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Sample records for nuclear atucha ii

  1. Atucha II nuclear power plant digital simulation

    International Nuclear Information System (INIS)

    Santome, D.; Rovere, L.A.T.

    1987-01-01

    This paper describes the start-up of a digital simulation code apt to be performed in real time of Atucha II nuclear power plant, foreseeing its subsequent usage in a Basic Principles Simulator. Adaptability and modification of existing routines and development of modules in order to incorporate the necessary variables dynamics to couple the different modes, were the main tasks. The mathematical model used allows the representation of the following sub-systems: a) a reactor's core point model, which comprehends the neutronic kinetics, fission and decaying powers, thermal transfer and Xe-poisoning calculation; b) pressurizer, which considers two sub-systems that may or may not be in thermodynamic equilibrium, both in two phases; c) coolants and moderators bonds considering separate moderator loops with the aim of introducing asymmetric perturbations; d) secondary sub-subsystem, which includes the feed water loop, pumps, steam generators and control valves; e) steam generators; f) control and safety systems, including power control, steam generators levels, moderator's temperature primary loop system, limitations and protection. (Author)

  2. Modeling of the core of Atucha II nuclear power plant

    International Nuclear Information System (INIS)

    Blanco, Anibal

    2007-01-01

    This work is part of a Nuclear Engineer degree thesis of the Instituto Balseiro and it is carried out under the development of an Argentinean Nuclear Power Plant Simulator. To obtain the best representation of the reactor physical behavior using the state of the art tools this Simulator should couple a 3D neutronics core calculation code with a thermal-hydraulics system code. Focused in the neutronic nature of this job, using PARCS, we modeled and performed calculations of the nuclear power plant Atucha 2 core. Whenever it is possible, we compare our results against results obtained with PUMA (the official core code for Atucha 2). (author) [es

  3. Reactor building design of nuclear power plant ATUCHA II, Argentina

    International Nuclear Information System (INIS)

    Rufino, R.E.; Hermann, E.R.; Richter, E.

    1984-01-01

    It is presented the civil engineering project carried out by the joint venture Hochtief - Techint-Bignoli (HTB) for the reactor building at the Atucha II power plant (PHWR of 745 MWe) in Buenos Aires. All the other civil projects at Atucha II are also being carried out by HTB. This building has the same general characteristics of the PWR plants developed by KWU in Germany, known for the spherical steel containment 56m in diameter. Nevertheless, it differs from those principally in the equipment lay-out and the remarkable foundation depth. From the basic engineering provided by ENACE, the joint venture has had to face the challenge of designing a tridimensional structure of large size. This has necessitated using simplified models which had to be superimposed, since the use of only one spatial mode would be highly inadequate, lacking the flexibility necessary to absorb the numerous modifications that this type of project undergoes during construction. In addition, this procedure has eliminated resorting to numerous and costly computer processings. (Author) [pt

  4. Concept and structure of instrumentation and control of the Atucha II nuclear power plant

    International Nuclear Information System (INIS)

    Garzon, D.; Roca, J.L.

    1987-01-01

    The general structure of instrumentation and control of Atucha II nuclear power plant as well as the technologies used, are described: concepts of functional decentralization and physical centralization; concept of functional group and functional complex; description of the technologies used (physical support) in the project of plant instrumentation and control; description of the different automation levels on the basis of concepts of control interface, automatism, regulation, group and subgroup controls; principles of signal conditioning; concept of announcement of alarms and state: supervisory computer, description of HAS (Hard wired Alarm System) and CAS (Computer Alarm System); application of the above mentioned structure to the project of another type of plants. (Author)

  5. The Argentine nuclear policy: evaluation and proposals of the National Atomic Energy Commission. Privatization of the nuclear power plants and the situation of Atucha II. Addendum 1

    International Nuclear Information System (INIS)

    2001-01-01

    An analysis is made of the possible privatization of the nuclear power plants in Argentina. The urgent definition of a nuclear policy for the short, medium and long term is strongly recommended to the Government together with the conclusion of the construction of the Atucha II nuclear power plant

  6. The world's reactors No. 82: Atucha II

    International Nuclear Information System (INIS)

    Anon.

    1982-01-01

    Detailed information on Atucha II, a 745 MWe pressure-vessel PHWR, is presented in the form of a wall chart. Plant specifications and full colour cutaway drawings of the power station are included. (U.K.)

  7. Model of automatic fuel management for the Atucha II nuclear central with the PUMA IV code

    International Nuclear Information System (INIS)

    Marconi G, J.F.; Tarazaga, A.E.; Romero, L.D.

    2007-01-01

    The Atucha II central is a heavy water power station and natural uranium. For this reason and due to the first floor reactivity excess that have this type of reactors, it is necessary to carry out a continuous fuel management and with the central in power (for the case of Atucha II every 0.7 days approximately). To maintain in operation these centrals and to achieve a good fuels economy, different types of negotiate of fuels that include areas and roads where the fuels displace inside the core are proved; it is necessary to prove the great majority of these managements in long periods in order to corroborate the behavior of the power station and the burnt of extraction of the fuel elements. To carry out this work it is of great help that a program implements the approaches to continue in each replacement, using the roads and areas of each administration type to prove, and this way to obtain as results the one regulations execution in the time and the average burnt of extraction of the fuel elements, being fundamental this last data for the operator company of the power station. To carry out the previous work it is necessary that a physicist with experience in fuel management proves each one of the possible managements, even those that quickly can be discarded if its don't fulfill with the regulatory standards or its possess an average extraction burnt too much low. For this it is of fundamental help that with an automatic model the different administrations are proven and lastly the physicist analyzes the more important cases. The pattern in question not only allows to program different types of roads and areas of fuel management, but rather it also foresees the possibility to disable some of the approaches. (Author)

  8. Hydrogeological modelling of the eastern region of Areco river locally detailed on Atucha I and II nuclear power plants area

    International Nuclear Information System (INIS)

    Grattone, Natalia I.; Fuentes, Nestor O.

    2009-01-01

    Water flow behaviour of Pampeano aquifer was modeled using Visual Mod-flow software Package 2.8.1 with the assumption of a free aquifer, within the region of the Areco river and extending to the rivers of 'Canada Honda' and 'de la Cruz'. Steady state regime was simulated and grid refinement allows obtaining locally detailed calculation in the area of Atucha I and II Nuclear power plants, in order to compute unsteady situations as the consequence of water flow variations from and to the aquifer, enabling the model to study the movement of possible contaminant particles in the hydrogeologic system. In this work the effects of rivers action, the recharge conditions and the flow lines are analyzed, taking always into account the range of reliability of obtained results, considering the incidence of uncertainties introduced by data input system, the estimates and interpolation of parameters used. (author)

  9. Experience gathered from the transport of a fuel element prototype of the CNA-II (Atucha-II nuclear power plant) type

    International Nuclear Information System (INIS)

    Pastorini, A.; Belinco, C.G.; El Bis, E.D.; Sacchi, M.A.; Mayans, C.O.; Martin Ghiselli, A.; Marcora, G.R.

    1990-01-01

    This work describes the needs to materialize the transport of a fuel element prototype of the CNA-II (Atucha-II nuclear power plant) type, under special conditions, from the Fabrication Pilot Plant sited at the Constituyentes Atomic Center and the Ezeiza Atomic Center, for its subsequent analysis at the High Pressure Experimental Loop. The special conditions under which the transport has been made responded to the fact that the prototype presents a fragile adjustment between rods and separators, necessary to be preserved. (Author) [es

  10. Ageing management program of wires for Atucha II Nuclear Power Plant

    International Nuclear Information System (INIS)

    Zorrilla, J.; Antonaccio, E.; Luraschi, C.; Simionato, E.; Nieto, H.

    2012-01-01

    Electrical cables constitute one of the most important components of NPP in terms o maintenance, safety and availability of the plant. Due to their large extension (thousands of kilometers) it is impossible to fully replace them and aging management becomes essential for long term operation of NPP. Since Atucha II is under construction there was a good opportunity to establish and implement a holistic cable aging management program from the beginning according to the state of the art in this field. The scope of this program involves safety related cables, including EQ cables and non EQ cables as well. Due to the diversity of the installed cables is impossible to address an aging management program of every single specimen. However it is possible to establish 'cables families' of similar aging behavior based on insulation and jacket material, manufacture, etc Several aging management strategies were set for the different 'cables families'. These strategies include cables deposit in plant, elaboration of procedures of visual and tactile inspection, NDE techniques, etc Currently the aging management program is being implemented covering topic such as: Cable screening and grouping. Review of EQ documentation and conventional qualification information referring to the installed cables. Establishment of base line of condition monitoring techniques. Calibration and set up of test parameters for NDE techniques. Aging mechanism characterization and determination of. Design construction and installation of cable deposit in plant (author)

  11. Atucha I nuclear power plant surveillance programme

    Energy Technology Data Exchange (ETDEWEB)

    Jinchuk, D [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

    1994-12-31

    After a review of the main characteristics of the Atucha I nuclear power plant and its pressure vessel, the embrittlement surveillance capsules and the irradiation conditions are described; Charpy impact tests and tensile tests were performed on the irradiated samples, and results are discussed and compared to theoretical calculations: transition temperature shifts, displacement per atom values. 6 refs., 16 figs., 7 tabs.

  12. Atucha I nuclear power plant surveillance programme

    International Nuclear Information System (INIS)

    Jinchuk, D.

    1993-01-01

    After a review of the main characteristics of the Atucha I nuclear power plant and its pressure vessel, the embrittlement surveillance capsules and the irradiation conditions are described; Charpy impact tests and tensile tests were performed on the irradiated samples, and results are discussed and compared to theoretical calculations: transition temperature shifts, displacement per atom values. 6 refs., 16 figs., 7 tabs

  13. Atucha I nuclear power plant transients analysis

    International Nuclear Information System (INIS)

    Castano, J.; Schivo, M.

    1987-01-01

    A program for the transients simulation thermohydraulic calculation without loss of coolant (KWU-ENACE development) to evaluate Atucha I nuclear power plant behaviour is used. The program includes systems simulation and nuclear power plants control bonds with real parameters. The calculation results show a good agreement with the output 'protocol' of various transients of the nuclear power plant, keeping the error, in general, lesser than ± 10% from the variation of the nuclear power plant's state variables. (Author)

  14. The pressure and leak tests in Atucha II

    International Nuclear Information System (INIS)

    Anon.

    1991-01-01

    This work deals with the pressure and leak tests of the containment sphere in the Atucha II Nuclear Power Plant's reactor building. This sphere is a metallic container, made in highly resistant steel plate, that is, built for providing the plant with a biological and structural barrier, which -in turn- provides safety and environmental protection. The applicable rules for these tests establish that the containment erection must be complete and in equivalent conditions to those that will prevail during the NPP operation. Particularly, pressure tests were carried out for assessing the structural condition of the sphere, while the leak test is aimed at the detection of tentative leaks [es

  15. Reliability analysis for Atucha II reactor protection system signals

    International Nuclear Information System (INIS)

    Roca, Jose Luis

    1996-01-01

    Atucha II is a 745 MW Argentine Power Nuclear Reactor constructed by ENACE SA, Nuclear Argentine Company for Electrical Power Generation and SIEMENS AG KWU, Erlangen, Germany. A preliminary modular logic analysis of RPS (Reactor Protection System) signals was performed by means of the well known Swedish professional risk and reliability software named Risk-Spectrum taking as a basis a reference signal coded as JR17ER003 which command the two moderator loops valves. From the reliability and behavior knowledge for this reference signal follows an estimation of the reliability for the other 97 RPS signals. Because the preliminary character of this analysis Main Important Measures are not performed at this stage. Reliability is by the statistic value named unavailability predicted. The scope of this analysis is restricted from the measurement elements to the RPS buffer outputs. In the present context only one redundancy is analyzed so in the Instrumentation and Control area there no CCF (Common Cause Failures) present for signals. Finally those unavailability values could be introduced in the failure domain for the posterior complete Atucha II reliability analysis which includes all mechanical and electromechanical features. Also an estimation of the spurious frequency of RPS signals defined as faulty by no trip is performed

  16. Reliability analysis for Atucha II reactor protection system signals

    International Nuclear Information System (INIS)

    Roca, Jose L.

    2000-01-01

    Atucha II is a 745 MW Argentine power nuclear reactor constructed by Nuclear Argentine Company for Electric Power Generation S.A. (ENACE S.A.) and SIEMENS AG KWU, Erlangen, Germany. A preliminary modular logic analysis of RPS (Reactor Protection System) signals was performed by means of the well known Swedish professional risk and reliability software named Risk-Spectrum taking as a basis a reference signal coded as JR17ER003 which command the two moderator loops valves. From the reliability and behavior knowledge for this reference signal follows an estimation of the reliability for the other 97 RPS signals. Because the preliminary character of this analysis Main Important Measures are not performed at this stage. Reliability is by the statistic value named unavailability predicted. The scope of this analysis is restricted from the measurement elements to the RPS buffer outputs. In the present context only one redundancy is analyzed so in the Instrumentation and Control area there no CCF (Common Cause Failures) present for signals. Finally those unavailability values could be introduced in the failure domain for the posterior complete Atucha II reliability analysis which includes all mechanical and electromechanical features. Also an estimation of the spurious frequency of RPS signals defined as faulty by no trip is performed. (author)

  17. A Coupled Calculation Suite for Atucha II Operational Transients Analysis

    International Nuclear Information System (INIS)

    Mazzantini, O.; Schivo, M.; Cesare, J.D.; Garbero, R.; Rivero, M.; Theler, G.

    2011-01-01

    While more than a decade ago reactor and thermal hydraulic calculations were tedious and often needed a lot of approximations and simplifications that forced the designers to take a very conservative approach, computational resources available nowadays allow engineers to cope with increasingly complex problems in a reasonable time. The use of best-estimate calculations provides tools to justify convenient engineering margins, reduces costs, and maximises economic benefits. In this direction, a suite of coupled best-estimate specific calculation codes was developed to analyse the behaviour of the Atucha II nuclear power plant in Argentina. The developed tool includes three-dimensional spatial neutron kinetics, a channel-level model of the core thermal hydraulics with subcooled boiling correlations, a one-dimensional model of the primary and secondary circuits including pumps, steam generators, heat exchangers, and the turbine with all their associated control loops, and a complete simulation of the reactor control, limitation, and protection system working in closed-loop conditions as a faithful representation of the real power plant. In the present paper, a description of the coupling scheme between the codes involved is given, and some examples of their application to Atucha II are shown

  18. Atucha II NPP full scope simulator modelling with the thermal hydraulic code TRACRT

    International Nuclear Information System (INIS)

    Alonso, Pablo Rey; Ruiz, Jose Antonio; Rivero, Norberto

    2011-01-01

    In February 2010 NA-SA (Nucleoelectrica Argentina S.A.) awarded Tecnatom the Atucha II full scope simulator project. NA-SA is a public company owner of the Argentinean nuclear power plants. Atucha II is due to enter in operation shortly. Atucha II NPP is a PHWR type plant cooled by the water of the Parana River and has the same design as the Atucha I unit, doubling its power capacity. Atucha II will produce 745 MWe utilizing heavy water as coolant and moderator, and natural uranium as fuel. A plant singular feature is the permanent core refueling. TRAC R T is the first real time thermal hydraulic six-equations code used in the training simulation industry for NSSS modeling. It is the result from adapting to real time the best estimate code TRACG. TRAC R T is based on first principle conservation equations for mass, energy and momentum for liquid and steam phases, with two phase flows under non homogeneous and non equilibrium conditions. At present, it has been successfully implemented in twelve full scope replica simulators in different training centers throughout the world. To ease the modeling task, TRAC R T includes a graphical pre-processing tool designed to optimize this process and alleviate the burden of entering alpha numerical data in an input file. (author)

  19. Integrated management for aging of Atucha Nuclear Power Plant

    International Nuclear Information System (INIS)

    Ranalli, J.M.; Marchena, M.H.; Sabransky, M.; Fonseca, M.; Santich, J.; Pedernera, P.

    2012-01-01

    Atucha NPP is a two PHWR unit site located in Lima, Province of Buenos Aires, 120 km north of Buenos Aires, Argentina.. With the start-up of Atucha II and aiming to integrate the Ageing Management of the plants, the Utility (Nucleolectrica Argentina Sociedad Anonima - NASA) created an Ageing Management Department to cope with all ageing issues of both Atucha I and II. In this project both organization has formed a joint working group. The role of CNEA is providing technical support to the plant in the development of procedures a methodological framework for the Ageing Management Program of Atucha NPP. The main documents that have being issued so far are: . An Ageing Management Manual, including standard definition of Materials, Ageing Related Degradation Mechanisms, Operation Environments customized for Atucha NPP. . Walk down procedures and checklists aimed to systematize data collection during outages. . Procedures for performing Ageing Management Reviews and Maintenance Reviews for passive and active components. . Condition Assessments of several safety related systems. . Condition assessment of electrical components. In the present work a summary of the activities, documental structure and first outputs of the Integrated Ageing Management Program of Atucha NPP is presented (author)

  20. Conceptual design of an interim dry storage system for the Atucha nuclear power plant spent fuels

    International Nuclear Information System (INIS)

    Nassini, Horacio E.P.; Fuenzalida Troyano, C.S.; Bevilacqua, Arturo M.; Bergallo, Juan E.

    2005-01-01

    The Atucha I nuclear power station, after completing the rearrangement and consolidation of the spent fuels in the two existing interim wet storage pools, will have enough room for the storage of spent fuel from the operation of the reactor till December 2014. If the operation is extended beyond 2014, or if the reactor is decommissioned, it will be necessary to empty both pools and to transfer the spent fuels to a dry storage facility. This paper shows the progress achieved in the conceptual design of a dry storage system for Atucha I spent fuels, which also has to be adequate, without modifications, for the storage of fuels from the second unity of the nuclear power station, Atucha II, that is now under construction. (author) [es

  1. Atucha II NPP full scope simulator modelling with the thermal hydraulic code TRAC{sub R}T

    Energy Technology Data Exchange (ETDEWEB)

    Alonso, Pablo Rey; Ruiz, Jose Antonio; Rivero, Norberto, E-mail: prey@tecnatom.e, E-mail: jaruiz@tecnatom.e, E-mail: nrivero@tecnatom.e [Tecnatom S.A., Madrid (Spain)

    2011-07-01

    In February 2010 NA-SA (Nucleoelectrica Argentina S.A.) awarded Tecnatom the Atucha II full scope simulator project. NA-SA is a public company owner of the Argentinean nuclear power plants. Atucha II is due to enter in operation shortly. Atucha II NPP is a PHWR type plant cooled by the water of the Parana River and has the same design as the Atucha I unit, doubling its power capacity. Atucha II will produce 745 MWe utilizing heavy water as coolant and moderator, and natural uranium as fuel. A plant singular feature is the permanent core refueling. TRAC{sub R}T is the first real time thermal hydraulic six-equations code used in the training simulation industry for NSSS modeling. It is the result from adapting to real time the best estimate code TRACG. TRAC{sub R}T is based on first principle conservation equations for mass, energy and momentum for liquid and steam phases, with two phase flows under non homogeneous and non equilibrium conditions. At present, it has been successfully implemented in twelve full scope replica simulators in different training centers throughout the world. To ease the modeling task, TRAC{sub R}T includes a graphical pre-processing tool designed to optimize this process and alleviate the burden of entering alpha numerical data in an input file. (author)

  2. Upgrading Atucha 1 nuclear power plant. Regulatory perspective

    International Nuclear Information System (INIS)

    Caruso, G.

    1998-01-01

    Atucha 1 nuclear power plant has unique design and its commercial operation started in 1974. The upgrading decisions, the basis for an upgrading program and its status of implementation are presented. Regulatory decisions derived from the performance-based approach have the advantage that they enable balancing of the overall plant risk and identifying at different plant levels the areas where improvements are necessary. (author)

  3. Methods and Model Development for Coupled RELAP5/PARCS Analysis of the Atucha-II Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Andrew M. Ward

    2011-01-01

    Full Text Available In order to analyze the steady state and transient behavior of CNA-II, several tasks were required. Methods and models were developed in several areas. HELIOS lattice models were developed and benchmarked against WIMS/MCNP5 results generated by NA-SA. Cross-sections for the coupled RELAP5/PARCS calculation were extracted from HELIOS within the GenPMAXS framework. The validation of both HELIOS and PARCS was performed primarily by comparisons to WIMS/PUMA and MCNP for idealized models. Special methods were developed to model the control rods and boron injection systems of CNA-II. The insertion of the rods is oblique, and a special routine was added to PARCS to treat this effect. CFD results combined with specialized mapping routines were used to model the boron injection system. In all cases there was good agreement in the results which provided confidence in the neutronics methods and modeling. A coupled code benchmark between U of M and U of Pisa is ongoing and results are still preliminary. Under a LOCA transient, the best estimate behavior of the core appears to be acceptable.

  4. Methods and Model Development for Coupled RELAP5/PARCS Analysis of the Atucha-II Nuclear Power Plant

    International Nuclear Information System (INIS)

    Ward, A.M.; Collins, B.S.; Xu, Y.; Downar, Th.J.; Madariaga, M.

