Institute for Nuclear Research of romenia
Institute for Nuclear Research of romenia
Glinatsis G.,ENEA |
Carta M.,ENEA |
Gugiu D.,Institute for Nuclear Research of romenia
Annals of Nuclear Energy | Year: 2017
The target of this study is to provide information on the actual boundary conditions encountered in the conceptual design of an ADS fuelled by (Th, LEU), as well as their impact on the design itself, on the performances of the system and on the possible solutions from the point of view of the choice of materials, the operating conditions, the neutron design optimization, etc. In this paper rationales and consequences of the (Th, U) fuel cycle adoption, together with the design constraints, are discussed. Identified reference core configuration and the steady state criticality results are also discussed. Stochastic approach for the neutronic evaluations has been adopted. © 2017 Elsevier Ltd
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.
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.
Radu V.,Institute for Nuclear Research of romenia |
Roth M.,Institute for Nuclear Research of romenia
Nuclear Engineering and Design | Year: 2012
For CANDU pressure tubes made from Zr-2.5%Nb alloy, the mechanism called delayed hydride cracking (DHC) is widely recognized as main mechanism responsible for crack initiation and propagation in the pipe wall. Generation of some blunt flaws at the inner pressure tube surface during refueling by fuel bundle bearing pad or by debris fretting, combined with hydrogen/deuterium up-take (20-40 ppm) from normal corrosion process with coolant, may lead to crack initiation and growth. The process is governed by hydrogen hysteresis of terminal solid solubility limits in Zirconium and the diffusion of hydrogen atoms in the stress gradient near to a stress spot (flaw). Creep and irradiation growth under normal operating conditions promote the specific mechanisms for Zirconium alloys, which result in circumferential expansion, accompanied by wall thinning and length increasing. These complicate damage mechanisms in the case of CANDU pressure tubes that are also are affected by irradiation environment in the reactor core. The structural integrity assessment of CANDU fuel channels is based on the technical requirements and methodology stated in the Canadian Standard N285.8. Usually it works with fracture mechanics principles in a deterministic manner. However, there are inherent uncertainties from the in-service inspection, which are associated with those from material properties determination; therefore a necessary conservatism in deterministic evaluation should be used. Probabilistic approach, based on fracture mechanics principle and appropriate limit state functions defined as fracture criteria, appears as a promising complementary way to evaluate structural integrity of CANDU pressure tubes. To perform this, one has to account for the uncertainties that are associated with the main parameters for pressure tube assessment, such as: flaws distribution and sizing, initial hydrogen concentration, fracture toughness, DHC rate and dimensional changes induced by long term irradiation in reactor. The paper describes a prospective way for the probabilistic approach of CANDU pressure tube failure by DHC mechanisms during cool-down cycles by using probabilistic fracture mechanics principles. The limit state functions are defined for fracture instability and plastic collapse according to the Canadian Standard N285.8-05 criteria. British Procedure R6 is used to define another limit state function based on reserve factors. Discussion of lifetime probability values obtained from both procedures is made. © 2012 Elsevier B.V. All rights reserved.
Farcasiu M.,Institute for Nuclear Research of romenia |
Prisecaru I.,Polytechnic University of Bucharest
Reliability Engineering and System Safety | Year: 2014
The results of many Probabilistic Safety Assessment (PSA) studies show a very significant contribution of human errors to nuclear installations failure. This paper is intended to analyze both the human performance importance in PSA studies and the elements that influence it. Starting from Man-Machine- Organization System (MMOS) concept a new approach (MMOSA) was developed to allow an explicit incorporation of the human and organizational factor in PSA studies. This method uses old techniques from Human Reliability Analysis (HRA) methods (THERP, SPAR-H) and new techniques to analyze human performance. The main novelty included in MMOSA is the identification of the machine-organization interfaces (maintenance, modification and aging management plan and state of man-machine interface) and the human performance evaluation based on them. A detailed result of the Human Performance Analysis (HPA) using the MMOSA methodology can identify any serious deficiencies of human performance which can usually be corrected through the improvement of the related MMOS interfaces. © 2013 Elsevier Ltd.
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.
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.
