic tested pipe results). A best-estimate fracture analysis is then conducted using the ASME Section XI Service Level D flaw size, but a best-estimate fracture analysis uses more realistic material properties and more accurate fracture analyses than the ASME Code. In this paper, the TBS flaw size was calculated by the Draft Regulatory Guide approach for the Atucha II nuclear plant in Argentina that is about to start up. Additionally a full 3D FE model of the plant including the whole NSSS, containment building, and supports between the building and the NSSS components was developed. Circumferential surface cracks were put in the nonlinear time-history FE model at the highest stressed locations in the primary pipe loop to determine the depth of the flaw that would fail at the 1e-6 seismic excitation to the plant building. This was done to assess the margins in the flaw size for the TBS analysis, and also characterize the magnitude of the LOCA. Where the TBS simplified model showed that a surface flaw of 0.709 and 0.669 of the thickness for the RPV/hot-leg and pump/cold-leg (respectively) and 270-degrees around the circumference could be tolerated, the full 3D FE analysis showed that even a surface crack of 90-percent of the thickness and 270-degrees around the circumference would not reach crack initiation for the material used in this plant with its seismic hazard curve. The information developed here may also be useful for assessing the piping integrity of a plant once it has exceeded the Service Level D limits of the ASME Code. Copyright © 2014 by ASME.

Gaute F.,Nucleoelectrica Argentina S.A.
International Journal of Low Radiation | Year: 2011

The main purpose of this paper is to explain the Radiological Assessment Programme implemented by Nucleoeléctrica Argentina SA (NASA) in the Atucha I NPP and Atucha II NPP sites. Atucha I NPP reached its first criticality in 1974, while Atucha II NPP will start its commercial operation in the last quarter of 2011. The Radiological Assessment Programme verifies the compliance of the basic safety standards with reference to the radiation exposure of the population in the surrounding area and the environmental impact searching for continuous improvement, as a result of applying the ALARA philosophy. The reference levels are calculated considering a theoretical dose of 0.05 mSv/year for the critical group. This value is considered for the discharge limits for Atucha I NPP (Ki) in the Operations Licence in force, and those limits comprise all the radionuclides established in the Argentine Regulatory Authority Requirements. Copyright © 2011 Inderscience Enterprises Ltd.

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