    2011-01-01

    In order to analyze the steady state and transient behavior of CNA-II, several tasks were required. Methods and models were developed in several areas. HELIOS lattice models were developed and benchmarked against WIMS/MCNP5 results generated by NA-SA. Cross-sections for the coupled RELAP5/PARCS calculation were extracted from HELIOS within the Gen PMAXS framework. The validation of both HELIOS and PARCS was performed primarily by comparisons to WIMS/PUMA and MCNP for idealized models. Special methods were developed to model the control rods and boron injection systems of CNA-II. The insertion of the rods is oblique, and a special routine was added to PARCS to treat this effect. CFD results combined with specialized mapping routines were used to model the boron injection system. In all cases there was good agreement in the results which provided confidence in the neutronics methods and modeling. A coupled code benchmark between U of M and U of Pisa is ongoing and results are still preliminary. Under a LOCA transient, the best estimate behavior of the core appears to be acceptable

  5. The application of the PARCS neutronics code to the Atucha-I and Atucha-II NPPs

    Energy Technology Data Exchange (ETDEWEB)

    Ward, Andrew; Collins, Ben; Xu, Yunlin; Downar, Thomas [Purdue University, West Lafayette, IN (United States); Madariaga, Marcelo [Autoridad Nuclear Regulatoria, Buenos Aires (Argentina)

    2008-07-01

    In order to analyze Central Nuclear Atucha II (CNA-II) with coupled RELAP5/PARCS, extensive benchmarking of the neutronics codes HELIOS and PARCS was completed. This benchmarking was performed using a range of test problems designed in collaboration with NA-SA. HELIOS has been previously used to model Candu systems, but the results were validated for this case as well. The validation of both HELIOS and PARCS was performed primarily by comparisons to MCNP results for the same problems. Though originally designed to model light water systems, the capability of the PARCS was validated for predicting the performance of a Pressurized Heavy Water Reactor. The other noteworthy issue was the control rods. Because the insertion of the rods is oblique, a special routine was added to PARCS to treat this effect. Lattice level and Core level calculations were compared to the corresponding NA-SA codes WIMS and PUMA. In all cases there was good agreement in the results which provided confidence that the neutronics methods and the core neutronics modelling would not be a significant source of error in coupled RELAP5/PARCS calculations. (authors)

  6. Water chemistry of Atucha II PHWVR. Design concepts and evolution

    International Nuclear Information System (INIS)

    Chocron, Mauricio; Rodriguez, Ivanna; Duca, Jorge; Fernandez, Ricardo; Rico, Jorge

    2007-01-01

    Full text: Atucha II is a pressurized heavy water vessel reactor designed by Siemens-KWU, currently part of AREVA NP, of 745 MWe and similar to Atucha I, which has been in operation over 25 years. The primary heat transport system (PHTS) is composed by vertical channels (277-313 C degrees) that allocate the fuel elements while the moderator circuit is composed by a partially separated circuit (142-173 C degrees). The moderation power is transferred to the feedwater through the moderator heat exchangers (HX). These HXs operate as the last, high pressure water-steam cycle heaters as well. Materials (with exception of fuel channels and fuel sheaths which are made of zirconium alloys) are all austenitic steels while cobalt containing alloys have been all replaced at the design stage. Steam generator and moderator HX tubing are Alloy 800 made. The core is operated without boron except with the first fresh nucleus. The secondary circuit or Balance of plant (BOP) is similar in conception to that of a PWR but the moderator HXs. It is entirely built of ferrous alloys, has a feedwater-deaerator tank and moisture separator. The energy sink is the Rio de la Plata River. The Reactors Chemistry Department, Chemistry Division, National Atomic Energy Commission, in its character of R and D institution has been committed by CNA II-N.A.S.A Project to prepare the water chemistry specifications, water chemistry engineering and manuals, considering the type of reactor, design and construction aspects and operation characteristics, taking into account the current state-of-the art and worldwide standards. This includes conceptual aspects and implementation and operative aspects as well. This documentation will be released after a designer's review as it has been stated in the respective agreement. Respecting the confidentiality agreement between CNEA and NASA and the confidentiality regarding handling original documentation provided by the designer, it is considered illustrative to

  7. Xenon oscillation in a large PHWR core (Atucha II type): TRISIC code applicability

    International Nuclear Information System (INIS)

    Solanilla, Roberto

    2000-01-01

    A three dimensional nuclear reactor simulation code (TRISIC) was developed many years ago to design a PHWR (pressurizer heavy water reactors - Atucha type) based in the 'source-sink model' (heterogeneous theory). The limited processor computational performance available at that time was the constraint of the code when a detailed reactor description was necessary. A modern PC (pentium) code version with a full reactor core representation (461 fuel channels) including diagonal control rod banks and flux-reading detectors with theirs tube guide was used in the present paper for simulation of the Xenon transient when a local asymmetric perturbation was produced in a large core (Atucha II type). The results obtained and the computer time required for the 70 hour's simulation with an adequate time step, established the potential of the code to deal with this kind of transients. The paper shows that the method of TRISIC allows to detect and control azimuthal, radial and axial oscillation. This code is a proper way to elaborate a program of control rods movement from the flux reading detectors to damp the oscillation. TRISIC could also be a accurate tool to supervise the full core flux distribution in real time during the operation of the reactor. (author)

  8. Proposal of Ex-Vessel dosimetry for pressure vessel Atucha II

    International Nuclear Information System (INIS)

    Chiaraviglio, N.; Bazzana, S.

    2013-01-01

    Nuclear reactor dosimetry has the purpose of guarantee that changes in material mechanical properties of critical materials do not compromise the reactor safety. In PWR in which the top of the reactor vessel is open once a year, is possible to use Charpy specimens to measure the change in mechanical properties. Atucha II nuclear power plant is a reactor with on-line refueling so there is no access to the inside of the pressure vessel. Because of this, ex-vessel dosimetry must be performed and mechanical properties changes must be inferred from radiation damage estimations. This damage can be calculated using displacement per atom cross sections and a transport code such as MCNP. To increase results reliability it is proposed to make a neutron spectrum unfolding using activation dosimeters irradiated during one operation cycle of the power plant. In this work we present a dosimetry proposal for such end, made in base of unfolding procedures and experimental background. (author) [es

  9. The text of the agreement of 15 July 1981 between Argentina and the Agency for the application of Safeguards in connection with contracts concluded between the Comision Nacional de Energia Atomica Argentina and the Kraftwerk Union AG (Federal Republic of Germany) for the supply of the Atucha II Nuclear Power Plant

    International Nuclear Information System (INIS)

    1995-01-01

    The Agreement between the Republic of Argentina, the Federative Republic of Brazil, the Brazilian-Argentine Agency for Accounting and Control of Nuclear Materials and the International Atomic Energy Agency for the Application of Safeguards came into force on 4 March 1994. As a result of the coming into force of the aforesaid Agreement for Argentina, the application of safeguards under the Agreement of 15 July 1981 between Argentina and the IAEA for the application of safeguards in connection with contracts concluded between the Comision Nacional de Energia Atomic Argentina and the Kraftwerk Union AG (Federal Republic of Germany) for the supply of the Atucha II Nuclear Power Plant has been suspended

  10. Full scale dynamic tests of Atucha II NPP

    International Nuclear Information System (INIS)

    Konno, T.; Alvarez, L.M.; Ceballos, M.A.; Prato, C.A.; Uchiyama, S.; Godoy, A.R.

    1995-01-01

    This paper summarizes the main results of a series of dynamic tests of the reactor building of Atucha II NPP performed to determine the dynamic properties of its massive structure deeply embedded in Quaternary soil deposits. Tests were performed under two different types of loading conditions: Steady state harmonic loads imposed by mechanical exciters and impulsive loads induced by dropping a weight on the ground surface in the vicinity. Natural frequencies and mode shapes were identified and the associated modal damping ratios were experimentally determined. Numerical analyses of the reactor building-foundation system by two different F.E. models were performed. One of them, based on an axisymmetric representation of the soil-structure system, was used to simulate the steady state vibration tests and to calculate the dynamic stiffness of the foundation slab and soil layers for comparison with those experimentally obtained. The other, a 3-D F.E. model of the superstructure, was used to assess the natural frequencies and mode shapes obtained from the tests, representing dynamic stiffness of the foundation with stiffness coefficients derived both from the tests and from the axisymmetric F.E. model. Good agreement of the natural frequencies given by two types of tests was generally found, with the largest difference between them in the fundamental frequency of the building. Estimates of modal damping derived from the tests showed significant differences depending on the technique used to calculate them. For the fundamental mode damping was found to be 23-42 %, gradually decreasing with frequency to 2-4 % for around 10 Hz. (author)

  11. Full scale dynamic tests of Atucha II NPP

    International Nuclear Information System (INIS)

    Prato, C.A.; Ceballos, M.A.; Konno, T.; Uchiyama, S.; Alvarez, L.M.; Godoy, A.R.

    1998-01-01

    This paper summarizes the main results of a series of dynamic tests of the reactor building of Atucha II NPP performed to determine the dynamic properties of its massive structure deeply embedded in quaternary soil deposits. Tests were performed under two different types of loading conditions: steady state harmonic loads imposed by mechanical exciters and impulsive loads induced by dropping a weight on the ground surface in the vicinity. Natural frequencies and mode shapes were identified and the associated modal damping ratios were experimentally determined. Numerical analyses of the reactor building-foundation system by two different F.E. models were performed. One of them, based on an axisymmetric representation of the soil-structure system, was used to simulate the steady state vibration tests and to calculate the dynamic stiffness of the foundation slab and soil layers for comparison with those experimentally obtained. The other, a 3-D F.E. model of the superstructure, was used to assess the natural frequencies and mode shapes obtained from the tests, representing dynamic stiffness of the foundation with stiffness coefficients derived both from the tests and from the axisymmetric F.E. model. Good agreement of the natural frequencies given by two types of tests were generally found, with the largest difference between them in the fundamental frequency of the building. Estimates of modal damping derived from the tests showed significant differences depending on the technique used to calculate them. For the fundamental mode, damping was found to be 23-42%, gradually decreasing with frequency to 2-4% for ∝10 Hz. (orig.)

  12. Containment failure modes preliminary analysis for Atucha-I nuclear power plant during severe accidents

    International Nuclear Information System (INIS)

    Baron, J.; Caballero, C.; Zarate, S.M.

    1997-01-01

    The present work has the objective to analyze the containment behavior of the Atucha-I nuclear power plant during a severe accident, as part of a probabilistic safety assessment (PSA). Initially, a generic description of the containment failure modes considered in other PSAs is performed. Then, the possible containment failure modes for Atucha I are qualitatively analyzed, according to it design peculiarities. These failure modes involve some substantial differences from other PSAs, due to the particular design of Atucha I. Among others, it is studied the influence of: moderator/coolant separation, existence of cooling Zircaloy channels, existence of filling bodies inside the pressure vessel, reactor cavity geometry, on-line refueling mode, and existence of a double shell containment (steel and concrete) with an annular separation room. As a functions of the before mentioning analysis, a series of parameters to be taken into account is defined, on a preliminary basis, for definition of the plant damage states. (author) [es

  13. Experimental and numerical determination of the dynamic properties of the reactor building of Atucha II NPP

    International Nuclear Information System (INIS)

    Ceballos, M.A.; Car, E.J.; Prato, T.A.; Prato, C.A.; Alvarez, L.M.; Godoy, A.R.

    1995-01-01

    Determination of the dynamic properties of the reactor building of Atucha II NPP is carried out in order to: i) Obtain valuable information for seismic qualification of the plant, and ii) Assess some procedures for testing and analysis that are used in the process of seismic evaluation of existing nuclear facilities founded on Quaternary soil deposits. Both steady state and impulsive dynamic tests were performed but attention is centered here in tile techniques used to determine natural frequencies and modal damping ratios with impulsive tests. Numerical analyses were performed by means of a 3-D model model of the superstructure together with foundation stiffness coefficients derived in a separate paper from steady state vibration tests, and also from analysis with a 2-D F.E. model of the soil layers capable of approximating the 3-D features of the problem. The computed foundation stiffness coefficients are compared both with those obtained from the tests and from an axisymmetric F.E. model; results indicate that foundation stiffness coefficients calculated with F.E. models with soil parameters given by laboratory tests performed on cored samples are significantly lower than those given by the steady state vibration tests. (author)

  14. Comparison of vibration test results for Atucha II NPP and large scale concrete block models

    International Nuclear Information System (INIS)

    Iizuka, S.; Konno, T.; Prato, C.A.

    2001-01-01

    In order to study the soil structure interaction of reactor building that could be constructed on a Quaternary soil, a comparison study of the soil structure interaction springs was performed between full scale vibration test results of Atucha II NPP and vibration test results of large scale concrete block models constructed on Quaternary soil. This comparison study provides a case data of soil structure interaction springs on Quaternary soil with different foundation size and stiffness. (author)

  15. Operation of Atucha I nuclear power plant with 25 cooling channels without fuel elements

    International Nuclear Information System (INIS)

    Perez, R.A.; Sidelnik, J.I.; Salom, G.F.

    1987-01-01

    In view of the need of removing the irradiation probes from the reactor of Atucha I nuclear power plant, a study about the consequences of operating with 25 channels without their respective fuel elements was performed. This condition was simulated by means of the code PUMA symmetry I and the consequences were analyzed. From the study resulted a program of stepped power reduction of the nuclear plant that would take place during the process of channel emptying. (Author)

  16. Atucha I nuclear power plant: repair works in QK02W01 moderator system heat exchanger; Central nuclar Atucha I. Intervencion al intercambiador nro2 del moderador

    Energy Technology Data Exchange (ETDEWEB)

    Olivieri, Luis E; Zanni, Pablo A [Nucleoelectrica Argentina SA (NASA), Lima (Argentina). Central Nuclear Atucha 1

    2000-07-01

    reduction (46 tubes). The exchanger has 1049 tubes. Personnel involved in the repair works came from different areas: Atucha I/II nuclear power plants, SPC Department, Embalse nuclear power plant, NASA Headquarters, CNEA and INVAP specialists. (author)

  17. Atucha I nuclear power plant: repair works in QK02W01 moderator system heat exchanger

    International Nuclear Information System (INIS)

    Olivieri, Luis E.; Zanni, Pablo A.

    2000-01-01

    reduction (46 tubes). The exchanger has 1049 tubes. Personnel involved in the repair works came from different areas: Atucha I/II nuclear power plants, SPC Department, Embalse nuclear power plant, NASA Headquarters, CNEA and INVAP specialists. (author)

  18. Hydrogen combustion study in the containment of Atucha-I nuclear power plant

    International Nuclear Information System (INIS)

    Baron, J.H.; Gonzalez Videla, E.

    1997-01-01

    In this paper the combustion of hydrogen was modeled and studied in the containment vessel of the Atucha I nuclear power station using the CONTAIN package. The hydrogen comes from the oxidation of metallic materials during the severe accidents proposed. The CONTAIN package is an integrated tool that analyzes the physical, chemical and radiation conditions that affect the containment structure of the radioactive materials unloaded from the primary system during a severe accident in the reactor. (author) [es

  19. The experience gained at various stages of the Atucha nuclear power plant project

    International Nuclear Information System (INIS)

    Cosentino, J.O.

    1977-01-01

    The paper describes the experience gained in Argentina at the successive stages of planning, feasibility study, decision-making, awarding of contracts, construction and operation of the first nuclear power plant in Latin America. In particular, the operating experience accumulated so far is summarized together with the requirements for preparing operating tables for the plant. The role of the Atucha plant is also described in connection with the second plant under construction and the third in the planning stage [es

  20. Mixed wastes treatment in Atucha I; Tratamiento de residuos mixtos de la CNA I (central nuclear Atucha I)

    Energy Technology Data Exchange (ETDEWEB)

    Varani, J L; Comandu, J F [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

    1998-07-01

    Full text: During decontamination works of the fueling machine of Atucha I nuclear power plant (AINPP), a liquid waste with special characteristics was generated, which needed the development of a treatment method. The waste consisted of an emulsion designed for the cleaning of mechanical components and was formed by an organic solvent dispersed in water with aid of an emulsifier additive. After several cleaning operations, the emulsion contained an important quantity of lubricants and radioactive dirt. The treatment had the objective of recycling a toxic waste such as the organic solvent and reducing the volume of the residual mass. Laboratory tests were made tending to the emulsion separation in their components. Ionic force and ionic mobility were modified for join the emulsion micelles and produce their coalescence. Different salts and working temperatures were tried and it was stated that the combination of 1% of Na{sub 2}SO{sub 4} added and 40 degree C temperature were the optimum taking into account the available equipment in AINPP and cost considerations. The process was carried out in batch mode and 3 residual streams were obtained, an aqueous one which was sent to Residual Water System of AINPP, an organic liquid consisting of decontaminated hydrocarbons, useful for other cleaning tasks and finally a solid one, sited in the in-between interface of the other two liquids, consisting of insoluble soaps used as lubricant thickness, containing the principal proportion of radioactivity. As a result of this process we have achieved a volume reduction higher than 90%, the recycling of the organic solvent and concentration of radioactivity in a solid greasy mass with low water solubility. (author) [Spanish] Texto completo: Como resultado de tareas de descontaminacion de la maquina de carga de la central nuclear Atucha I (CNAI) se genero un residuo liquido de caracteristicas especiales, que requirio el desarrollo de un metodo de tratamiento. El residuo consistia en

  1. Commissioning of the laboratory of Atucha II NPP. Implementation and optimization of analytical techniques, quality aspects

    International Nuclear Information System (INIS)

    Schoenbrod, Betina; Quispe, Benjamin; Cattaneo, Alberto; Rodriguez, Ivanna; Chocron, Mauricio; Farias, Silvia

    2012-09-01

    Atucha II NPP is a Pressurized Vessel Heavy Water Reactor (PVHWR) of 740 MWe designed by SIEMENSKWU. After some years of delay, this NPP is in advanced construction state, being the beginning of commercial operation expected for 2013. Nucleoelectrica Argentina (N.A.S.A.) is the company in charge of the finalization of this project and the future operation of the plant. The Comision Nacional de Energia Atomica (C.N.E.A.) is the R and D nuclear institution in the country that, among many other topics, provides technical support to the stations. The Commissioning Chemistry Division of CNAII is in charge of the commissioning of the demineralization water plant and the organization of the chemical laboratory. The water plant started operating successfully in July 2010 and is providing the plant with nuclear grade purity water. Currently, in the conventional ('cold') laboratory several activities are taking place. On one hand, analytical techniques for the future operation of the plant are being tested and optimized. On the other hand, the laboratory is participating in the cleaning and conservation of the different components of the plant, providing technical support and the necessary analysis. To define the analytical techniques for the normal operation of the plant, the parameters to be measured and their range were established in the Chemistry Manual. The necessary equipment and reagents were bought. In this work, a summary of the analytical techniques that are being implemented and optimized is presented. Common anions (chloride, sulfate, fluoride, bromide and nitrate) are analyzed by ion chromatography. Cations, mainly sodium, are determined by absorption spectrometry. A UV-Vis spectrometer is used to determine silicates, iron, ammonia, DQO, total solids, true color and turbidity. TOC measurements are performed with a TOC analyzer. To optimize the methods, several parameters are evaluated: linearity, detection and quantification limits, precision and

  2. Micrometeorological study of the Atucha Nuclear Power Plant site

    International Nuclear Information System (INIS)

    Berri, G.J.; Robbio, C.A.

    1986-01-01

    The evaluation of time meteorological data obtained at the micrometeorological station of the Atucha Power Plant during 1979, is presented. Special attention is given to the transport and atmospheric dispersion characteristics through the evaluation of the mean and hourly wind behaviour and the stability classes. Furthermore, it is obtained an estimation of the dispersion factors both for short-term and long-term releases using Gaussians models. As these factors are representative of mean conditions, they should not be applied to the analysis of isolated situations. Finally it is emphasized that, although the results were obtained by means of 1979 data, significative differences are not expected for other years. (M.E.L.) [es

  3. Implementation of the utilization program for the fuel elements of the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Martin, H.R.; Serra, O.H.; Parker, Alejandro

    1981-01-01

    The programming operation for the use of the fuel elements in the Atucha-1 nuclear power plant was initially under the responsibility of the KWU Company, as part of the services rendered due for the manufacturing of said elements. This job was done with the help of the TRISIC program, developed in the early seventies by CNEA and SIEMENS staff. From april 21, 1979 on, CNEA took over the responsibility and strategy of the interchange of fuel elements. The several stages carried out for the implementation of this service are detailed. (M.E.L.) [es

  4. Model and simulation of the hydraulic turbine speed regulator of the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Copparoni, G.; Etchepareborda, A.; Urrutia, G.

    1992-01-01

    The hydraulics turbines of Atucha I Nuclear Power Plant takes advantage of condenser cooling water level difference between the plant and the river to recover about 2,5 MW e. It also supplies emergency power until diesel generators start up. Speed regulation is needed due to the transients that during this process occur. The purpose is to minimize the diesels start up time, and to avoid overshoots on the internal grid frequency. The hydraulic turbine, its speed regulator and the electric system associated with this transient have been modeled. The models and some simulation results are presented in this work. (author)

  5. Activity determination for neutron dosimetry in the vigilance programme for the pressure vessel in Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Furnari, J.C.; Cohen, I.M.; Ciriani, D.F.; Helzel Garcia, J.

    1993-01-01

    The methodologies for the activity determination of Co-60, Nb-93m and Nb-94 in flux monitors are presented. This was done in order to evaluate dose and damage caused by radiation received by pressure vessel materials of the Atucha I nuclear power plant for its surveillance program. (author)

  6. Temporary storage facility for spent nuclear fuels at the Atucha I nuclear power station (CNA)

    International Nuclear Information System (INIS)

    Wasinger, K.