Dinca M.,Institute for Nuclear Research of romenia |
Mandescu D.,Arges County Museum
Physics Procedia | Year: 2015
The neutron and gamma imaging facility placed at the tangential channel of the TRIGA-ACPR from INR was used for tomography investigations on a test object with good results and shortly followed its involvement for tomography investigations on prehistoric statues of clay from the Arges County Museum. This activity was performed in connection with a research contract with IAEA with title "The neutron and gamma imaging method combined with neutron-based analytical methods for cultural heritage research", in the frame of a current CRP, that helps curators to reveal the internal structure and composition of the objects. The detector system has been developed based on two interchangeable scintillators, one for thermal neutrons and the other one for gamma radiations, a mirror of float glass coated with aluminum and two interchangeable CCD cameras. Experiments of tomography imaging for two prehistoric statues of clay with CCD STARLIGHT XPRESS SXV-H9 camera with XD-4 type image intensifier are presented in this paper. The tomography reconstructions with Octopus software have shown the potential of good results even for 100 projections/1800. This was a good opportunity for the dissemination of the investigation methods based on neutrons for cultural heritage and beyond this area. © 2015 The Authors.
Constantin A.,Institute for Nuclear Research of romenia |
Constantin M.,Institute for Nuclear Research of romenia |
Diaconu D.,Institute for Nuclear Research of romenia
Mineralogical Magazine | Year: 2015
Many countries encourage national forums for transparency, dialogue and participation with regards to radioactive waste disposal. However, the local actors (authorities, non-government organisations and the public) often note a lack of public participation in the decision making process. Civil society is often frustrated with its limited involvement in the consultative process. Participation is regulated by national laws and rules and the right to participate in environmental decision-making is covered by the Aarhus Convention. Continuous dialogue amongst stakeholders is seen as important in building sustainable solutions in radioactive waste management. In addition, understanding public concerns and needs can increase the trust between the partners and build confidence in the process. Different national and local contexts have contributed to the development of quite a broad set of approaches and tools for stakeholder engagement. This paper describes the use of such tools in the engagement with the Saligny community in the siting process of a repository for low- and intermediate-level wastes in Romania. Some specific issues are highlighted such as: the low level of interest amongst the public in relation to long-term projects; over-estimation of benefits in comparison to the negative aspects of hosting a repository; lack of a coherent public voice; and a perceived lack of information on the project from the authorities and the implementer. The present study describes the setting up of the participatory approach to engage with the public and the different methods employed (including citizen juries, workshops, open days, etc.). A number of criteria were developed for evaluating the effectiveness of these methods particularly with regards to their adaptability to a local context such as Saligny. The paper then focuses on the results of one of these methods - the use of focus groups covering a cross-section of civil society - including members of the general public, a group of professionals and a group of local councillors. The study has resulted in a number of recommendations to the implementer on how to build a new programme for public participation. © 2016 by Walter de Gruyter Berlin/Boston.
Dianu M.,Institute for Nuclear Research of romenia
Fusion Engineering and Design | Year: 2012
The aim of the present study is to investigate a method to evaluate the tritium activity in hydraulic oil waste generated during the operation of Romanian Cernavoda Nuclear Power Plant. The method is based on a combustion technique using the 307 PerkinElmer ® Sample Oxidizer model. The hydraulic oil samples must be processed prior to counting to avoid color quenching (the largest source of inaccuracy) because these samples absorb in the region of 200-500 nm, where scintillation phosphors emit. Prior to combustion of the hydraulic oil waste, tritium recovery degree and tritium retention degree in the circuits of combustion system were evaluated as higher than 98% and less than 0.08%, respectively. After combustion, tritium activity was measured by a 2100 Tri-Carb ® Packard model liquid scintillation analyzer. The blank counts were 16.25 ± 0.50 counts/min, measured for 60 min. The significant activity level value was 6.53 counts/min, at a preselected confidence level of 95%. The Minimum Detectable Activity of a 0.2 mL hydraulic oil sample was calculated to 1.09 Bq/mL. Therefore, the developed method is sensitive enough for the tritium evaluation in the ordinary hydraulic oil waste samples. © 2012 Elsevier B.V.