    1983-01-01

    According to plans of the Argentine Atomic Energy Commission (CNEA), the spent nuclear fuel elements of the Atucha I Nuclear Power Station are to be stored temporarily pending a decision about the ultimate disposal concept. The holding capacity of the first fuel storage facility built by the German KWU together with the whole power plant had been expanded in 1978 to a level good until mid-1982. In 1977, KWU drafted the concept of another fuel storage facility. Like the first one, it was designed as a wet storage system attached to the power plant installations and had a holding capacity of 6944 fuel elements, which corresponds to some 1100 te of uranium. This extends the storage capacity up until 1996. In 1978, KWU was commissioned by CNEA to plan the whole facility and deliver the mechanical and electrical equipment. CNEA themselves assumed responsibility for the construction work. The second fuel storage facility was commissioned three years after the start of construction. (orig.) [de

  7. Proposal of modification of the Atucha I nuclear power plant's emergency power supply system

    International Nuclear Information System (INIS)

    Palacio, Pedro; Dabove, Mario

    1989-01-01

    The emergency power supply system of Atucha I N.P.P. consists of three 50% diesel generators. During the transient from normal power supply to emergency power supply (approximately 15 seconds) an hydraulic generator takes care of the emergency system. By this way, the emergency busbars constitute themselves an interruption free system. The two emergency busbars work normally coupled. This proposal consists of the following modifications: 1) Add a new diesel generator in order to allow the operation with two diesel generators per busbar. 2) To work with the two emergency busbars not coupled as normal operation mode. 3) To eliminate the hydraulic generator from the emergency power supply system, in order to simplify the operation and to reduce the failure possibility. Without the hydraulic turbine generator, the emergency busbars loose the interruption free condition. For this reason, for the loads that are not able for this mode of operation and are connected to the emergency power supply system, two additional low-voltage interruption free busbars are necessary. Finally, this proposal is compared with the Atucha II N.P.P. emergency power supply system. (Author)

  8. Compact spent fuel storage at the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Antonaccio, Carlos; Conde, Alberto; Flores, Alexis; Masciotra, Humberto; Sala, Guillermo; Zanni, Pablo

    2000-01-01

    The object of this report is to verify the possibility to increase the available storage of irradiated fuel assemblies, placed in the spent fuel pools of the Atucha I nuclear power plant. There is intends the realization of structural modifications in the storage bracket-suspension beam (single and double) for the upper and lower level of the four spent fuel pools. With these modifications that increase the storage capacity 25%, would arrive until the year 2014, it dates dear for the limit of the commercial operation of nuclear power plant. The increase of the capacity in function of the permissible stress for the supports of the bracket-suspension beam. They should be carried out 5000 re-accommodations of irradiated fuel assemblies. The task would demand approximately 3 years. (author)

  9. Development of an acoustic emission equipment for valves of the Nuclear Power Station Atucha 1

    International Nuclear Information System (INIS)

    Giaccheta, R.; Lopez Pumarega, I.; Straus, A.; Ruzzante, J.; Herzovich, P.

    1994-01-01

    A four channel Acoustic Emission was developed by the Acoustic Emission Group, INEND Department, of the Atomic Energy Commission of Argentina, for the detection of leaks in valves of the pressurized air system: ''Sistema de desconexion de emergencias por acido deuteroborico''. Basically, the system consists of four piezoelectric transducers with their corresponding preamplifiers coupled to the piping close to the valves. The following stages: amplifiers, threshold levels, channel identifications and visual alarm system are gathered in a box. The system was installed in the controlled zone of the Nuclear Power Stations Atucha I. It was calibrated and works on line. The values shown on the display are registered daily in order to separate the normal values from the leak ones. (author). 4 refs, 9 figs

  10. Alternatives to reach safeguards goals at Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Palacios, E.; Orpet, P.; Marzo, M.; Valentino, L.; Vicens, H.

    2001-01-01

    Full text: This paper describes the main features of Atucha I Nuclear Power Plant and the current safeguards' approach applied to this installation by the International Atomic Energy Agency (IAEA) and the Brazilian-Argentine Agency for Accounting and Control of Nuclear Materials (ABACC). The reasons for not completely fulfilling the IAEA safeguards criteria with the current approach are also described and a conceptual proposal of an unattended system developed jointly by ABACC and the Nuclear Regulatory Authority of Argentina (ARN) is presented. Finally, the paper addresses an alternative proposal to the previous one aiming at fulfilling the above mentioned objectives. Atucha I Nuclear Power Plant (NPP) was built in the 70's and has been under operation since 1974, This is an On Load Reactor, moderated and refrigerated with heavy water (PHWR). From its starting up to about a year ago, this NPP operated with natural uranium fuel assemblies but presently the reactor core is fed with slightly enriched uranium fuel assemblies (0,85 %). This Plant generates up to 357 Mwe. An outstanding operating characteristic of this power reactor is that low burn-up fuels assemblies already discharged into the pond may be re-used when necessary upon neutron flux requirements (re-shuffling). This installation has a pond storage capacity of about 10,000 fuel assemblies. At the highest power rate, the reactor core must be fed with a frequency of about 0,72 fuel assemblies per day. Before the application of the Agency Safeguards Criteria (IAEA-SC) in (1991), Atucha l had always satisfied the IAEA safeguards goals. Since 1991 the IAEA-SC demanded for On Load Reactors the control of the flow of irradiated fuel assemblies that leave or enter into the core (re-shuffling). By that time, Atucha I had been working for about seventeen years and there was no possibilities to install specific safeguards equipment without making significant construction modifications on this installation. Under the

  11. Pressure waves transient occurred in the steam generators feedwater lines of the Atucha-1 Nuclear Power Plant

    International Nuclear Information System (INIS)

    Balino, J.L.; Carrica, P.M.; Larreteguy, A.E.

    1993-01-01

    The pressure transient occurred at Atucha I Nuclear Power Plant in March 1990 is simulated. The transient was due to the fast closure of a flow control valve at the steam generators feedwater lines. The system was modelled, including the actuation of the relief valves. The minimum closure time for no actuation of the relief valves and the evolution of the velocity and piezo metric head for different cases were calculated. (author)

  12. Detection and analysis of thermal energy loss in the Atucha I nuclear power plant residual heat removal system

    International Nuclear Information System (INIS)

    Berra, Sandra; Guala, Mariana I.; Khon, Hector; Lorenzo, Andrea T.; Raffo Calderon, Maria C.; Urrutia, Guillermo

    1999-01-01

    It is presented the methodology used to detect and to measure energy losses which are existent in the Atucha I nuclear power plant. They were not directly detected, since the magnitude of those was below of the instrumentation precision which is used to measure the electric and thermal power in the plant. To achieve this work temperature special measurements were made. In this way it was possible to quantify the energy losses after operational long periods. (author)

  13. Operative modes of the primary circuit degasser of Atucha II N.P.P

    International Nuclear Information System (INIS)

    Rodriguez, Ivanna; Contino, Maximiliano; Chocron, Mauricio; Duca, Jorge

    2012-09-01

    Atucha II (N.A.S.A., Buenos Aires Province, Argentina) is a Pressurized Vessel Heavy Water Reactor designed by Siemens with a capacity of 740 MWe. After a long delay in construction the plant is close to the commissioning and among the many task that are carried out, chemistry and operation of devices related to it are under consideration [1]. As it is known, Hydrogen or Deuterium dosing has the purpose of both: limitation of the water radiolysis and to provide an appropriate reductive media for the structural materials, mainly stainless steel, A800 and Zr-4. Dealing with a heavy water plant, it is critical to determine whether it is necessary to add D 2 or if it is feasible to dose H 2 , by considering heavy water degradation and heavy water upgrading system capability. Those aspects have been previously analyzed and presented [2]. It is also necessary to consider blankets and venting locations that address to losses of the expensive D 2 . In the present work several alternatives of hydrogenation are presented and evaluated, considering the Degasser (D), the Volume Control Tank (TCV) and the special features of the purification and volume control system of a pressurized vessel heavy water plant where the primary circuit and moderator are partially mixed. Also the influence of venting through the pressurizer is analyzed. Conclusions are obtained in connection to (i) the maintenance of a permanent blanket of H 2 /He, 4%, in the TCV dome at a given initial pressure, (ii) The same but constant pressure to reach 0.6 ppm of H 2 in the Primary and Moderator water circuit, (iii) transients while reducing pressure in the Degasser and considering contribution of pressurizer venting, (iv) estimated contribution of the general corrosion of the system and (iv) differences if D 2 is used. (authors)

  14. Permission of change of limits in the vapor generators of the Atucha I Nuclear Central

    International Nuclear Information System (INIS)

    Ventura, M.

    2006-01-01

    In the mark of the modification of the Atucha-I Nuclear Central Installation (CNA-I) as consequence of the Introduction of the System 'Second Drain of Heat' (SSC), the Entity Responsible for the CNA-I (NASA) requested authorization to the Nuclear Regulatory Authority (ARN) to modify the value of the minimum level of water in the secondary side in the Steam generators (GVs) to activate the signal 'shoot of the Cut of the Reactor' (RESA-LLV). As the level in the GVs is one of those parameters that are used to shoot the Emergency Feeding System (RX), component of the SSC System, also was analyzed the change in the activation of the shoot signal of the 'Second Drain of Heat' (2SSC-LLV). The ARN uses for the study of the nuclear safety of nuclear power plants, the series of prediction programs RELAP5/MOD3.X. It participates of the evaluation and maintenance activities of these codes through specific agreements with the U.S. Nuclear Regulatory Commission (US-NRC). It is necessary to account with programs of this type since the ARN it licenses the construction and operation of Nuclear Power Plants (NPPs) and other outstanding facilities and it inquires its operation according to its own standards. With these tools its are auditing the calculations that the Responsible Entities of the operation make to guarantee the operability of the NPPs assisting the mentioned standards. The analysis with computational codes is used as a tool to achieve the best understanding in the behavior of the plant in union with the engineering approach, the manual calculations, the data analysis and the experience in the operation of the machine. (Author)

  15. Integrated ageing management of Atucha NPP

    International Nuclear Information System (INIS)

    Ranalli, Juan M.; Marchena, Martin H.; Zorrilla, Jorge R.; Antonaccio, Elvio E.; Brenna, Pablo; Yllanez, Daniela; Cruz, Gerardo Vera de la; Luraschi, Carlos; Sabransky, Mario

    2013-01-01

    Atucha NPP is a two PHWR unit site located in Lima, Province of Buenos Aires, 120 km north of Buenos Aires, Argentina. Until recent, the site was split in Atucha I NPP, a 350 MW pressure vessel heavy water reactor in operation since 1974; and Atucha II, a similar design reactor, twice as big as Atucha I finishing a delayed construction. With the start-up of Atucha II and aiming to integrate the management of the plants, the Utility (Nucleolectrica Argentina Sociedad Anonima - NASA) has reorganized its operation units. Within this reorganization, an Ageing Management Department has been created to cope with all ageing issues of both Atucha I and II units. The Atomic Energy Commission of Argentina (Comision Nacional de Energia Atomica - CNEA) is a state-owned R and D organization that; among other functions such as designing and building research reactors, developing uranium mining and supplying radioisotopes to the medical market; is in charge of providing support and technological update to all Argentinean NPPs. The Ageing Management Department of Atucha NPP and the Ageing Management Division of CNEA has formed a joint working group in order to set up an Integrated Ageing Management Program for Atucha NPP following IAEA guidelines. In the present work a summary of the activities, documental structure and first outputs of the Integrated Ageing Management Program of Atucha NPP is presented. (author)

  16. Integrated Ageing Management of Atucha NPP

    International Nuclear Information System (INIS)

    Ranalli, J.M.; Marchena, M.H.; Zorrilla, J.R.; Sabransky, M.

    2012-01-01

    Atucha NPP is a two PHWR unit site located in Lima, Province of Buenos Aires, 120 km north of Buenos Aires, Argentina. Until recent, the site was split in Atucha I NPP, a 350 MW pressure vessel heavy water reactor in operation since 1974; and Atucha II, a similar design reactor twice as big as Atucha I finishing a delayed construction . With the start-up of Atucha II and aiming to integrate the management of the plants, the Utility (Nucleolectrica Argentina Sociedad Anonima - NASA) has reorganized its operation units. Within this reorganization, an Ageing Management Department has been created to cope with all ageing issues of both Atucha I and II units. The Atomic Energy Commission of Argentina (Comision Nacional de Energia Atomica - CNEA) is a state-owned R and D organization that; among other functions such as designing and building research reactors, developing uranium mining and supplying radioisotopes to the medical market; is in charge of providing support and technological update to all Argentinean NPPs. The Ageing Management Department of Atucha NPP and the Ageing Management Division of CNEA has formed a joint working group in order to set up an Integrated Ageing Management Program for Atucha NPP following IAEA guidelines. In the present work a summary of the activities, documental structure and first outputs of the Integrated Ageing Management Program of Atucha NPP is presented. (author)

  17. Integrated ageing management of Atucha NPP

    Energy Technology Data Exchange (ETDEWEB)

    Ranalli, Juan M.; Marchena, Martin H.; Zorrilla, Jorge R.; Antonaccio, Elvio E.; Brenna, Pablo; Yllanez, Daniela; Cruz, Gerardo Vera de la; Luraschi, Carlos, E-mail: ranalli@cnea.gov.ar [Gerencia Coordinacion Proyectos CNEA-NASA, Comision Nacional de Energia Atomica, Buenos Aires (Argentina); Sabransky, Mario, E-mail: msabransky@na-sa.com.ar [Departamento Gestion de Envejecimiento, Central Nuclear Atucha I-II Nucleoelectrica Argentina S.A., Provincia de Buenos Aires (Argentina)

    2013-07-01

    Atucha NPP is a two PHWR unit site located in Lima, Province of Buenos Aires, 120 km north of Buenos Aires, Argentina. Until recent, the site was split in Atucha I NPP, a 350 MW pressure vessel heavy water reactor in operation since 1974; and Atucha II, a similar design reactor, twice as big as Atucha I finishing a delayed construction. With the start-up of Atucha II and aiming to integrate the management of the plants, the Utility (Nucleolectrica Argentina Sociedad Anonima - NASA) has reorganized its operation units. Within this reorganization, an Ageing Management Department has been created to cope with all ageing issues of both Atucha I and II units. The Atomic Energy Commission of Argentina (Comision Nacional de Energia Atomica - CNEA) is a state-owned R and D organization that; among other functions such as designing and building research reactors, developing uranium mining and supplying radioisotopes to the medical market; is in charge of providing support and technological update to all Argentinean NPPs. The Ageing Management Department of Atucha NPP and the Ageing Management Division of CNEA has formed a joint working group in order to set up an Integrated Ageing Management Program for Atucha NPP following IAEA guidelines. In the present work a summary of the activities, documental structure and first outputs of the Integrated Ageing Management Program of Atucha NPP is presented. (author)

  18. Volatile amines treatment: Corrosion rates and Atucha I nuclear power plant experience

    International Nuclear Information System (INIS)

    Iglesias, Alberto M.; Jimenez Rebagliati, Raul; Raffo Calderon, Maria C.; Manzi, Ricardo

    2000-01-01

    Steam generators water treatment with volatile amines in place of ammonia is usual today. This option seems an acceptable alternative to the generalize use of ammonia-sodium phosphate and has advantages when copper alloys are present. There are several amines that can work as corrosion inhibitor but the most useful for plant applications are: morpholine, ethanolamine and cyclohexylamine. In this work, are present the obtained results of corrosion rates measurements by electrochemical methods. The hydrothermal conditions of our experiences were similar to that of the Atucha I nuclear power plant (CNA I). pH, conductivity and dissolved oxygen measures were correlated with corrosion rates of the CNA I materials as carbon steel and admiralty brass. The faradaic impedance spectroscopy techniques allows a more detailed interpretation of corrosion rates process. Morpholine and ammonia behavior can be evaluated under power plant operations conditions with the accumulated experience of CNA I. Results are present throughout material release and his effects over heat transfer parameters. (author)

  19. Irradiation program of slightly enriched fuel elements at the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Casario, J.A.; Cesario, R.H.; Perez, R.A.; Sidelnik, J.I.

    1987-01-01

    An irradiation program of fuel elements with slightly enriched uranium is implemented, tending to the homogenization of core at Atucha I nuclear power plant. The main benefits of the enrichment program are: a) to extend the average discharge burnup of fuel elements, reducing the number of elements used to generate the same amount of energy. This implies a smaller annual consumption of elements and consequently the reduction of transport and replacement operations and of the storage pool systems as well as that of radioactive wastes; b) the saving of uranium and structural materials (Zircaloy and others). In the initial stage of program an homogeneous core enrichment of 0.85% by weight of U-235 is anticipated. The average discharge burnup of fuel elements, as estimated by previous studies, is approximately 11.6 MW d/kg U. The annual consumption of fuel elements is reduced from 396 of natural uranium to 205, with a load factor of 0.85. It is intended to reach the next equilibrium steps with an enrichment of 1.00 and 1.20% in U-235. (Author)

  20. Atucha I nuclear power plant azimuthal ex-vessel flux profile evaluation

    International Nuclear Information System (INIS)

    Ferraro, Diego

    2008-01-01

    Irradiation damage in RPV (Reactor Pressure Vessel) in nuclear power plants is a key parameter to be analyzed in order to assess the plant integrity up to end of life and planning for a possible plant life extension. In this work a neutronic model in MCNP that represents a sector of 30 degrees of the Atucha I power plant nucleus has been consolidated with the results of an ex-vessel dosimetry made in the outer surface of the RPV s power plant in order to analyse the irradiation damage through the dpa rate. A strong dependents of the maximum point of damage with the loading of a peripheral channel was found, so a mitigation strategy was proposed, which is basically to empty this channel and its analogs in the rest of the nucleus. Analysing this second case a notable decrease of the damage is found in the zone considerated on the model (shown through the drop of de dpa rate in the zone). [es

  1. Summary of severe accident assessment for Atucha 2 Nuclear Power Plant using RELAP5/SCDAPSIM Mod3.6

    International Nuclear Information System (INIS)

    Bonelli, Analia; Mazzantini, Oscar; Siefken, Larry; Allison, Chris

    2014-01-01

    A severe accident assessment was performed for the Atucha 2 Nuclear Power Plant in Argentina. Atucha 2 is a PHWR, cooled and moderated by heavy water, presently in commissioning process. Its 451 fuel assemblies are 6.03m high and each composed of 37 Zircaloy clad fuel rods. Each assembly is placed inside an individual Zircaloy coolant channel. Heavy water coolant flows inside the channels which are all immersed inside the moderator tank. The RPV lower plenum is occupied by a massive steel structure called 'filling body' that was designed to minimize heavy water inventory. Due to some unique design characteristics, severe accident progression in Atucha 2 is expected to be somewhat different from that predicted for regular PWRs. Therefore, a very detailed assessment was performed, focused on the different accident stages and expected phenomena by the use of different input models and nodalizations. When possible, linking to available experimental data was performed. RELAP/SCDAPSIM Mod 3.6 was the computer code selected to perform this task. The modeling of Atucha 2's unique characteristics required several extensions to the code. For the severe accident assessment of Atucha 2, three different input models were developed that were key instruments for the debugging and evaluation process. A Single Channel Model was used to evaluate the first stages of core heatup (including the boiloff of the channels and moderator tank), an RPV standalone model was used to assess the interaction between components in the complete core and for the evaluation of late in-core melting and relocation. Then, a Lower Plenum standalone model was developed to assess the behavior of the melted and slumped core material on top of the filling body and to analyze ex-vessel cooling as a possible severe accident management action. For each of the cases, highlights of key results are shown and general conclusions are drawn. In the case of a severe accident with significant meltdown of

  2. Measurement of the CNA I's (Atucha I nuclear power plant) control rods reactivity during its commissioning on January 8th, 1990

    International Nuclear Information System (INIS)

    Waldman, R.M.; Gomez, A.

    1990-01-01

    Measurements were made on integral and differential calibration of rod 16, fuel racks RG and R3 and extinction reactivity during Atucha I nuclear power plant's commissioning on January 8th., 1990. These were the first physical measurements performed after the first critical nuclear power plant's commissioning. (Author) [es

  3. Neutronic calculations for the reactor pressure vessel of Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Lerner, Ana M.; Madariaga, Marcelo R.

    1999-01-01

    In 1974 a surveillance program for the Atucha I nuclear power plant pressure vessel was initiated which included the construction of different types of specimens, distributed in 30 irradiation capsules located under the core at the lower part of some of the fuel channels. The capsules containing the irradiated specimens were withdrawn in two stages; the first set (SET 1) of 15 specimens in 1980 and the second one (SET 2) of the remaining 15, in 1987. Both fracture mechanic tests and dosimetry analysis were carried out by the designer (KWU) for SET1 and by the owner National Atomic Energy Commission (CNEA) for SET2. The calculations performed in the case of SET1 showed that there was a significant spectrum difference between the position where the specimens had been and the reactor pressure vessel (RPV) - inner surface (IS). It was established that the ratio of thermal flux (E 1 MeV) varied, approximately, from 1000 to 10 from the irradiation position to the RPV- IS. The purpose of this report is to show the calculations recently performed at the Nuclear Regulatory Authority, with particular emphasis on the difference in the results generated by the modification to sightly enriched fuel. A simplified 1-D calculations show that there is a slight increase (4% approximately) in the flux along the whole energy range. As it has already been mentioned, this is due, more than to the isotopic composition of the new fuel, to the difference in power density spatial distribution, which is a consequence of a different fuel management, necessary to preserve operational limits below their maximum allowed values with the same total thermal power generated. More detailed calculations are nevertheless foreseen in order to verify these first results. (author)

  4. GIS Application in Atucha I Nuclear Power Plant Exercise Argentina, 2007

    International Nuclear Information System (INIS)

    Sadaniowski, I.V.; Telleria, D.M.; Jordan, O.D.; Boutet, L.I.; Kunst, J.J.; Bruno, H.A.; Hernandez, D.G.; Rodriguez, M.; Cateriano, M.A.; Rey, H.L.

    2011-01-01

    Geographic Information Systems (GIS) are tools applied to assist in the assessment and solution of many geographical related issues. Recently, their applications have been extended to the areas of disasters and environmental emergencies. GIS not only could be used as a diagnostic tool. Combined with adequate information and other tools capable to predict the transfer of pollutants in the environment and the associated impacts to the public, GIS could be used to support emergency planning and response. The Nuclear Regulatory Authority (NRA) of Argentina has incorporated in 2003 the GIS technology like an innovative resource for its preparation and response activities in emergencies. For this, the NRA acquired the necessary technology (hardware and software) and the technical specialist who were joined to expert's team in the nuclear and radiological emergencies field. The GIS stays operative as support tool in the Emergencies Control Center of NRA. In this paper, the use of GIS as a tool for analysis and advice is presented. The GIS is being used for preparation and development of nuclear emergencies trials and exercises, carried out on-site and off-site at the Nuclear Power Plant Atucha I Buenos Aires, Argentina, in cooperation with civil defense, national and state security and army forces and intensive public involvement. The databases were conformed with information from different sources, including the result of interviews to different actors, as well as other local and national government agencies and forces. Also, educational institutions, local medical centers, etc., were consulted. The information was enriched with outings to field in the surroundings of nuclear power plant. The scope and the detail of the information for this exercise covers 30 kilometers surroundings the nuclear power plant, with a range of significantly different geographical and population conditions. When loading the information in the GIS, a classification scheme is applied and

  5. GIS application in Atucha I nuclear power plant exercise Argentina, 2007

    International Nuclear Information System (INIS)

    Sadaniowski, Ivana; Jordan, Osvaldo; Boutet, Luis; Kunst, Juan; Bruno, Hector; Hernandez, Daniel; Rodriguez, Monica; Cateriano, Miguel; Rey, Hugo; Telleria, Diego

    2008-01-01

    Full text: Geographic Information Systems (GIS) are tools applied to assist in the assessment and solution of many geographical related issues. Recently, their applications have been extended to the areas of disasters and environmental emergencies. GIS not only could be used as a diagnostic tool. Combined with adequate information and other tools capable to predict the transfer of pollutants in the environment and the associated impacts to the public, GIS could be used to support emergency planning and response. The Nuclear Regulatory Authority (NRA) of Argentina has incorporated in 2003 the GIS technology like an innovative resource for its preparation and response activities in emergencies. For this, the NRA acquired the necessary technology (hardware and software) and the technical specialist who were joined to expert's team in the nuclear and radiological emergencies field. The GIS stays operative as support tool in the Emergencies Control Center of NRA. In this paper, the use of GIS as a tool for analysis and advice is presented. The GIS is being used for preparation and development of nuclear emergencies trials and exercises, carried out on-site and off-site at the Nuclear Power Plant Atucha I Buenos Aires, Argentina, in cooperation with civil defense, national and state security and army forces and intensive public involvement. The databases were conformed with information from different sources, including the result of interviews to different actors, as well as other local and national government agencies and forces. Also, educational institutions, local medical centers, etc., were consulted. The information was enriched with outings to field in the surroundings of nuclear power plant. The scope and the detail of the information for this exercise covers 30 kilometers surroundings the nuclear power plant, with a range of significantly different geographical and population conditions. When loading the information in the GIS, a classification scheme is applied

  6. C A R A fuel element for Atucha nuclear power plants and development plan

    International Nuclear Information System (INIS)

    Brasnarof, D. O; Marino, A. C; Bianchi, D; Giorgis M A; Orlando, O; Munoz, C; Taboada, H; Florido, P. C

    2006-01-01

    This paper presents the current state and the development plan of the C A R A fuel element.Main activities were carried out towards to welding of the end plates of the C A R A fuel element by a new process, and the assembling and hanging of the C A R A fuel element in its Atucha configuration, by using an external basket [es

  7. ND online software development for data acquisition of replacement operations of fuel assemblies of Atucha I Nuclear power plant

    International Nuclear Information System (INIS)

    Calvo, Maria Dolores; Wentzeis, Luis

    2012-01-01

    The ND Online software was developed in order to acquire data on a real-time basis of the refueling operations at the Atucha I nuclear power plant. The fuel elements containing slightly enriched uranium dioxide are located in the nuclear reactor core inside the cooling channels. The refueling operations are made periodically while the reactor is operating at full power. The acquired signals during the refueling operations are: pressure, force and position of the fuel element. In order to improve the safety and availability of the installation, monitoring of the refueling operations is important for the early detection of anomalies related to the fuel element itself, the cooling channels or the refueling machine (author)

  8. Assessment of theoretical and experimental results in the calculation of atmospheric dilution factors in the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Quintana, E.E.; Tossi, M.H.; Telleria, D.M.

    1990-01-01

    Collective doses produced during the normal working of the Atucha I Nuclear Power Plant are calculated using annual atmospheric factors. This work studies the behaviour of the dilution factors in different periods of the year in order to fit the calculated dose model applying factors from seasonal, monthly or weekly periods. The Radiation Protection Group of the C.N.E.A. have carried out continuous environmental monitoring in the surroundings of the Atucha I Nuclear Power Plant. These studies include the measurement of air tritium concentration, radionuclide that is found principally as tritiated water vapour. This isotope, normally released by the nuclear power plant was used as a tracer to assess the atmospheric dilution factors. Factors were calculated by two methods: an experimental one, based on environmental measurements of the tritium concentration in the surroundings of the nuclear power plant and another one by applying a theoretical model based on information from the micrometeorological tower located in the mentioned place. To carry out the environmental monitoring, four monitoring stations in the surroundings of the power plant were chosen. Three of them are approximately one kilometer from the plant and the fourth is 7.5 km away, near the city of Lima. To condense and collect the atmospheric water vapour, an overcooling system was used. The measurement was performed by liquid scintillation counting, previous alkaline electrolytical enrichment of the samples. The theoretical model uses hourly values of direction and wind intensity, as well as the atmospheric dispersive properties. Values obtained during the period 1976 to 1988 allowed, applying statistical tests, to validate the theoretical model and to observe seasonal variation of the dilution factors throughout the same year and between different years. Finally, results and graphics are presented showing that the behaviour of the dilution factors in different periods of the year. It is recommended to

  9. Theoretical-experimental assessment of the variables affecting fretting of Atucha I nuclear power plant utility steam generators tubes

    International Nuclear Information System (INIS)

    Kulichevsky, Raul M.

    1995-01-01

    Fretting wear of Steam Generator tubes caused by flow induced vibrations generates uncertainty on their integrity. The knowledge of the controlling variables of the wear process may give a criterion to evaluate the tubes residual life. Information on vibratory response and dynamic interaction between tubes and their supports are prerequisites for understanding the relationship between fretting wear and tube vibration. Experimental results of the vibratory response of an Atucha-I nuclear power plant type U-tube, the influence of tube/support clearance on this response and a study of tube/support dynamic interaction, which allow the verification of a finite element model of this type of tubes, are presented in this work. Also wear results for the Incoloy 800/DIN 1.4550 austenitic stainless steel pair of materials and a first evaluation of the wear constant of this pair are presented. (author)

  10. Simulation program for the dynamic behaviour of the primary system and moderators's circuit of the Atucha I Nuclear Power Plant

    International Nuclear Information System (INIS)

    Castano, Jorge; Gvirtzman, H.A.

    1981-01-01

    A model of digital computation is presented to simulate the primary system of heat transportation, moderator system and the associated systems for adjustment, regulation and control in the PHWR reactor at the Atucha-1 nuclear power plant. The model discusses in a concentrated way the different components and allows the study of the dynamical behaviour of the power plant facing disturbances with respect to a state of stationary regime. General considerations and description of the model are made. The method is described showing flow sheets, graphs and developing basic formulas, simulating a primary system, moderator and secondary system of the steam generator and the main system of regulation. Also an analysis of the results is made, for the case of disturbances which reduce or increase the power of the reactor by 10%. (V.B.) [es

  11. Economical benefits for the use of slightly enriched fuel elements at the Atucha-I nuclear power plant

    International Nuclear Information System (INIS)

    Sidelnik, J.I.; Sosa, M.A.

    1987-01-01

    The fuel represents a very important factor in the operative cost of the Atucha I nuclear power plant. This cost is drastically reduced with the use of fuel elements of slightly enriched uranium. The annual saving is analyzed with actual values for fuel elements with an enrichment of 0.85% by weight of U-235. With the reactor core in equilibrium state the annual saving achieved is approximately 7.5-10 u$s. According to the present irradiation plan, the benefit for the transition period is studied. An analysis of the sensitivity to differential increments in factors determining the cost of fuel elements or to changes in manufacturing losses is also performed, calculating its effect on the waste, the storage of irradiated elements and the amount of UO 2 required. (Author)

  12. Main pumps lost incident in the nuclear power plant Atucha I. Modelling with RELAP5/MOD3.2

    International Nuclear Information System (INIS)

    Ventura, M.A.; Rosso, R.D.

    1998-01-01

    Time evolution of natural circulation in the nuclear power plant Atucha I (CNA-I), in a main pumps lost incident because of the lost of external power feed, is analyzed. It leads to a strong stop transient, without an important blow down, from a forced nominal flow to a natural circulation one. The results are obtained from RELAP5/MOD3.2 code's modeling. The study is based on the refrigeration conditions analysis, during the first minutes of the reactor out of service. Previously to the transient, work had been done to obtain the plant steady state, with design parameters in operation conditions at 100 % of power. The object is that the actual plant state would be represented. In this way, each plant part (steam generators, reactor, pressurizer, pumps) had been modeled in separated form with the appropriate boundary conditions, to be used in the whole circuit simulation. The developed model, had been validated making use of the comparison between the values obtained to the principal thermodynamic parameters with the plant recorded values, in the same incident. The results are satisfactory in a way. On the other hand, it has suggested some modeling changes. The RELAP5/MOD3.2 capability to model the thermodynamic phenomena in a PHWR plant has been verified when, according to the mentioned incident, the flow pass from a nominal forced flow, to one which is governed by natural circulation, still with the CNA-I untypical design conditions. (author) [es

  13. Thermoelastic analysis for the fuel claddings of the nuclear power reactor at Atucha in the skid's region

    International Nuclear Information System (INIS)

    Sanchez Sarmiento, Gustavo; Basombrio, F.G.

    1979-01-01

    For the fuel elements of the Nuclear Power Reactor at Atucha, a two-dimensional thermoelastic analysis has been made in the region of the skids of the fuel cladding, when the gap between them and the fuel rod separator's support becomes zero. In such a case the latter forces exert on the skids an elastic reaction opposite to the cladding's expansion. The internal pressure reaching the yield stress for the cladding material has been calculated, as a function of the initial gap; for several possible fuel rod locations within the separator; for the actual dimensions and also for reduced thickness of the cladding; with a given external pressure and, with a known temperature spatial distribution. The latter has been calculated by solving the heat conduction equation along the fuel element for a certain power level in the reactor. The calculations are made with two FORTRAN IV computer codes developed at C.A.B., using the finite-element method: the NOLICUARM, to solve the nonlinear quasi-harmonic equation, and the ELASTEF 3, for the solution of thermoelastic problems with plane symmetry. (author) [es

  14. The CNA-1 (Nuclear Power Plant Atucha-1) QK-01 repairing project; Proyecto de reparacion del QK-01 de CNA-1 (Central Nuclear Atucha-1)

    Energy Technology Data Exchange (ETDEWEB)

    Pizzaferri, J C [Nucleoelectrica Argentina S.A. (NASA), Buenos Aires (Argentina); Cabot, P [Comision Nacional de Energia Atomica, San Martin (Argentina). Centro Atomico Constituyentes

    1998-12-31

    The repair/maintenance of the CNA-1 QK-01 Moderator Cooler will be a leading case of the repair of a class 1 nuclear component in a high radiation environment; utilizing for the work, sophisticated remotely operated equipment. This paper describes the component, the repair-maintenance objective, and the equipment-procedures developed for the intervention. (author) 9 refs., 5 figs. [Espanol] La reparacion/mantenimiento del enfriador del moderador QK-01 de CNA-1 sera un caso sobresaliente de reparacion de un componente nuclear clase 1 en un ambiente de alta radiacion, utilizando equipamiento sofisticado de accion remota. En este trabajo se describen caracteristicas de diseno del componente, objetivos del mantenimiento, y las facilidades y procedimientos desarrollados en el marco del proyecto para realizar la tarea en cuestion. (autor)

  15. Application of the weld in maintenance mechanics at the Nuclear Power Station Atucha I; Aplicaciones de soldadura en mantenimiento de la Central Nuclear Atucha I

    Energy Technology Data Exchange (ETDEWEB)

    Cosentino, R E

    1988-12-31

    The application of the `weld procedures`, in the field of activity of nuclear power is an special chapter of weld. The so called `Nuclear Installations` are actually under control from their contruction up to their life extension operation to special control programs and quality assurance. This situation obliges the implementation of procedures to assure the fulfilment of the programs for the need to make the reparations or mechanics construction. This paper describes the considerations that has been taken into account to repare some components of the plant. The works carried out constitute applications to the TIG weld procedure. The `lip weld` is a mechanic component required in pressurized systems subject to air pressure. (Author).

  16. Atucha II: building a 745 MWe pressure-vessel PHWR in the Argentine

    International Nuclear Information System (INIS)

    Madero, C.C.

    1982-01-01

    The history of the development of nuclear power in Argentina is outlined. Future policy is described. The aim is to close the fuel cycle and secure a domestic capability to design and construct nuclear power plants. (U.K.)

  17. Improvements related with the safety required by the Argentine Regulatory Authority to the Atucha I Nuclear Central; Mejoras relacionadas con la seguridad requeridas por la Autoridad Regulatoria Argentina a la Central Nuclear Atucha I

    Energy Technology Data Exchange (ETDEWEB)

    Calvo, J.; Michelin, C.; Navarro, R.; Waldman, R. [Autoridad Regulatoria Nuclear, Av. Del Libertador 8250, 1429- Ciudad de Buenos Aires (Argentina)]. e-mail: jcalvo@sede.arn.gov.ar

    2006-07-01

    The Argentinean Nuclear Regulation Authority (ARN) verified the existence of changes in the state of some internal components of the reactor of the Atucha I Nuclear Power station that, of continuing in the time, it could take to an inconvenient degradation for the safety operation of the installation. In consequence, to the effects of preventing that reach this situation, at the end of 1999, the ARN required to the Responsible Entity for the operation of this power station the implementation of an important improvements program in the internal components of the reactor. Additionally, and based on the results of the Probabilistic Safety analysis, it was added the one mentioned improvements program the implementation of an alternative cooling system of the reactor core denominated Second Drain of Heat, due to it was determined that, for some accidental sequences, their performance would reduce considerably the probability of damage to the core. The concretion of the improvements program implied to the Responsible Entity the realization of an important quantity of engineering studies, tests and specific inspections that allowed to carry out changes on the control bars of the reactor and its guide tubes; the coolant channels; the sensors of neutron flow; and diverse components of the primary and moderator systems. On the other hand also it was implemented the system Second Drain of Heat, what represents a considerable effort to make compatible the instrumentation and control of last generation, with the instrumentation and existent control systems in the power station. Also, it was requested to be carried out an integrity of the pressure recipient for to demonstrate the existence of an acceptable margin for the difference among the acceptable limit temperatures and of ductile/fragile transition of the material for all the possible accidental scenarios during the useful life of the reactor. (Author)

  18. Sampling and characterization of spent exchange resins of Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Varani, Jose L.; Cernadas, D.; Iglesias, Alberto M.; Raffo Calderon, Maria del C.

    2004-01-01

    The present storage facilities for spent resins in Atucha I NPP would be full within the next 5 years, accordingly some tasks are being planned to conditioning these waste and so generate enough storage capacity for the remaining life of the plant. Among this tasks is the characterization of spent resins that has several objectives: to know their chemical and radiochemical composition; the homogeneity of these parameters in the total volume of spent resins; the existence or not of compact zones; the proportion of 'crud'; the breakage degree of the micro-spheres; etc. The first step was to analyse the criterion to follow for sampling resins in the storage deposit of 40 m 3 . In order to take some samples from different points, a special device was required. It had to be introduced closed in the resin bed, opened to take the sample and then closed again to return to the surface. A device used in cereal industry for sampling silos to different depths was modified in its internal capacity for reducing operator dose and increasing the length of rod in order to reaching the bottom of the pit. The device was tested in cold mock up before to taking actual samples. Active resins samples, five in total up to now, were taken from deposit to different depths and kept in lead containers. After analysing the samples, the following average results were extracted: 1.7 x 10 5 Bq/g of Co-60, 9.7 x 10 5 Bq/g of Cs-137 and 774 Bq/g of total alpha, which corresponds to intermediate activity waste. The differences between the values of activity of the different samples are of up to 310 % for Co-60 and of up to 788 % for the Cs-137 what indicates a great inhomogeneity. The direct observation of resin grains, placed in a transparent glass burette, did not demonstrate an important proportion of broken or divided resins. (author)

  19. Improvements related with the safety required by the Argentine Regulatory Authority to the Atucha I Nuclear Central

    International Nuclear Information System (INIS)

    Calvo, J.; Michelin, C.; Navarro, R.; Waldman, R.

    2006-01-01

    The Argentinean Nuclear Regulation Authority (ARN) verified the existence of changes in the state of some internal components of the reactor of the Atucha I Nuclear Power station that, of continuing in the time, it could take to an inconvenient degradation for the safety operation of the installation. In consequence, to the effects of preventing that reach this situation, at the end of 1999, the ARN required to the Responsible Entity for the operation of this power station the implementation of an important improvements program in the internal components of the reactor. Additionally, and based on the results of the Probabilistic Safety analysis, it was added the one mentioned improvements program the implementation of an alternative cooling system of the reactor core denominated Second Drain of Heat, due to it was determined that, for some accidental sequences, their performance would reduce considerably the probability of damage to the core. The concretion of the improvements program implied to the Responsible Entity the realization of an important quantity of engineering studies, tests and specific inspections that allowed to carry out changes on the control bars of the reactor and its guide tubes; the coolant channels; the sensors of neutron flow; and diverse components of the primary and moderator systems. On the other hand also it was implemented the system Second Drain of Heat, what represents a considerable effort to make compatible the instrumentation and control of last generation, with the instrumentation and existent control systems in the power station. Also, it was requested to be carried out an integrity of the pressure recipient for to demonstrate the existence of an acceptable margin for the difference among the acceptable limit temperatures and of ductile/fragile transition of the material for all the possible accidental scenarios during the useful life of the reactor. (Author)

  20. Criticality and shielding calculations of an interim dry storage system for the spent fuel from Atucha I Nuclear Power Plant

    International Nuclear Information System (INIS)

    Silva, M

    2006-01-01

    The Atucha I Nuclear Power Plant (CNA-I) has enough room to store its spent fuel (SF) in damp in its two pool houses until the middle of 2015.Before that date there is the need to have an interim dry storage system for spent fuel that would make possible to empty at least one of the pools, whether to keep the plant operating if its useful life is extended, or to be able to empty the reactor core in case of decommissioning.Nucleolectrica Argentina S.A. (NA-SA) and the Comision Nacional de Energia Atomica (CNEA), due to their joint responsibility in the management of the SF, have proposed interim dry storage systems.These systems have to be evaluated in order to choose one of them by the end of 2006.In this work the Monte Carlo code MCNP was used to make the criticality and shielding calculations corresponding to the model proposed by CNEA.This model suggests the store of sealed containers with 36 or 37 SF in concrete modules.Each one of the containers is filled in the pool houses and transported to the module in a transference cask with lead walls.The results of the criticality calculations indicates that the solutions of SF proposed have widely fulfilled the requirements of subcriticality, even in supposed extreme accidental situations.Regarding the transference cask, the SF dose rate estimations allow us to make a feedback for the design aiming to the geometry and shielding improvements.Regarding the store modules, thicknesses ranges of concrete walls are suggested in order to fulfill the dose requirements stated by the Autoridad Regulatoria Nuclear Argentina [es

  1. Design, test and start up of a cleaning system for the moderator tank bottom of Atucha I Nuclear Power Plant

    International Nuclear Information System (INIS)

    Duca, J.; Gerber, O.; Ibero, M.; Riga, N.

    1989-01-01

    In order to perform the cleaning of the moderator tank bottom, during the repair of the Atucha I nuclear power plant (CNA I) failure, the Empresa Nuclear Argentina de Centrales Electricas (ENACE S.A.) designed a system with the following requirements (asked by CNA I): a) To aspirate and retain free solid particles, uranium dioxide pellets and coolant channels isolations (foils) of minor size settled at the moderator tank bottom, being the reactor at middle loop state. b) To allow a radially cleaning up to 1.4 m from the extracted channel. c) To design a lay-out attaining the ALARA dose exposure. The designed system basically consists in: a) Flexible intake for suction: allows the movement inside the moderator tank and provides the adequate speed to raise the particles. b) Filter: retains the aspirated particles, pellets and foils. Its capacity is 1.8 dm 3 and the minimum size of retained particles is 200 m. The ALARA dose exposure concept is attained due to that the filter is located inside the moderator tank. c) Filtering column: contains the filter and allows the entrance of the extraction and exchange tool (for the flexible intake and filter). d) Suction hose: connects the filtering column with the pump. Its flexibility allows its use in any channel maintaining the same positions of the discharge pump and the return piping. e) Discharge pump: it is a canned centrifugal pump with low-low net positive suction head. f) Return piping: discharges the filtered water into the moderator tank. The system fulfilled satisfactorily all requirements during its operation. (Author)

  2. Atucha II: building a 745 MWe pressure-vessel PHWR in the Argentine

    International Nuclear Information System (INIS)

    Leibovich, H.; Coll, J.; Backhaus, K.

    1982-01-01

    The history of the Comision Nacional de Energia Atomica and its involvement with the development in Argentina of a nuclear power plant manufacturing technology is described. The way in which technology transfer from German industry has enabled increasing Argentine participation is outlined. The formation of the joint Argentine/German engineering company ENACE in 1981 will promote this process. (U.K.)

  3. Operational experience with the first eighteen slightly enriched uranium fuel assemblies in the Atucha-1 nuclear power plant

    International Nuclear Information System (INIS)

    Higa, M.; Perez, R.; Pineyro, J.; Sidelnik, J.; Fink, J.; Casario, J.A.; Alvarez, L.

    1997-01-01

    Atucha I is a 357 Mwe nuclear station, moderated and cooled with heavy water, pressure vessel type of German design, located in Argentina. Fuel assemblies (FA) are 36 active natural UO2 rod clusters, 5.3 meters long and fuel is on power. Average FA exit burnup is 6 MWd/kg U. The reactor core contains 252 FA. To reduce the fuel costs about 6 MU$S/yr a program of utilization of SEU (0.85 %w U235) fuel was started at the beginning of 1995 with the introduction of 12 FA in the first step. The exit burnup of FA is approx. 10 MWd/kgU. It is planned to increase gradually the number of them up to having a full core with SEU fuel with an expected FA average exit burnup of 11 MWd/kgU. The SEU program has also the advantage of a strong reduction of spent fuel volume, and a moderate reduction of fuelling machine use. This paper presents the satisfactory operation experience with the introduction of the first 12 SEU fuel assemblies and the planned activities for the future. The fresh SEU fuel assemblies were introduced in six fuel channels located in an intermediate zone located 136 cm from the center of the reactor and selected because they have higher margins to the channel powers limits to accommodate the initial 15 to 20 % relative channel power increase. To verify the design and fuel management calculations, comparisons have been made of the calculated and measured values of the variation of channel ΔT, regulating rods insertion and flux reading in in-core detectors near to the refueled channel. The agreement was good and in most of the cases within the measurement errors. Cell calculations were made with WIMS-D4, and reactor calculations with PUMA. a fuel management 3D diffusion program developed in Argentina. With SEU fuel with a greater burnup in the central high power core region, new operating procedures were developed to prevent PCI failures in fuel power ramps that arise during operation. Some fuel rod and structural assembly design changes were introduced on the

  4. The CNA-1 (Nuclear Power Plant Atucha-1) QK-01 repairing project

    International Nuclear Information System (INIS)

    Pizzaferri, J.C.; Cabot, P.

    1997-01-01

    The repair/maintenance of the CNA-1 QK-01 Moderator Cooler will be a leading case of the repair of a class 1 nuclear component in a high radiation environment; utilizing for the work, sophisticated remotely operated equipment. This paper describes the component, the repair-maintenance objective, and the equipment-procedures developed for the intervention. (author) [es

  5. Application of the weld in maintenance mechanics at the Nuclear Power Station Atucha I

    International Nuclear Information System (INIS)

    Cosentino, R.E.

    1988-01-01

    The application of the 'weld procedures', in the field of activity of nuclear power is an special chapter of weld. The so called 'Nuclear Installations' are actually under control from their contruction up to their life extension operation to special control programs and quality assurance. This situation obliges the implementation of procedures to assure the fulfilment of the programs for the need to make the reparations or mechanics construction. This paper describes the considerations that has been taken into account to repare some components of the plant. The works carried out constitute applications to the TIG weld procedure. The 'lip weld' is a mechanic component required in pressurized systems subject to air pressure. (Author)

  6. LDC nuclear power: Argentina

    International Nuclear Information System (INIS)

    Tweedale, D.L.

    1982-01-01

    Argentina's 31-year-old nuclear research and power program makes it a Third World leader and the preeminent Latin American country. Easily accessible uranium fuels the heavy water reactor, Atucha I, which provides 10% of the country's electric power. Atucha II and III are under construction. Several domestic and international factors combined to make Argentina's program succeed, but achieving fuel-cycle independence and the capacity to divert fissionable material to military uses is a cause for some concern. 60 references

  7. 37-Active rods fuel element for Atucha 1 nuclear power plant. Effects of this change in design over the neutronic behavior, decay power and radioactive inventory

    International Nuclear Information System (INIS)

    Villar, Javier E.

    1999-01-01

    The influence of the use of 37-rods fuel element on the behavior of the Atucha 1 nuclear power plant homogeneous core with slightly enriched fuel to 0.85 w % were studied through representative parameters such as average discharge burnup, channel powers, reactivity coefficients, kinetic parameters, radioactive inventory and decay power. In general, the values of mentioned parameters are similar to those corresponding to a core with the 36-rods fuel element actually in use, although it must be emphasized a decrease both in linear power and, in minor degree, in the efficiency of shut-off and control rods and a slight increase in the discharge burnup. The fuel management strategy developed for a core with 36-rods elements can be maintained. (author)

  8. Blackout sequence modeling for Atucha-I with MARCH3 code

    International Nuclear Information System (INIS)

    Baron, J.; Bastianelli, B.

    1997-01-01

    The modeling of a blackout sequence in Atucha I nuclear power plant is presented in this paper, as a preliminary phase for a level II probabilistic safety assessment. Such sequence is analyzed with the code MARCH3 from STCP (Source Term Code Package), based on a specific model developed for Atucha, that takes into accounts it peculiarities. The analysis includes all the severe accident phases, from the initial transient (loss of heat sink), loss of coolant through the safety valves, core uncovered, heatup, metal-water reaction, melting and relocation, heatup and failure of the pressure vessel, core-concrete interaction in the reactor cavity, heatup and failure of the containment building (multi-compartmented) due to quasi-static overpressurization. The results obtained permit to visualize the time sequence of these events, as well as provide the basis for source term studies. (author) [es

  9. Permission of change of limits in the vapor generators of the Atucha I Nuclear Central; Permiso de cambio de limites en los GVs de la CNA-I

    Energy Technology Data Exchange (ETDEWEB)

    Ventura, M. [Autoridad Regulatoria Nuclear, Av. Libertador 8250 (1429), Capital Federal (Argentina)]. e-mail: mventura@sede.arn.gov.ar

    2006-07-01

    In the mark of the modification of the Atucha-I Nuclear Central Installation (CNA-I) as consequence of the Introduction of the System 'Second Drain of Heat' (SSC), the Entity Responsible for the CNA-I (NASA) requested authorization to the Nuclear Regulatory Authority (ARN) to modify the value of the minimum level of water in the secondary side in the Steam generators (GVs) to activate the signal 'shoot of the Cut of the Reactor' (RESA-LLV). As the level in the GVs is one of those parameters that are used to shoot the Emergency Feeding System (RX), component of the SSC System, also was analyzed the change in the activation of the shoot signal of the 'Second Drain of Heat' (2SSC-LLV). The ARN uses for the study of the nuclear safety of nuclear power plants, the series of prediction programs RELAP5/MOD3.X. It participates of the evaluation and maintenance activities of these codes through specific agreements with the U.S. Nuclear Regulatory Commission (US-NRC). It is necessary to account with programs of this type since the ARN it licenses the construction and operation of Nuclear Power Plants (NPPs) and other outstanding facilities and it inquires its operation according to its own standards. With these tools its are auditing the calculations that the Responsible Entities of the operation make to guarantee the operability of the NPPs assisting the mentioned standards. The analysis with computational codes is used as a tool to achieve the best understanding in the behavior of the plant in union with the engineering approach, the manual calculations, the data analysis and the experience in the operation of the machine. (Author)

  10. Manufacturing at industrial level of UO2 pellets for the fuel elements of the Atucha I Nuclear Power Plant

    International Nuclear Information System (INIS)

    Dyment, I.G.; Noguera Rojas, Francisco

    1982-01-01

    The interest to produce fuel elements within a policy of self sufficiency arose with the installation of Atucha I. The first steps towards this goal consisted in processing the uranium oxide, transforming it into fuel pellets of high density. The developments towards the fabrication of said pellets, performed by CNEA since 1968, first at a laboratory level and afterwards on an industrial scale, allowed CNEA to obtain its own technological capability to produce 400 kg of UO 2 per day. The fuel pellets manufacturing method developed by CNEA is a powder-metallurgical process, which, besides conventional equipment, involves the use of special equipment that required the performance of systematic testing programmes, as well as special training at operational level. The developed processes respond to a modern and advanced technology. A general scheme of the process, starting with a directly sinterable UO 2 powder, is described, including compacting of the powder into pellets, sintering, control of the temperature in the sintering and reduction zones and of the time of permanence in both zones, and cylindric rectifying of the pellets. During the whole process, specialized personnel controls the operations, after which the material is released by the Quality Control Department. The national contribution to the manufacturing technology of the pellets for fuel elements of power and research reactors was of 100%. (M.E.L.) [es

  11. Chemical and radiochemical control of the primary circuit of Atucha INPP (Nuclear Power Plant) since the start up in January 1990

    International Nuclear Information System (INIS)

    Ali, S.P; Baungartner, E.C.; Blesa, M.A.

    1990-01-01

    Since the start up of Atucha I Nuclear Power Plant in January 1990, an exhaustive chemical and radiochemical control of primary media was undertaken. The main objectives were the evaluation of the water condition after the long outage and the determination of activity measurements limitations to detect and localize fuel failures. Chemical and radiochemical techniques were critically proved. At the same time, a complete program of updating and optimization of those procedures was developed, including the revision of the analytical parameters, range of applicability and accuracy. A more adequate processing of data was adopted. They were compared with historical values corresponding to periods with and without fuel elements failures, used as references. The analysis of theoretical models of total gamma activity concentration and some specific radionuclides activity concentration evolution and their rates, and the comparison with experimental data obtained during normal operation including some failure events, generated tables of alarm criteria through a combination of parameters. Additionally, actions are suggested for different combination of parameters. Operative conditions that might interfere in the detection and localization of a failed fuel element are also pointed out. (Author)

  12. Simulation with the MELCOR code of two severe accident sequences, Station Blackout and Small Break LOCA, for the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Valle Cepero, Reinaldo

    2004-01-01

    The results of the PSA-I applied to the Atucha I nuclear power plant (CNA I) determine the accidental sequences with the most influence related to the probability of the core reactor damage. Among those sequences are include, the Station Blackout and lost of primary coolant, combine with the failure of the emergency injection systems by pipe breaks of diameters between DN100 - DN25 or equivalent areas, Small LOCA. This paper has the objective to model and analyze the behavior of the primary circuit and the pressure vessel during the evolution of those two accidental sequences. It presented a detailed analysis of the main phenomena that occur from the initial moment of the accident to the failure moment of the pressure vessel and the melt material fall to the reactor cavity. Two sequences were taken into account, considering the main phenomena (core uncover, heating, fuel element oxidation, hydrogen generation, degradation and relocation of the melt material, failure of the support structures, etc.) and the time of occurrence, of those events will be different, if it is considered that both sequences will be developed in different scenarios. One case is an accident with the primary circuit to a high pressure (Station Blackout scenario) and the other with a early primary circuit depressurization due to the lost of primary coolant. For this work the MELCOR 1.8.5 code was used and it allows within a unified framework to modeling an extensive spectrum of phenomenology associated with the severe accidents. (author)

  13. Methodology and Software for Gross Defect Detection of Spent Nuclear Fuel at the Atucha-I Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sitaraman, Shivakumar; Ham, Young S.; Gharibyan, Narek [Lawrence Livermore National Laboratory, 7000 East Avenue, Livermore, CA 94550 (United States); Peixoto, Orpet J.M. [Brazilian-Argentine Agency for Accounting and Control of Nuclear Materials, Avenida Rio Branco, 123/Grupo 515- Centro, CEP: 20040-005, Rio de Janeiro (Brazil); Diaz, Gustavo [National Regulatory Authority - Argentina, Av. Del Libertador 8250, (1429) Buenos Aires (Argentina)

    2015-07-01

    At the Atucha-I pressurized heavy water reactor in Argentina, fuel assemblies in the spent fuel pools are stored by suspending them in two vertically stacked layers. This introduces the unique problem of verifying the presence of fuel in either layer without physically moving the fuel assemblies. Since much of the fuel is very old, Cerenkov viewing devices are often not very useful even for the top layer. Given that the facility uses both natural uranium and slightly enriched uranium at 0.85 w% {sup 235}U, and has been in operation since 1974, a wide range of burnups and cooling times can exist in any given pool. A spent fuel neutron counting tool consisting of a fission chamber, SFNC, has been used at the site to verify the presence of fuel up to burnups of 8000 MWd/t. At higher discharge burnups to levels up 11,000 MWd/t, the existing signal processing software of the tool was found to fail due to non-linearity of the source term with burnup. A new Graphical User Interface software package based on the LabVIEW platform was developed to predict expected neutron signals covering all ranges of burnups and cooling times and establish maps of expected signals at various pool locations. The algorithm employed in the software uses a set of transfer functions in a 47-energy group structure which are coupled with a 47-energy group neutron source spectrum based on various cooling times and burnups for each of the two enrichment levels. The database of the software consists of these transfer functions for the three different inter-assembly pitches that the fuel is stored in at the site. The transfer functions were developed for a 6 by 6 matrix of fuel assemblies with the detector placed at the center surrounded by four near neighbors, eight next nearest neighbors and so on for the 36 assemblies. These calculations were performed using Monte Carlo radiation transport methods. The basic methodology consisted of starting sources in each of the assemblies and tallying the

  14. Study of the Atucha I nuclear power plant's residual heat removal system unavailability through the fault tree analysis and common cause failures

    International Nuclear Information System (INIS)

    Terrado, C.A.

    1991-06-01

    The present essay offers a comprehensive research of the Atucha I nuclear power plant's residual heat removal system unavailability, including Fault Tree Analysis and Common Cause Failures (CCF) treatment. The study is developed within the Event Tree perspective that considers the loss of external electrical power of the initiating event. The event was constructed by the Safety Evaluations Division of the Ezeiza Atomic Center in Argentina. According to the Event Tree, the research includes system demand during plant operation with 132 KV and emergency generation (Diesel motor generators). The system unavailability assessment is approached in two different ways: a) Considering independent failures only. b) Taking into account the existence of Common Cause Events, and modeling dependent failures. The Fault Tree quantification is played using the AIEA PSAPACK Code. The assessment data base is compiled from plant specific records and generic data bases like TECDOC 478. After Fault Tree model logic development, some general procedures used in common cause failures treating are applied to pick up another set of solutions. The results of the study are: a) Four Fault Trees have been developed to model the abovementioned system: 132 KV and emergency generation, both including and excluding CCF. b) The following unavailability values were obtained: 132 KV independent failures only: 7 10 -4 . Emergency generation independent failures only: 1.53 10 -2 . 132 KV dependent and independent failures: 3.6 10 -3 . Emergency generation dependent and independent failures: 1.74 10 -2 . The major conclusions obtained from the precedent results are: a) When using 132 KV system configuration, minimal cut sets involving common cause failures represents 81%from total system unavailability. b) The dependent failures treatment is an important task to be considered in safety assessments in order to reach more realistic values. (Author) [es

  15. Knowledge preservation of Atucha type reactor: Practical approaches and lessons learned

    International Nuclear Information System (INIS)

    Eppenstein, M.; Vetere, C.

    2004-01-01

    As Siemens, designer of the Atucha type HWRs, has transferred its nuclear activities to Framatome ANP, Argentina must undertake the knowledge preservation of this type of reactor, both if life extension is decided for the operating Atucha I NPP, and/or if ending the construction of Atucha II is decided. Another reason for undertaking a knowledge preservation program is aging and increasing retirement of personnel in the nuclear field, and the small number of young people in nuclear related disciplines at the universities. This situation motivated CNEA to implement a Knowledge Management system (KM), in order to capture and capitalize the tacit and explicit knowledge, to spread it and share it, making use of the technical and suitable tools through the organization. The strategy was based on recognizing the critical knowledge by means of a methodology incorporating the critical knowledge map technique. This map is a tool that uses the cognitive surfing in order to access to the organization heritage knowledge. Different techniques and methodologies are applied for identifying the critical knowledge domains. The result is a vigorous graphical tool with a certain formalism, able to describe knowledge in a hierarchical way in order to preserve it. It is used to analyze the criticality, and as an access portal to the knowledge patrimony, pointing out, according to each knowledge area, the people skills, publications, related documents and others. Experience obtained through the KM system development shows how training techniques are put into practice, in order not to interfere with normal plant operation, and how to initiate the KM processes in order to improve the criticity. (author)

  16. Some conclusions obtained from the thermo-hydraulic behavior analysis of the nuclear power plant Atucha I, in case of loss of coolant accident with second heat sink

    International Nuclear Information System (INIS)

    Ventura, Mirta A.

    2003-01-01

    This paper is based on the recompilation, analysis and elaboration of the results of the operator (NA-SA), in the framework of the Atucha I Second Heat Sink project. The results have been compared with those obtained for the same power plant without second heat sink. The conclusions of the work permit the establishment of the operation rules of the plant. (author)

  17. ATUCHA I NPP - Emergency drill practice

    International Nuclear Information System (INIS)

    Sanda, Alejandro; Rosales, Gabriel

    2008-01-01

    Full text: Atucha I NPP performs an Emergency Drill Practice once a year. Its main goals are: -) Fulfill the requirements of the Argentine Nuclear Regulatory Authority (ARN) regarding Atucha I NPP's Operating License; -) Fulfill the commitment with the community regarding the safe and reliable operation Atucha I NPP; -) Verify the response of the Civil Organizations, Security Forces, and Armed Forces, as well as the correct application of the Emergency Plan; -) Perform the 'General Alarm Drill' periodic control; -) Perform a re-training of the members of the Security Advisor Internal Committee (CIAS) on the Internal and External Aspects of the Emergency Plan and on the related procedures; -) Test the Emergency Communications System. New goals are added every year, considering the Drill's scope. This drill comprises two different kinds of practices: Internal practices (practices in the station, with our personnel) and external practices (practices outside the station with governmental organizations). Internal practices comprise: -) Internal and external communications practices; -) Acoustic alarms; -) Personnel gathering in the Meeting Points; -) Safety of selected Meeting Points; -) Personnel count, selective evacuation; -) Iodide Potassium pills distribution; -) CICE (Internal Group for Emergency Control) Coordination. External practices comprise: -) Nuclear Regulatory Authority; -) Argentine Navy, Comando Area Naval Fluvial, Base Naval Zarate; -) Lima firemen; -) Zarate firemen; -) Municipal Civil Defense (Zarate and Lima); -) National Guard, Escuadron Atucha; -) Zarate Regional Hospital; -) Lima Police Department; -) Zarate Police Department; -) Argentine Coast Guard, Zarate; -) Local radios: Radio FM Libre, FM El Sitio; -) First Aid clinic. The following activities are performed together with the aforementioned organizations: -) Formation of an 'Operative committee'; -) Evacuation of citizens in a 3 km radio; -) Control of every access to Lima; -) Control of

  18. Extended burnup with SEU fuel in Atucha-1 NPP

    International Nuclear Information System (INIS)

    Alvarez, L.; Casario, J.; Fink, J.; Perez, R.; Higa, M.

    2002-01-01

    Atucha-1 is a Pressurized Heavy Water Reactor originally fuelled with natural uranium. Fuel Assemblies consist of 36 fuel rods and the active length is 5300 mm. The total length of the fuel assembly is about 6 m. The average discharge burnup of natural UO 2 fuel is 5900 MWd/tU. After the deregulation of the Argentine electricity market there was an important incentive to reduce the impact of fuel cost on the cost of generation. To keep the competitiveness of the nuclear energy against another sources of electricity it was necessary to reduce the cost of the nuclear fuel. With this objective a program to introduce SEU (0.85 % 235 U) fuel in Atucha-1 was launched in 1993. As a result of this program the average SEU fuel discharge burnup increased to more than 11000 MWd/tU. The first SEU fuels were introduced in Atucha-1 in 1995 and, in the present stage of the program, 71% of core positions are loaded with this type of fuel. This paper describes key aspects of Atucha-1 fuel design and their relevance limiting the burnup extension and shows relevant data regarding the SEU in-reactor performance. At the present time 125 SEU Fuel Assemblies have been irradiated without failures associated with the extended burnup or unfavorable influences on the operation of the power station. (author)

  19. Full-scale vibration tests of Atucha II N.P.P. Part I: objectives, instrumentation and test description

    International Nuclear Information System (INIS)

    Konno, T.; Tsugawa, T.; Sala, G.; Friebe, T.M.; Prato, C.A.; Godoy, A.R.

    1995-01-01

    The main purpose of the tests was to provide experimental data on the dynamic characteristics of the main reactor building and adjacent structures of a full-scale nuclear power plant built on deep Quaternary soil deposits. Test results were intended to provide a benchmark case for control and calibration of state-of-the-art numerical techniques used for engineering design of new plants and assessment of existing facilities. Interpretation of test results and calibration of numerical analyses are described in other associated papers. (author). 5 figs

  20. Nuclear physics II

    International Nuclear Information System (INIS)

    Elze, T.

    1988-01-01

    This script consisting of two parts contains the matter of the courses Nuclear Pyhsics I and II, as they were presented in the winter term 1987/88 and summer term 1988 for students of physics at Frankfurt University. In the present part II the matter of the summer term is summarized. (orig.) [de

  1. Mechanical and Radiological Characterization of Different parts of an Irradiation Coolant Channel Tube from Atucha I Nuclear Plant; Caracterizacion Mecanica y Radiologica de Partes de Canales Refrigerantes Irradiados Extraidos del Reactor de la Central Nuclear Atucha I

    Energy Technology Data Exchange (ETDEWEB)

    Piquin, Ruben [Instituto Balseiro, Universidad Nacional de Cuyo, Centro Atomico Bariloche, Universidad Nacional de Mar del Plata (Argentina)

    2001-07-01

    The widespread replacement of reactor internals has generated a substantial volume of active material. It is essential to work with these components at least in a partial way before the next planned stop, which will take place during the second semester of the year 2002. Due to the fact that the reactor internals pool and the storage pool for irradiated nuclear fuel have limited capacities, it has been proposed to compact an experimental shift of 50 irradiated coolant channels, that are currently placed in storage pools. Basically the processed waste will be put in baskets at the bottom pools.The alternative choice proposes to divide an irradiation coolant channel tube into different parts: stainless steel section, zircaloy-4 section and stainless steel section with hardened zones with cobalt alloys named Estelite-6. The person in charge has already planned the constructive and operative solutions but the mechanical characterization of the different parts of the channel tube is necessary in order to dimension the compaction tool needed for the semi-industrial installation.In the present special report, two well-differentiated actions will be described. The necessary compacted strength of the irradiation coolant channel tube will be estimated for the stainless steel section and the zircaloy-4 section starting from experiment with unirradiated material and considering effects of radiation damage and hydrides on the ductility.These results will be used to design the necessary compacted tools for the semi-industrial installation. The necessary equipment for the radiological characterization of the different material sections already specified will be described and the most important emitting particles of radiation that could be detected will be mentioned. Also the decontamination process to use including the radiological characterization of every stage of the process will be described in order to establish the decontamination factor. Finally the most important

  2. Mechanical and Radiological Characterization of Different parts of an Irradiation Coolant Channel Tube from Atucha I Nuclear Plant

    International Nuclear Information System (INIS)

    Piquin, Ruben

    2001-01-01

    The widespread replacement of reactor internals has generated a substantial volume of active material. It is essential to work with these components at least in a partial way before the next planned stop, which will take place during the second semester of the year 2002. Due to the fact that the reactor internals pool and the storage pool for irradiated nuclear fuel have limited capacities, it has been proposed to compact an experimental shift of 50 irradiated coolant channels, that are currently placed in storage pools. Basically the processed waste will be put in baskets at the bottom pools.The alternative choice proposes to divide an irradiation coolant channel tube into different parts: stainless steel section, zircaloy-4 section and stainless steel section with hardened zones with cobalt alloys named Estelite-6. The person in charge has already planned the constructive and operative solutions but the mechanical characterization of the different parts of the channel tube is necessary in order to dimension the compaction tool needed for the semi-industrial installation.In the present special report, two well-differentiated actions will be described. The necessary compacted strength of the irradiation coolant channel tube will be estimated for the stainless steel section and the zircaloy-4 section starting from experiment with unirradiated material and considering effects of radiation damage and hydrides on the ductility.These results will be used to design the necessary compacted tools for the semi-industrial installation. The necessary equipment for the radiological characterization of the different material sections already specified will be described and the most important emitting particles of radiation that could be detected will be mentioned. Also the decontamination process to use including the radiological characterization of every stage of the process will be described in order to establish the decontamination factor. Finally the most important

  3. Fracture mechanical analysis of relevant transients in the pressure vessel of Atucha I reactor

    International Nuclear Information System (INIS)

    Saavedra, Fernando M.

    2001-01-01

    The evolution of the applied stress intensity factor K I for 10 relevant transients of the nuclear power station Atucha I obtained from thermohydraulic data is analyzed according to the methodology proposed in Section XI of ASME Boiler and Pressure Vessel Code. Vast knowledge was thus obtained about basic concepts of fracture mechanics and its application to remanent life of nuclear components. Basic knowledge which commands the performance of nuclear power stations was also obtained, especially that related to the Atucha I utility [es

  4. Distribution of equilibrium burnup for an homogeneous core with fuel elements of slightly enriched uranium (0.85% U-235) at Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Sidelnik, J.I.; Perez, R.A.; Salom, G.F.

    1987-01-01

    At Atucha I, the present fuel management with natural uranium comprises three burnup areas and one irradiation path, sometimes performing four steps in the reactor core, according to the requirements. The discharge burnup is 6.0 Mw d/kg U for a waste reactivity of 6.5 m k and a heavy water purity of 99.75%. This is a preliminary study to obtain the distribution of equilibrium burnup of an homogeneous core with slightly enriched uranium (0.85% by weight U-235), using the time-averaged method implemented in the code PUMA and a representative model of one third of core and fixed rod position. It was found a strategy of three areas and two paths that agrees with the present limits of channel power and specific power in fuel rod. The discharge burnup obtained is 11.6 Mw d/kg U. This strategy is calculated with the same method and a full core representation model is used to verify the obtained results. (Author)

  5. A challenging task: cleaning and repairing at nuclear power plant ATUCHA I (CAN-I) the primary's moderators cooling circuits heat exchangers

    International Nuclear Information System (INIS)

    Amaya, D.; Alaniz, A.; Bernasconi, R.

    2005-01-01

    A set of automatic and semiautomatic machines and tools were designed for the accomplishment of remotely controlled works in high radiation fields on the ATUCHA-I moderator heat exchangers. The object of this equipment is to carry out works related to the cleaning, inspection and eventual blocking of the heat exchanger's tubes. Due to the special characteristics of the area, such as difficult access, not much space and high dose rates, the remote operation of highly trained and specialized personnel with specially designed tools, is mandatory. The Principal operations consist of: 1. Equipment manually taken to the area by specialized personnel. 2. Remote cutting the bolting and cutting and re-weld seals with custom designed equipment. 3. Remote cutting and re-weld piping connections with equipment on customized tracks, special supports, drives and commands. 4. Remote cleaning, leak testing, machining and plugging of the tube sheets with a custom-made master-slave/Cartesian robotic manipulators. 5. Monitoring with video cameras and lighting systems incorporated into the equipment. 6. Ands others task as piping stabilization, supporting and moving flanges, re-alignment of seals and pipes, etc. This paper describes the entire development of this project, starting from the initial work plan to the completion of the first on-site work carried out at the facility. Including descriptions, drawings and pictures of the custom designed equipment, description of the performed works and comparisons between the actual doses and estimated manual operation doses. (authors)

  6. Cost estimation of interim dry storage for Atucha I NPP

    International Nuclear Information System (INIS)

    Bergallo, Juan E.; Fuenzalida Troyano, Carlos S.

    2007-01-01

    A joint effort between NASA and CNEA has been performed in order to evaluate and fix the strategy of interim spent fuel storage for Atucha I nuclear power plant. In this work the cost estimation on the proposed system was performed in order to fix the parameter and design criteria for the next engineering step. The main results achieved show that both alternatives are all in the same range of costs per unit of mass to be stored, the impact on electricity cost is less than 1 US mills/KWh and the scaling factor achieved is 0.85. (author) [es

  7. Mixed wastes treatment in Atucha I

    International Nuclear Information System (INIS)

    Varani, J.L.; Comandu, J.F.

    1998-01-01

    Full text: During decontamination works of the fueling machine of Atucha I nuclear power plant (AINPP), a liquid waste with special characteristics was generated, which needed the development of a treatment method. The waste consisted of an emulsion designed for the cleaning of mechanical components and was formed by an organic solvent dispersed in water with aid of an emulsifier additive. After several cleaning operations, the emulsion contained an important quantity of lubricants and radioactive dirt. The treatment had the objective of recycling a toxic waste such as the organic solvent and reducing the volume of the residual mass. Laboratory tests were made tending to the emulsion separation in their components. Ionic force and ionic mobility were modified for join the emulsion micelles and produce their coalescence. Different salts and working temperatures were tried and it was stated that the combination of 1% of Na 2 SO 4 added and 40 degree C temperature were the optimum taking into account the available equipment in AINPP and cost considerations. The process was carried out in batch mode and 3 residual streams were obtained, an aqueous one which was sent to Residual Water System of AINPP, an organic liquid consisting of decontaminated hydrocarbons, useful for other cleaning tasks and finally a solid one, sited in the in-between interface of the other two liquids, consisting of insoluble soaps used as lubricant thickness, containing the principal proportion of radioactivity. As a result of this process we have achieved a volume reduction higher than 90%, the recycling of the organic solvent and concentration of radioactivity in a solid greasy mass with low water solubility. (author) [es

  8. Utilization of noise analysis technique for mechanical vibrations estimation in the ATUCHA{sub 1} and Embalse Argentine NPP; Uso de la tecnica de analisis de ruido para la estimacion de vibraciones mecanicas en las centrales nucleares argentinas Atucha I y Embalse

    Energy Technology Data Exchange (ETDEWEB)

    Lescano, V.H.; Wentzeis, L.M. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Centro Atomico Constituyentes; Guevara, M.; Moreno, C. [Nucleoelectrica Argentina S.A., Cordoba (Argentina). Central Nuclear Embalse; Pineyro, J. [Nucleoelectrica Argentina S.A., Buenos Aires (Argentina). Central Nuclear Atucha I

    1996-07-01

    In Argentine, comprehensive noise measurements have been performed with the reactor instrumentation of the PHWR power plant Atucha I and Embalse. The Embalse reactor is a CANDU-600 (600 Mwe) type pressurized heavy water reactor. It's a heavy water moderator and heavy water cooled natural uranium fueled pressure tube system. Signal of vanadium and platinum type in core-self power neutron detectors of ex-core ion chambers and of a moderator pressure sensor have been recorded and analysed. The vibration of reactor internals as vertical and horizontal in-core neutron flux detectors units and the coolant channels systems, consisting of calandria and pressure tubes with fuel bundles, have been identified and monitored during normal reactor operation. Atucha I, is a PHWR reactor natural uranium fueled, and heavy water moderated and cooled. Neutron noise techniques using of ex-core ionization chambers and in-core Vanadium SPND's were implemented, among others, in order to produce early detection of anomalous vibrations in the reactor internals. Noise analysis was successfully performed to identify normal and peculiar vibrations in particular reactor internals. (author)

  9. How Belgium helped establish a surveillance programme for Argentina's Atucha-2

    Energy Technology Data Exchange (ETDEWEB)

    Mitev, Lubomir [NucNet, Brussels (Belgium)

    2015-02-15

    Collaboration between Belgian experts and Argentina on the commissioning of Argentine reactors helped overcome problems caused by delays with construction. Marc Scibetta, deputy manager for nuclear materials science, from the Belgian Nuclear Research Centre (SCK CEN) gave an interview in which he told some facts to Lubomir Mitev of NucNet. The cooperation between SCK CEN an Argentina's Comision Nacional de Energia Atomica (CNEA) started in 2002. The first project was a support for the safety evaluation of the Atucha-1 reactor pressure vessel. When Argentina resumed the construction of Atucha-2 in 2006 - originally, construction started in 1981 but was suspended in 1985 due to financial reasons -, SCK CEN was asked to develop and implement a surveillance programme for the unit.

  10. Modelling of blackout sequence at Atucha-1 using the MARCH3 code

    International Nuclear Information System (INIS)

    Baron, J.; Bastianelli, B.

    1997-01-01

    This paper presents the modelling of a complete blackout at the Atucha-1 NPP as preliminary phase for a Level II safety probabilistic analysis. The MARCH3 code of the STCP (Source Term Code Package) is used, based on a plant model made in accordance with particularities of the plant design. The analysis covers all the severe accident phases. The results allow to view the time sequence of the events, and provide the basis for source term studies. (author). 6 refs., 2 figs

  11. Reflections on the development of local suppliers for the Argentine nuclear industry

    International Nuclear Information System (INIS)

    Quilici, Domingo F.

    2008-01-01

    Argentina has given recently a new start to its nuclear power activities. Looking for background and experiences that can be useful under the new reality, the paper is a survey of the past development of local suppliers for the national nuclear industry. Based on the intention to answer the questions: Why so early it was decided to build a nuclear power plant? Why it was decided to buy it under a turnkey basis rather than developing an indigenous design? and what was the meaning of the 'opening of the technology package' at that time?, the paper describes the actions that led to the purchase of the Atucha I, Embalse and Atucha II nuclear power plants and how these decisions were implemented in order to maximize local participation and the technology transfer. It also analyzes the influence of the Argentine Nuclear Plan of the late seventies on the development of endogenous technology and describes the facts that helped to preserve until now the technological nuclear capabilities of the country in spite of the stopping of the Atucha II construction, and to create positive expectations regarding the revival of the local industry as a supplier of nuclear goods and services. (author) [es

  12. Regulatory actions towards dose reduction at Atucha 1 NPS

    International Nuclear Information System (INIS)

    Spano, F.; Curti, A.R.; Telleria, D.M.; Rudelli, M.D.

    1998-01-01

    Atucha 1, a nuclear power plant designed in the late sixties, is in commercial operation since June 1974. In some internal components such as the coolant channels, the station has Stellite-6, a high cobalt content alloy (up to 60%) for hard-facing application. The erosion and corrosion processes on the surfaces of the piping components of the primary coolant and moderator systems generate a varied type of particles oxides called 'crud'. The crud and cobalt 60 produced by neutron activation of cobalt are transferred by the water along the circuit of the coolant and moderator systems, producing deposits on internal surfaces. The cobalt deposits are dominant in radiation fields at working locations. For years, the Authority allowed a considerable number of station workers incurring doses near the limit since that installation had been built previously to introduction of the optimisation concept by ICRP publication 26. The recommendations included in ICRP publication 60 made more than difficult the radiological situation at the Atucha 1. For the facility, to comply with the new limit established by the Authority in January 1995, meant to carry out a substantial modification of the radiological conditions, specially the radiation fields due to cobalt 60. Some options to reduce individual and collective doses were analysed by the Authority. To carry out the evaluation of the deposit mechanisms and the real activity level of cobalt 60, a model of compartments connected by means of constant transfer coefficients was designed. It was concluded that there was a necessity to the change of coolant channels by new ones free of cobalt. It has been shown experimentally that radiation fields and occupational doses were reduced, due to the replacement programme carried out by the utility, in a similar way to the model predictions. At present after more than three years from the beginning of the application of the new limits, and after carrying out partially the tasks for the

  13. Nuclear Activities in Argentina, 2010

    International Nuclear Information System (INIS)

    Ferreri, J.C.; Ferreri, J.C.; Clausse, A.; Clausse, A.; Clausse, A.; Ordonez, J.P.; Mazzantini, O.A.

    2011-01-01

    Nuclear activities in Argentina are restarted. After almost two decades of near stagnation, the governments political decision of August 2006 regarding electrical energy production, considered the nuclear option as a valid one to solve the problems of the growing demand of electrical energy. This decision triggered again the activities related to the finalization of the third nuclear power reactor (Atucha-II), now actively progressing, the construction of a prototype of the CAREM integral advanced reactor, the life extension of the Embalse CANDU nuclear power plant (NPP) and the studies for the emplacement of a fourth NPP in an appropriate site. In all those years of near stagnation, there were notable exceptions related to the design and construction of experimental and radioisotope production reactors, led by INVAP, a state-owned industry, which exported its production. The accompanying industries of nuclear fuel elements production also remained active, given the demand of the two active NPPs. Meanwhile, the National Atomic Energy Commission of Argentina continued the efforts on research and development that were at the base of the technological achievements of the nuclear activities in Argentina. Nuclear safety studies associated with Atucha II and Embalse NPPs and radiological safety were also a substantive part of the continued efforts by Nucleo-Electrica de Argentina SA and the Nuclear Regulatory Authority of Argentina

  14. Atucha-I source terms for sequences initiated by transients

    International Nuclear Information System (INIS)

    Baron, J.; Bastianelli, B.

    1997-01-01

    The present work is part of an expected source terms study in the Atucha I nuclear power plant during severe accidents. From the accident sequences with a significant probability to produce core damage, those initiated by operational transients have been identified as the most relevant. These sequences have some common characteristics, in the sense that all of them resume in the opening of the primary system safety valves, and leave this path open for the coolant loss. In the case these sequences continue as severe accidents, the same path will be used for the release of the radionuclides, from the core, through the primary system and to the containment. Later in the severe accident sequence, the failure of the pressure vessel will occur, and the corium will fall inside the reactor cavity, interacting with the concrete. During these processes, more radioactive products will be released inside the containment. In the present work the severe accident simulation initiated by a blackout is performed, from the point of view of the phenomenology of the behavior of the radioactive products, as they are transported in the piping, during the core-concrete interactions, and inside the containment buildings until it failure. The final result is the source term into the atmosphere. (author) [es

  15. The power control system of the Siemens-KWU nuclear power station of the PWR [pressurized water reactors] type

    International Nuclear Information System (INIS)

    Huber, Horacio

    1989-01-01

    Starting with the first nuclear power plant constructed by Siemens AG of the pressurized light water reactor line (PWR), the Obrigheim Nuclear Power Plant (340 MWe net), until the recently constructed plants of 1300 MWe (named 'Konvoi'), the design of the power control system of the plant was continuously improved and optimized using the experience gained in the operation of the earlier generations of plants. The reactor power control system of the Siemens - KWU nuclear power plants is described. The features of this design and of the Siemens designed heavy water power plants (PHWR) Atucha I and Atucha II are mentioned. Curves showing the behaviour of the controlled variables during load changes obtained from plant tests are also shown. (Author) [es

  16. Nuclear India. Vol. II. [India's nuclear policy

    Energy Technology Data Exchange (ETDEWEB)

    Jain, J P

    1974-01-01

    The book contains 186 documents on India's nuclear policy covering a period from November 1948 to May 1974. It thus forms a comprehensive documentary account of India's nuclear policy. They include: texts of India's agreements for cooperation on the peaceful uses of atomic energy with the USA and Canada, the summary conclusions of India's atomic energy program for the decade 1970-80, the resolutions and amendments moved by India, the communications sent and the statements made by Indian representatives in various international forums--the conference of the IAEA statute, the Annual General Conference of the IAEA and its committees and the Board of Governors, the UN General Assembly and its First Committee, the conference of the Committee on Disarmaments etc. It also contains texts or extracts from the papers presented, statements made, and addresses and talks delivered by H. J. Bhabha, V. A. Sarabhai, H. N. Sethna and other eminent scientists at the international conferences on the peaceful uses of atomic energy, IAEA discussions on PNE, etc. Policy statements by India's Prime Ministers Nehru, Shastri and (Mrs.) Gandhi, and Foreign Ministers Chagla and Swaran Singh, made from time to time in the Lok Sabha and the Rajya Sabha--the two houses of the Indian parliaments--are also included. The sources of these documents are listed at the end. (MCB)

  17. Nuclear relevant installations licensing methodology in the Argentine Republic

    International Nuclear Information System (INIS)

    Paganini, C.E.

    1986-01-01

    A review of the requeriments of the Nuclear Installations Advisory Committee on Licensing (CALIN) from the nuclear security point of view, is presented. The methodology applied by the CALIN for the licensing in the Argentine Republic is included as well as codes, standards of applications and the interaction between the licensing Authority and the Responsible Entity during the whole process. Finally, the Atucha II nuclear power plant's licensing, in construction at present, is explained and the standard, of the licensing schedule, is presented graphically. (author) [es

  18. Nuclear medicine and thyroid disease - part II

    International Nuclear Information System (INIS)

    Chatterton, B.E.

    2005-01-01

    Part 1 of this article discussed the anatomy, physiology and basic pathology of the thyroid gland. Techniques of thyroid scanning and a few clinical examples are shown part II Copyright (2005) The Australian and New Zealand Society Of Nuclear Medicine Inc

  19. CARA Project: development of the advanced ULE fuel element for heavy water nuclear power plants

    International Nuclear Information System (INIS)

    Brasnarof, Daniel O.; Marino, Armando C.; Florido, Pablo C.; Munoz, C.; Bianchi, Daniel R.; Giorgis, Miguel A.

    2006-01-01

    The CARA Project (Spanish acronym of Combustible Avanzado para Reactores Argentinos) is a national fuel element technology development, compatible with our nuclear power plants (Atucha I, Embalse and Atucha II). It takes into account the experience obtained in our nuclear organisations (CNEA-CONUAR-NASA). The goal of the CARA fuel element is the performance improvement for those reactors and the enhancing of their normal operative conditions. The CARA design allows the burnup extension by using 52 rods of the same diameter. Likewise it keeps good thermo-hydraulic behaviour. The fuel bundle can be directly used in nuclear power plants with horizontal channels. By using an additional system it can be installed in the PHWR with vertical channels. The expected profits, by the use of the CARA in our reactors, broadly guaranty the recovery of the fund for its development, due to a reduction of the NPP fuels and back end cost. We estimate a reduction in the generation cost between 20 or 25 % in relation to the present one if we use 0.85 or 0.90% SEU (Slightly Enriched Uranium). The use of the CARA fuel in our reactors will also reduce the amount of spent fuel to be treated. The shortening could be between 17 to 27 % in Atucha I in relation to the present ULE (0.85%), between 38 to 46% for Embalse, and 45 to 53% for Atucha II. The mechanical behaviour and hydraulic compatibility have been verified. Several CARA prototypes were fabricated with a new design of the end plate and with new processes for the welding for the rods. We present in this paper the current status of the CARA fuel element development. (author) [es

  20. Comparison between a finite difference model (PUMA) and a finite element model (DELFIN) for simulation of the reactor of the atomic power plant of Atucha I; Comparacion entre un modelo de diferencias finitas (PUMA) y uno de elementos finitos (DELFIN) para la simulacion del reactor de la CNA-I (central nuclear Atucha-I)

    Energy Technology Data Exchange (ETDEWEB)

    Grant, C R [Comision Nacional de Energia Atomica, San Martin (Argentina). Unidad de Actividad Reactores y Centrales Nucleares

    1997-12-31

    The reactor code PUMA, developed in CNEA, simulates nuclear reactors discretizing space in finite difference elements. Core representation is performed by means a cylindrical mesh, but the reactor channels are arranged in an hexagonal lattice. That is why a mapping using volume intersections must be used. This spatial treatment is the reason of an overestimation of the control rod reactivity values, which must be adjusted modifying the incremental cross sections. Also, a not very good treatment of the continuity conditions between core and reflector leads to an overestimation of channel power of the peripherical fuel elements between 5 to 8 per cent. Another code, DELFIN, developed also in CNEA, treats the spatial discretization using heterogeneous finite elements, allowing a correct treatment of the continuity of fluxes and current among elements and a more realistic representation of the hexagonal lattice of the reactor. A comparison between results obtained using both methods in done in this paper. (author). 4 refs., 3 figs.

  1. Atucha II NPP (nuclear power plant). Analysis of the stress generated by special loads in the upper lateral support of the reactor pressure vessel

    International Nuclear Information System (INIS)

    Mancini, G.R.; Jaichenco, M.; Alvarez, L.M.

    1988-01-01

    This report is aimed at introducing the results of a study performed for assessing the mechanical behavior occurred after the introduction of the tangential component of stresses generated by accident-related loads in the RPV's support shield. Significant modifications have been made to the original structural design of the support on the basis of the results from such study, while taking into account the confinement effects produced by the joint action of an adequate steel reinforcing arrangement and of an external armoring plate. (Author) [es

  2. Assurance of sodium concentration measurement on line in water supply to the secondary system in Atucha I and II nuclear power plants

    International Nuclear Information System (INIS)

    Ormando, Miguel-Angel; Galarza, Guillermo-Dario

    2012-09-01

    Sodium measurement is used for quality control in high purity water application, to monitor break-through of mixed bed ion exchanger, condenser leaks and also to prevent caustic corrosion in turbines. The measurement principle is based on a selective electrode that responds to Nernst equation. The samples were measured in Swan's Trace Sodium/Conductivity Analyzer, Model 2114 July 1993. The aim of this work was to present a method in order to assure sodium concentration measurement on line using the ion selective method for water supply to the secondary system. Conductivity, less sensitive but more reliable than sodium analysis, is an overall quality parameter of water. It is traditionally used to back-up sodium analyzer and is sensitive to any ionic impurities. (authors)

  3. Desarrollo de proveedores para la industria nuclear argentina Visión desde las Centrales Nucleares

    Directory of Open Access Journals (Sweden)

    DOMINGO QUILICI

    2013-06-01

    Full Text Available Frente al inicio de una nueva etapa en la instalación de capacidad núcleo eléctrica en el país, se recorrerá la historia del desarrollo de la industria nuclear nacional (1964-1986 en búsqueda de antecedentes útiles para esta nueva realidad. Partiendo de la intención de dar repuesta a las preguntas: ¿Por qué se decidió tan tempranamente construir una central nuclear (en adelante CN; ¿por qué se decidió comprarla con una modalidad particular de los contratos “llave en mano”, en vez del desarrollo de una versión “criolla”? Y cuál fue el significado de la apertura del “paquete tecnológico” en aquel momento; se indagará sobre los antecedentes del desarrollo de proveedores para la industria nuclear en la Argentina. Se describirán las acciones que llevaron a la compra de las centrales de Atucha I, Embalse y Atucha II y como a partir de esas decisiones se implementaron políticas para maximizar la participación nacional en la construcción de las mismas y para la transferencia de tecnología del exterior hacia la industria local. Se analizará el Plan Nuclear puesto en vigencia a fines de los años setenta, desde el punto de vista de su influencia sobre el desarrollo tecnológico endógeno. Abstract The history of the development of national nuclear industry (1964-1986 will be reviewed in the search of useful patterns for the present new phase in the installation of nucleo-electric capacity in the country Precedents of development of suppliers for the argentinean nuclear industry will be considered, taking as starting point the following questions: Why the early decision of constructing a Nuclear Power Plant was taken? Why was it decided to buy it under a peculiar version of a turnkey contract instead of developing a “native” design? What were the implications of opening “technological packages” at that time? Actions leading to the construction of Atucha I, Embalse and Atucha II stations will be described, as well

  4. Analysis of the main causes of failures in the Atucha I PWR moderator circuit branch piping

    International Nuclear Information System (INIS)

    Porto, J.; Sarmiento, G.S.

    1983-01-01

    From 1977 to 1979 four through cracks were detected in the auxiliary connection of the moderator piping with the coolant circuit in the PWR Atucha I Nuclear Plant. The failures were observed to occur systematically in the same place of the pipe, where mechanical stresses were detected experimentally and thermal stresses were calculated based on temperature values measured on the pipe. The temperature field in steady state conditions as well as during thermal shocks was modelled by finite element codes, and the corresponding thermal stresses were than numerically calculated. Considering those thermal and mechanical solicitations, a crack propagation analysis based on the elastoplastic fracture mechanics and the finite element method is now being developed. Among other causes such as fatigue corrosion and vibrations, the results of the analysis show that the most preponderant factors determining the cracking are mechanical stress, thermal stress and thermal fatigue

  5. European Nuclear Decommissioning Training Facility II

    International Nuclear Information System (INIS)

    Demeulemeester, Y.

    2005-01-01

    SCK-CEN co-ordinates a project called European Nuclear Decommissioning Training Facility II (EUNDETRAF II) in the Sixth Framework Programme on Community activities in the field of research, technological development and demonstration for the period 2002 to 2006. This was a continuation of the FP5 project EUNDETRAF. EUNDETRAF II is a consortium of main European decommissioners, such as SCK-CEN, EWN (Energie Werke Nord, Greifswald Germany), Belgatom (Belgium), SOGIN Societa Gestione Impiantio Nucleari, Italy), Universitaet Hannover (Germany), RWE NUKEM (United Kingdom), DECOM Slovakia Slovakia), CEA Centre d'Energie Atomique, France), UKAEA (United Kingdom's Atomic Energy Agency, United Kingdom) and NRG (Nuclear Research and consultancy Group, Netherlands). The primary objective of this project is to bring together this vast skill base and experience; to consolidate it for easy assimilation and to transfer to future generations by organising a comprehensive training programme.Each training course has a one-week theoretical and a one-week practical component. The theoretical part is for a broader audience and consists of lectures covering all the main aspects of a decommissioning. The practical part of the course includes site visits and desk top solutions of anticipated decommissioning problems. Due to operational constraints and safety considerations, the number of participants to this part of the course is strictly limited. The partners intend to organise altogether two two-week EUNDETRAF II training courses over a period of three years. Another goal is to disseminate the existing theory as well as the practical know-how to personnel of the third countries. Finally it is important to bring together the principal decommissioning organisations undertaking various decommissioning activities. The project creates a forum for regular contacts to exchange information and experiences for mutual benefit of these organisations as well as to enhance skill base in Europe to

  6. Experiments with radioactive nuclear beams II

    International Nuclear Information System (INIS)

    Aguilera R, E.F.; Martinez Q, E.; Gomez C, A.; Lizcano C, D.; Garcia M, H.; Rosales M, P.

    2001-12-01

    The studies of nuclear reactions with heavy ions have been carried out for years for the group of heavy ions of the laboratory of the Accelerator of the ININ. Especially in the last years the group has intruded in the studies of nuclear reactions with radioactive beams, frontier theme at world level. Presently Technical Report is presented in detailed form the experimental methods and the analysis procedures of the research activities carried out by the group. The chpater II is dedicated to the procedures used in the analysis of the last two experiments with radioactive beams carried out by the group. In the chapter III is presented the procedure followed to carrying out an extended analysis with the CCDEF code, to consider the transfer channel of nucleons in the description of the fusion excitation functions of a good number of previously measured systems by the group. Finally, in the chapter IV the more important steps to continue in the study of the reaction 12 C + 12 C experiment drifted to be carried out using the available resources of the Tandem Accelerator Laboratory of the ININ are described. At the end of each chapter some of the more representative results obtained in the analysis are presented and emphasis on the scientific production generated by the group for each case is made. (Author)

  7. Instrumentation and control engineering at ENACE (Argentine Nuclear Enterprise of Electric Power Plants S.A.)

    International Nuclear Information System (INIS)

    Roca, J.L.; Garzon, D.

    1987-01-01

    This paper describes the techniques used in the project of instrumentation and control for the Atucha II nuclear power plant, from the original flow diagram of the system whose instrumentation and control is requested to the functional binary diagrams and control loops, through measurement sheets and other documentation. An account of the organization and handling of this mass of information is given, using an electronic processing system of data file for the project. A brief description of the task implied in the completing and updating of these files defines the scheme in which all the documentation development associated with a given process is included. (Author)

  8. Overview of the SEU project for extended burnup at the Atucha-I NPP. Four years of operating experience

    International Nuclear Information System (INIS)

    Fink, J.M.; Higa, M.; Perez, R.; Pineyro, J.; Sidelnik, J.; Casario, J.A.; Alvarez, L.

    2002-01-01

    Atucha I is a 357 MWe nuclear station moderated and cooled with heavy water, of German design located in Argentina. Fuelling is on-power and the plant was originally fuelled with natural uranium. To reduce fuel costs a program was initiated in August 1993 to introduce gradually slightly enriched uranium (SEU) fuel (0.85 w% U-235) with an associated burnup increase from 5900 MWd/tU to 11300 MWd/tU. The introduction of SEU fuel started in January 1995 and the program was divided in three Phases with an upper limit of SEU FA in the core: 12, 60 and 252 (full core) and licensing documentation was prepared for each Phase. This paper describes the most important aspects of the operating and project experience, and some factors limiting the burnup extension from an operation point of view. After four years of the program and with 181 SEU FA (71%) of the core, the operating experience has been good and without unfavourable effects due to the use of SEU fuel with the only exception of a small increase of the time to reach full power in plant startups or power cycling. In particular, the new criteria to prevent PCI failures in power ramps for higher burnup SEU fuel in refueling operations, plant startups or power cycling has been effective. The average discharge burnup of the SEU fuel taken out of the reactor in 1998 was 11263 MWd/tU. The average discharge burnup of the natural fuel in the same year was 6640 MWd/tU, with an increase of about 12% of the original value for a natural fuel core. The average number of fresh fuel assemblies per full power day was being reduced from 1.31 to 0.92 in 1998 and 0.83 in 1999. The fuel costs dropped gradually during the program from 9.38 (with natural uranium fuel) to 6.57 $/MWh in the first four months of 1999 (taking as reference the NU and SEU FA costs for 1999). Because of this the SEU program has been an important contribution to the reduction of Atucha I operating costs and to the competitiveness of nuclear power generation against

  9. The text of the agreement between the Agency and Argentina for the application of safeguards to the Atucha power reactor facility

    International Nuclear Information System (INIS)

    1995-01-01

    The Agreement between the Republic of Argentina, the Federative Republic of Brazil, the Brazilian-Argentine Agency for Accounting and Control of Nuclear Materials and the International Atomic Energy Agency for the Application of Safeguards came into force on 4 March 1994. As a result of the coming into force of the aforesaid Agreement for Argentina, the application of safeguards under the Agreement of 3 October 1972 between Argentina and the IAEA for the application of safeguards to the Atucha Power Reactor Facility has been suspended

  10. Nuclear proliferation. II. Monopoly or cartel

    International Nuclear Information System (INIS)

    Baker, S.J.

    1976-01-01

    Increasing competition between a growing number of nations exporting nuclear technology and recent exporting of full fuel-cycle facilities raise fears of nuclear proliferation and widespread nuclear weapons. As a result of the 1973 oil crisis, industrial nations seeking a share in the international nuclear market in order to protect their economic interests must also cooperate to protect these same interests from nuclear risks. Disagreement over the form of cooperation centers on the competing exporters' tactics of undercutting safeguards and political restrictions. Monopoly was never an option for even the United States. Government intervention in the international nuclear market in the form of subsidies and financial incentives is a more practical approach than a free market. A cartel arrangement is appropriate to nuclear energy in the sense of reducing economic uncertainties, but political objections would be strong and there would be some risk of independent nuclear development. As a strategy to forestall proliferation, however, the cartel can control exports of enrichment and reprocessing facilities and make it more expensive for nations to independently develop nuclear weapons. An enlargement of safeguards arrangements by nuclear suppliers will require nations to trade some of their economic interests in order to achieve international political objectives

  11. Nuclear emergency buildings of Asco and Vandellos II nuclear power plants; Centros alternativos de emergencias de las centrales nucleares de Asco y Vandellos II

    Energy Technology Data Exchange (ETDEWEB)

    Massuet, J.; Sabater, J.; Mirallas Esteban, S.

    2016-08-01

    The Nuclear Emergency Buildings sited at Asco and Vandellos II Nuclear Power Plants (NPP) are designed to safety manage emergencies in extreme situations, beyond the design basis of the Nuclear Power Plants. Designed in accordance with the requirements of the Spanish Nuclear Regulator (Consejo de Seguridad Nuclear-CSN) these buildings are ready to operate over a period of 72 hours without external assistance and ensure habitability for crews of 120 and 70 people respectively. This article describes the architectural conception, features and major systems of the Nuclear Emergency Buildings sited at Asco and Vandellos II. (Author)

  12. Experiments with radioactive nuclear beams II; Experimentos con haces nucleares radiactivos II

    Energy Technology Data Exchange (ETDEWEB)

    Aguilera R, E.F.; Martinez Q, E.; Gomez C, A.; Lizcano C, D.; Garcia M, H.; Rosales M, P. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2001-12-15

    The studies of nuclear reactions with heavy ions have been carried out for years for the group of heavy ions of the laboratory of the Accelerator of the ININ. Especially in the last years the group has intruded in the studies of nuclear reactions with radioactive beams, frontier theme at world level. Presently Technical Report is presented in detailed form the experimental methods and the analysis procedures of the research activities carried out by the group. The chpater II is dedicated to the procedures used in the analysis of the last two experiments with radioactive beams carried out by the group. In the chapter III is presented the procedure followed to carrying out an extended analysis with the CCDEF code, to consider the transfer channel of nucleons in the description of the fusion excitation functions of a good number of previously measured systems by the group. Finally, in the chapter IV the more important steps to continue in the study of the reaction {sup 12}C + {sup 12}C experiment drifted to be carried out using the available resources of the Tandem Accelerator Laboratory of the ININ are described. At the end of each chapter some of the more representative results obtained in the analysis are presented and emphasis on the scientific production generated by the group for each case is made. (Author)

  13. Energy paper II: Nuclear energy revival

    International Nuclear Information System (INIS)

    Anonymous

    2008-01-01

    ESI Energy paper is called 'Issue Paper' awarded by think-tank Energy Security Institute. The second issue focuses on the energy security of countries from the perspective of Renaissance of construction of nuclear power plants. Topicality is documented by fluctuations in fossil fuel prices on the world commodity markets and by extortionate potential, disposed by their main producers. The Slovak Republic is actively engaged into international dialogue on the need for the development of nuclear energy.

  14. Materials qualification for nuclear power plants

    International Nuclear Information System (INIS)

    Braconi, F.

    1987-01-01

    The supply of materials to be used in the fabrication of components submitted to pressure destined to Atucha II nuclear power plant must fulfill the quality assurance requirements in accordance with the international standards. With the aim of promoting the national participation in CNA II, ENACE had the need to adapt these requirements to the national industry conditions and to the availability of official entities' qualification and inspection. As a uniform and normalized assessment for the qualification of materials did not exist in the country, ENACE had to develop a materials suppliers qualification system. This paper presents a suppliers qualification procedure, its application limits and the alternative procedures for the acceptance of individual stock and for the stock materials purchase. (Author)

  15. Nuclear Fuel Cycle System Analysis (II)

    Energy Technology Data Exchange (ETDEWEB)

    Ko, Won Il; Kwon, Eun Ha; Yoon, Ji Sup; Park, Seong Won

    2007-04-15

    As a nation develops strategies that provide nuclear energy while meeting its various objectives, it must begin with identification of a fuel cycle option that can be best suitable for the country. For such a purpose, this paper takes four different fuel cycle options that are likely adopted by the Korean government, considering the current status of nuclear power generation and the 2nd Comprehensive Nuclear Energy Promotion Plan (CNEPP) - Once-through Cycle, DUPIC Recycle, Thermal Reactor Recycle and GEN-IV Recycle. The paper then evaluates each option in terms of sustainability, environment-friendliness, proliferation-resistance, economics and technologies. Like all the policy decision, however, a nuclear fuel cycle option can not be superior in all aspects of sustainability, environment-friendliness, proliferation-resistance, economics, technologies and so on, which makes the comparison of the options extremely complicated. Taking this into consideration, the paper analyzes all the four fuel cycle options using the Multi-Attribute Utility Theory (MAUT) and the Analytic Hierarchy Process (AHP), methods of Multi-Attribute Decision Making (MADM), that support systematical evaluation of the cases with multi- goals or criteria and that such goals are incompatible with each other. The analysis shows that the GEN-IV Recycle appears to be most competitive.

  16. Nuclear spectroscopy with direct relations II. Proceedings

    International Nuclear Information System (INIS)

    Throw, F. E.

    1964-01-01

    The Symposium on Nuclear Spectroscopy with Direct Reactions, sponsored and organized by Argonne National Laboratory under the auspices of the U. S. Atomic Energy Commission, was held on 9-11 March 1964 at the Center for Continuing Education, University of Chicago. The present volume contains the invited papers along with abstracts or summaries of the few short papers selected for their special relevance to the topics of the invited lecturers . Edited versions of the discussions are also included

  17. Nuclear instrumentation system operating experience and nuclear instrument testing in the EBR-II

    International Nuclear Information System (INIS)

    Yingling, G.E.; Curran, R.N.

    1980-01-01

    In March of 1972 three wide range nuclear channels were purchased from Gulf Atomics Corporation and installed in EBR-II as a test. The three channels were operated as a test until April 1975 when they became a permanent part of the reactor shutdown system. Also described are the activities involved in evaluating and qualifying neutron detectors for LMFBR applications. Included are descriptions of the ANL Components Technology Division Test Program and the EBR-II Nuclear Instrument Test Facilities (NITF) used for the in-reactor testing and a summary of program test results from EBR-II

  18. Moderator behaviour and reactor internals integrity at Atucha I NPP

    International Nuclear Information System (INIS)

    Berra, S.; Guala, M.; Herzovich, P.; Chocron, M.; Lorenzo, A.; Raffo Calderon, Ma. C. del; Urrutia, G.

    1996-01-01

    Atucha I is a Pressure Vessel Heavy Water Cooled Heavy Water Moderator Reactor. In this kind of reactor the moderator tank is physically connected to the primary coolant. Since neutron economy requires the moderator to be as cold as possible, it is necessary that even when physically connected, it should have a separated cooling system, which in this case is also used as a feed-water preheater, and also heat mass transfer with primary coolant should be minimized. This condition requires that some reactor internals are designed in principle to last the whole life of the plant. However, in 1988 the failure of one internal produced a 16 month shut down. This incident could have been prevented but the idea that reactor internals would not have failures due to aging was dominant at that time avoiding the early detection of the failure. However, the analysis of the records after the incident showed that some process variables had changed previously to the incident, i.e., power exchanged at the moderator heat exchanger had increased. Since the station restart up some changes in the moderator process variables and a flow rate reduction of about 10% through the primary side of one moderator cooler were observed. In order to understand the flow reduction and the overall behaviour of moderators parameters, two models were developed that predict moderator and moderator cooler behavior under the new conditions. The present paper refers to these models, which together with the improvement of process variables measurements mentioned in another paper presented at this meeting permits to understand current moderator behaviour and helps to early diagnostic of an eventual reactor internal failure. (author). 2 refs, 4 figs, 1 tab

  19. Moderator behaviour and reactor internals integrity at Atucha I NPP

    Energy Technology Data Exchange (ETDEWEB)

    Berra, S; Guala, M; Herzovich, P [Central Nuclear Atucha I, Nucleoelectrica Argentina, Lima, Buenos Aires (Argentina); Chocron, M; Lorenzo, A; Raffo Calderon, Ma. C. del; Urrutia, G [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Centro Atomico Constituyentes

    1997-12-31

    Atucha I is a Pressure Vessel Heavy Water Cooled Heavy Water Moderator Reactor. In this kind of reactor the moderator tank is physically connected to the primary coolant. Since neutron economy requires the moderator to be as cold as possible, it is necessary that even when physically connected, it should have a separated cooling system, which in this case is also used as a feed-water preheater, and also heat mass transfer with primary coolant should be minimized. This condition requires that some reactor internals are designed in principle to last the whole life of the plant. However, in 1988 the failure of one internal produced a 16 month shut down. This incident could have been prevented but the idea that reactor internals would not have failures due to aging was dominant at that time avoiding the early detection of the failure. However, the analysis of the records after the incident showed that some process variables had changed previously to the incident, i.e., power exchanged at the moderator heat exchanger had increased. Since the station restart up some changes in the moderator process variables and a flow rate reduction of about 10% through the primary side of one moderator cooler were observed. In order to understand the flow reduction and the overall behaviour of moderators parameters, two models were developed that predict moderator and moderator cooler behavior under the new conditions. The present paper refers to these models, which together with the improvement of process variables measurements mentioned in another paper presented at this meeting permits to understand current moderator behaviour and helps to early diagnostic of an eventual reactor internal failure. (author). 2 refs, 4 figs, 1 tab.

  20. Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation

    Energy Technology Data Exchange (ETDEWEB)

    Pecchia, M.; D' Auria, F. [San Piero A Grado Nuclear Research Group GRNSPG, Univ. of Pisa, via Diotisalvi, 2, 56122 - Pisa (Italy); Mazzantini, O. [Nucleo-electrica Argentina Societad Anonima NA-SA, Buenos Aires (Argentina)

    2012-07-01

    Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3D{sup C}/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI for performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)

  1. SAFEPAQ-II. User manual[Nuclear data

    Energy Technology Data Exchange (ETDEWEB)

    Forrest, R.A

    2001-03-01

    SAFEPAQ-II is the new software tool that has been developed to enable efficient production of the EAF nuclear data libraries that are required as input to the FISPACT activation code. It forms part of the European Activation System (EASY), and replaces SAFEPAQ and SYMPAL that were used previously. It enables all the nuclear data to be stored in relational databases (Access) and by using an interactive user interface allows the data to be viewed, modified, validated and then produced in the required EAF format as text files. It is written in Visual Basic and runs under the Windows NT4 and 98 operating systems. The Windows operating system has the great advantage of portability and SAFEPAQ-II has been successfully installed at two external sites for use by UKAEA's international collaborators. It has been used in the production of the EAF-2001 data libraries. (author)

  2. Nuclear Electronics II. Proceedings of the Conference on Nuclear Electronics. V. II

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1962-04-15

    Nuclear instruments are used in almost every phase of atomic energy work, from assessing health hazards and prospecting for nuclear materials to plant control and nuclear physics experiments. The demands on nucleonic instrumentation are growing steadily. High-energy particle physics need such instruments for measuring extremely short times; in various research experiments most advanced electronic systems are required; and routine applications of radioisotopes call for more reliable instruments for automated counting facilities. In order to give designers and users of nuclear instrumentation an opportunity to discuss the research results and to exchange information on recent developments and new designs, the International Atomic Energy Agency, in co-operation with the Federal Nuclear Energy Commission of Yugoslavia, organized a Conference on Nuclear Electronics which was held in Belgrade from 15-20 May 1961. It was attended by more than 300 scientists from nearly 30 countries and five international organizations. Over 150 papers were read and discussed. As the field of nuclear electronics has expanded considerably, it was impossible to discuss all aspects of nuclear electronics in one series of meetings. Included in the main topics were radiation detectors, electronic circuitry in conventional and fast-pulse techniques and advanced electronic systems used in nuclear research. The Proceedings presented in these volumes contain the full records of the Conference, including discussions. The present state of technique, together with current trends and developments, are outlined. Of particular value should be the world-wide survey on progress recently made in such fields as those connected with semiconductor detectors, spark counters, luminescence chambers and fast electronic facilities for nuclear physics research. Together with the Proceedings of the Symposium on the same subject held in Paris and also published by the International Atomic Energy Agency, these volumes

  3. Current situation and future of a nuclear power plan in Argentina

    International Nuclear Information System (INIS)

    Messi, E.

    2006-01-01

    After the crisis suffered by Argentina during the first years of the new millennium, the economic variables improved significantly, thus showing the need for an increase in the electric supply. While the growth rate of the electric demand has been of a steady 4%, the electric supply didn't come along as expected; mainly due to the decrease in the gas reserves, among other things. This scenario opens new opportunities for the nuclear area in our country: especially with Atucha II NPP termination that will contribute with 745 MWe gross output, together with Embalse NPP's life extension and refurbishment, and the analysis for the installation of new generation nuclear power plants in the decades to come, therefore strengthening the nuclear industry in Argentina. (author)

  4. TIBER II/ETR: Nuclear Performance Analysis Group Report

    International Nuclear Information System (INIS)

    1987-09-01

    A Nuclear Performance Analysis Group was formed to develop the nuclear technology mission of TIBER-II under the leadership of Argonne National Laboratory reporting to LLNL with major participation by the University of California - Los Angeles (test requirements, R and D needs, water-cooled test modules, neutronic tests). Additional key support was provided by GA Technologies (helium-cooled test modules), Hanford Engineering Development Laboratory (material-irradiation tests), Sandia National Laboratory - Albuquerque (high-heat-flux component tests), and the Idaho National Engineering Laboratory (safety tests). Support also was provided by Rennselaer Polytechnic Institute, Grumman Aerospace Corporation, and the Canadian Fusion Fuels Technology Program. This report discusses these areas and provides a schedule for their completion

  5. Improved nuclear gage development - phase i and ii. Interim report

    International Nuclear Information System (INIS)

    Chan, E.L.; Champion, F.C.; Castanon, D.R.; Chang, J.C.; Hannon, J.B.

    1976-09-01

    This report contains Phase I and II of an investigation covering the design and construction of a prototype nuclear-moisture-density backscatter gage. Gage development was based upon the analysis of several factors which affect gage performance. This research indicated that the prototype gage measurements are approximately equivalent to measurements obtained by a commercial transmission gage. The implication of this research finding concerns the qualification of the backscatter test method as a valid, reliable, and expedient procedure for determining in-situ soil conditions

  6. EMPIRE-II statistical model code for nuclear reaction calculations

    Energy Technology Data Exchange (ETDEWEB)

    Herman, M [International Atomic Energy Agency, Vienna (Austria)

    2001-12-15

    EMPIRE II is a nuclear reaction code, comprising various nuclear models, and designed for calculations in the broad range of energies and incident particles. A projectile can be any nucleon or Heavy Ion. The energy range starts just above the resonance region, in the case of neutron projectile, and extends up to few hundreds of MeV for Heavy Ion induced reactions. The code accounts for the major nuclear reaction mechanisms, such as optical model (SCATB), Multistep Direct (ORION + TRISTAN), NVWY Multistep Compound, and the full featured Hauser-Feshbach model. Heavy Ion fusion cross section can be calculated within the simplified coupled channels approach (CCFUS). A comprehensive library of input parameters covers nuclear masses, optical model parameters, ground state deformations, discrete levels and decay schemes, level densities, fission barriers (BARFIT), moments of inertia (MOMFIT), and {gamma}-ray strength functions. Effects of the dynamic deformation of a fast rotating nucleus can be taken into account in the calculations. The results can be converted into the ENDF-VI format using the accompanying code EMPEND. The package contains the full EXFOR library of experimental data. Relevant EXFOR entries are automatically retrieved during the calculations. Plots comparing experimental results with the calculated ones can be produced using X4TOC4 and PLOTC4 codes linked to the rest of the system through bash-shell (UNIX) scripts. The graphic user interface written in Tcl/Tk is provided. (author)

  7. Role of nuclear hexokinase II in DNA repair

    International Nuclear Information System (INIS)

    Khanna, S.; Bhatt, A.N.; Dwarakanath, B.S.; Kalaiarasan, P.; Brahmachari, V.

    2012-01-01

    A common signature of many cancer cells is a high glucose catabolic rate primarily due to the over expression of Type II hexokinase (HKII; responsible for the phosphorylation of glucose), generally known as cytosolic and mitochondrial bound enzyme that also suppresses cell death. Although, nuclear localization and transcriptional regulation of HKII has been reported in yeast; we and few others have recently demonstrated its nuclear localization in malignant cell lines. Interestingly, modification of a human glioma cell line (BMG-1) for enhancing glycolysis through mitochondrial respiration (OPMBMG cells) resulted in a higher nuclear localization of HKII as compared to the parental cells with concomitant increase in DNA repair and radio-resistance. Further, the glucose phosphorylation activity of the nuclear HKII was nearly 2 folds higher in the relatively more radioresistant HeLa cells (human cervical cancer cell line) as compared to MRC-5 cells (human normal lung fibroblast cell line). Therefore, we hypothesize that nuclear HKII facilitates DNA repair, in a hither to unknown mechanism, that may partly contribute to the enhanced resistance of highly glycolytic cells to radiation. Sequence alignment studies suggest that the isoenzymes, HKI and HKII share strong homology in the kinase active site, which is also found in few protein kinases. Interestingly HKI has been shown to phosphorylate H2A in-vitro. Further, in-silico protein-protein interaction data suggest that HKII can interact with several DNA repair proteins including ATM. Taken together; available experimental evidences as well as in-silico predictions strongly suggest that HKII may play a role in DNA repair by phosphorylation of certain DNA repair proteins. (author)

  8. A level III PSA for the inherently safe CAREM-25 nuclear power station

    International Nuclear Information System (INIS)

    Baron, Jorge H.; Nunez McLeod, J.; Rivera, S.S.

    2000-01-01

    A Level III PSA has been performed for the inherently safe CAREM-25 nuclear power station, as a requirement for licensing according to argentinian regulations. The CAREM-25 project is still at a detailed design state, therefore only internal events have been considered, and a representative site has been assumed for dose estimations. Several conservative hypothesis have been formulated, but even so an overall core melt frequency of 2.3E -5 per reactor year has been obtained. The risk estimations comply with the regulations. The risk values obtained are compared to the 700MW(e) nuclear power plant Atucha II PSA result, showing an effective risk reduction not only in the severe accident probability but alto in the consequence component of the risk estimation. (author)

  9. Study of essential safety features of a three-loop 1,000 MWe light water reactor (PWR) and a corresponding heavy water reactor (HWR) on the basis of the IAEA nuclear safety standards

    International Nuclear Information System (INIS)

    1989-02-01

    Based on the IAEA Standards, essential safety aspects of a three-loop pressurized water reactor (1,000 MWe) and a corresponding heavy water reactor were studied by the TUeV Baden e.V. in cooperation with the Gabinete de Proteccao e Seguranca Nuclear, a department of the Ministry which is responsible for Nuclear power plants in Portugal. As the fundamental principles of this study the design data for the light water reactor and the heavy water reactor provided in the safety analysis reports (KWU-SSAR for the 1,000 MWe PWR, KWU-PSAR Nuclear Power Plant ATUCHA II) are used. The assessment of the two different reactor types based on the IAEA Nuclear Safety Standards shows that the reactor plants designed according to the data given in the safety analysis reports of the plant manufacturer meet the design requirements laid down in the pertinent IAEA Standards. (orig.) [de

  10. Time series analysis of nuclear instrumentation in EBR-II

    International Nuclear Information System (INIS)

    Imel, G.R.

    1996-01-01

    Results of a time series analysis of the scaler count data from the 3 wide range nuclear detectors in the Experimental Breeder Reactor-II are presented. One of the channels was replaced, and it was desired to determine if there was any statistically significant change (ie, improvement) in the channel's response after the replacement. Data were collected from all 3 channels for 16-day periods before and after detector replacement. Time series analysis and statistical tests showed that there was no significant change after the detector replacement. Also, there were no statistically significant differences among the 3 channels, either before or after the replacement. Finally, it was determined that errors in the reactivity change inferred from subcritical count monitoring during fuel handling would be on the other of 20-30 cents for single count intervals

  11. Nuclear material inventory estimation in solvent extraction contractors II

    International Nuclear Information System (INIS)

    Beyerlein, A.

    1987-11-01

    The effectiveness of near-real-time nuclear materials accounting in reprocessing facilities can be limited by inventory variations in the separations contactors. Investigations are described in three areas: (i) Improvements in the model that the authors have described previously for the steady state inventory estimation in mixer-settler contactors, (ii) extension for the model for steady state inventory estimation to transient inventory estimation for non-steady state conditions, and (iii) the development of a computer model CUSEP (Clemson University Solvent Extraction Program) for simulating the concentration profiles and nuclear material inventories in pulsed column contactors. Improvements in the steady state model that are described in this report are the simplification of the methods for evaluating model parameters and development of methods for reducing the equation which estimates the total inventory of the set of contactors directly. The pulsed column computer model CUSEP (Clemson University Solvent Extraction Program) was developed. Concentration profiles and inventories calculated from CUSEP are compared with measured data from pilot scale contactors containing uranium. Excellent agreement between measured and simulated data for both the concentration profile and inventories is obtained, demonstrating that the program correctly predicts the concentration dispersion caused by pulsing and the dispersed phase holdup within the contactor. Further research to investigate (i) correction of the MUF (Material Unaccounted For) and CUMUF (Cumulative Material Unaccounted For) tests for mixer-settler contactor inventory using the simplified model developed in this work, (ii) development of a simple inventory estimation model for pulsed column contactors similar to that developed for mixer-settler contactors using CUSEP to provide necessary database, and (iii) sources of bias appearing in the MUF and CUMUF tests using computer simulation techniques are planned. Refs

  12. Hydraulic Design of the CARA Fuel Assembly for Atucha-I

    International Nuclear Information System (INIS)

    Juanico, Luis; Brasnarof, Daniel

    2000-01-01

    In this paper a hydraulic model of the CARA fuel assembly within the Atucha I fuel channel is developed. Besides, a experimental test running in the CBP low pressure loop have been designed.This model is used for design purpose of the assembly system such as the whole channel pressure drop remains the same that it is at the present.It is observed that choosing the right thickness and hole surface of the assembly system, it is possible tune up the CARA pressure drop, releases the azimuth alignment condition on the fuel element neighbors

  13. Study on alternatives of inertisation of nuclear power plant containment

    International Nuclear Information System (INIS)

    Baron, J.H.; Zarate, S.M.

    1998-01-01

    In the course of a severe accident in a nuclear power plant, the hydrogen generation and other flammable gases, during the core degradation phase and the interaction corium-concrete, could produce the failure of the containment by overpressure of by combustion. According to the analysis of the potential effects of hydrogen evolution, following accidents inside the containment trough a Defense-in depth principle, which attempts to assure that the containment must not fail catastrophically, two techniques have been evaluated: a: Inertisation pre-accident and b: Inertisation post-accident. The technique of inertisation pre-accident consists in replacing the air of the containment with inert-gas like nitrogen (N 2 ) or carbon dioxide (CO 2 ) during the normal operation. The inertisation post-accident in combination with early venting system consists in replacing the air of the containment with inert-gas like nitrogen (N 2 ) or carbon dioxide (CO 2 ), immediately after the beginning of the accident, while the radioactivity is still negligible inside the containment. A system of inertisation pre-accident with nitrogen is used on BWR Mark I and Mark II. Investigations on the inertisation post-accident of the containment atmosphere during severe accidents have been carried out with different objectives from principles of the decade of 1980. Studies concerning hydrogen problem for the nuclear power plants Atucha I and CAREM-25 have permitted to know that the hydrogen generation during an accidental sequence with core degradation, would result important, being able to arrive to form explosive mixtures. In the present work, the applicability of the techniques of inertisation is analyzed for the containment of the Atucha I and CAREM-25, considering the particular design characteristics of these plants. (author) [es

  14. Development and production of nuclear valves. Forging and welding. Pt. 2

    International Nuclear Information System (INIS)

    Bernal Castro, J.B.; Perez, J.C.; Labonia, R.N.

    1987-01-01

    The first part of this work deals with the obtainment of the austenitic stainless steel DIN 1.4541 (AISI 321) stabilized titanium for Atucha II nuclear valves. The second part presented herein, continued with the development process and part of the production of the bodies' forging and valves leads. This development has been also carried out in the country and a detailed set up of the process with its corresponding Inspection and Assay Program was needed. The last part of this stage has been initiated at the welding process, so it was necessary to develop specific welding procedures to qualify them and use the equipment specially applied to this requirement. The set of assays and criteria certification for the qualifications is presented. (Author)

  15. Security programs for Category I or II nuclear material or certain nuclear facilities. Regulatory guide G-274

    International Nuclear Information System (INIS)

    2003-03-01

    The purpose of this regulatory guide is to help applicants for a Canadian Nuclear Safety Commission (CNSC) licence in respect of Category I or II nuclear material - other than a licence to transport - , or a nuclear facility consisting of a nuclear reactor that may exceed 10 MW thermal power during normal operation, prepare and submit the security information to be included with the application, pursuant to the Nuclear Safety and Control Act (NSCA). Category I and II nuclear material are defined in Appendix B to this guide. This guide describes: the security information that should typically be included with the application for any licence referred to above; how the security information may be organized and presented in a separate document (hereinafter 'the security program description'), in order to assist CNSC review and processing of the application; and, the administrative procedures to be followed when preparing, submitting or revising the security program description. (author)

  16. Vibration monitoring of pressure vessel in Atucha-1 power plant

    International Nuclear Information System (INIS)

    Belinco, C.; Pastorini, A.; Martin Ghiselli, A.; Sacchi, M.

    1994-01-01

    The Vibration Monitoring Systems are described to obtain information about the mechanical state of different components in the main coolant system of nuclear power plants to ensure that changes in the mechanical integrity of this components are detected at an early point in time, even during operation. 9 figs

  17. Steam generator materials and secondary side water chemistry in nuclear power stations

    International Nuclear Information System (INIS)

    Rudelli, M.D.

    1979-04-01

    The main purpose of this work is to summarize the European and North American experiences regarding the materials used for the construction of the steam generators and their relative corrosion resistance considering the water chemestry control method. Reasons underlying decision for the adoption of Incoloy 800 as the material for the secondary steam generator system for Atucha I Nuclear Power Plant (Atucha Reactor) and Embalse de Rio III Nuclear Power Plant (Cordoba Reactor) are pointed out. Backup information taken into consideration for the decision of utilizing the All Volatil Treatment for the water chemistry control of the Cordoba Reactor is detailed. Also all the reasonswhich justify to continue with the congruent fosfatic method for the Atucha Reactor are analyzed. Some investigation objectives which would eventually permit the revision of the decisions taken on these subjects are proposed. (E.A.C.) [es

  18. The Text of the Agreement between the Agency and Argentina for the Application of Safeguards to the Atucha Power Reactor Facility

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1973-01-25

    The text of the Agreement between the Agency and the Government of the Republic of Argentina for the Application of Safeguards to the Atucha Power Reactor Facility is reproduced in this document for the information of all Members.

  19. Pokhran II and Beyond (Emerging Indian Nuclear Posture)

    National Research Council Canada - National Science Library

    Mishra, Jeetendra

    2002-01-01

    .... The nuclear forces, however, are sought only to be minimum possible to credibly deter nuclear weapons use or coercion against India, Considering the imperatives of the Indian deterrence posture...

  20. Proton nuclear magnetic resonance and spectrophotometric studies of nickel(II)-iron(II) hybrid hemoglobins

    International Nuclear Information System (INIS)

    Shibayama, N.; Inubushi, T.; Morimoto, H.; Yonetani, T.

    1987-01-01

    Ni(II)-Fe(II) hybrid hemoglobins, α(Fe) 2 β(Ni) 2 and α(Ni) 2 β(Fe) 2 , have been characterized by proton nuclear magnetic resonance with Ni(II) protoporphyrin IX (Ni-PP) incorporated in apoprotein, which serves as a permanent deoxyheme. α(Fe) 2 β(Ni) 2 , α(Ni) 2 β(Fe) 2 , and NiHb commonly show exchangeable proton resonances at 11 and 14 ppm, due to hydrogen-bonded protons in a deoxy-like structure. Upon binding of carbon monoxide (CO) to α(Fe) 2 β(Ni) 2 , these resonances disappear at pH 6.5 to pH 8.5. On the other hand, the complementary hybrid α(Ni) 2 β(Fe-CO) 2 showed the 11 and 14 ppm resonances at low pH. Upon raising pH, the intensities of both resonances are reduced, although these changes are not synchronized. Electronic absorption spectra and hyperfine-shifted proton resonances indicate that the ligation of CO in the β(Fe) subunits induced changes in the coordination and spin states of Ni-PP in the α subunits. In a deoxy-like structure, the coordination of Ni-PP in the α subunits is predominantly in a low-spin (S = 0) four-coordination state, whereas in an oxy-like structure the contribution of a high-spin (S = 1) five-coordination state markedly increased. Ni-PP in the β subunits always takes a high-spin five-coordination state regardless of solution conditions and the state of ligation in the partner α(Fe) subunits. In the β(Ni) subunits, a significant downfield shift of the proximal histidyl N/sub δ/H resonance and a change in the absorption spectrum of Ni-PP were detected, upon changing the quaternary structure of the hybrid. The chemical shifts were analyzed in terms of the E11-Val methyls vs. the porphyrin rings in hybrid Hbs

  1. Provisions relating to Nuclear Energy. II - International Conventions

    International Nuclear Information System (INIS)

    This book published by the Portuguese Junta de Energia Nuclear (Nuclear Energy Commission) reproduces in Portuguese and in the original language (English or French), texts of a series of international conventions in the nuclear field and the Statutes of international nuclear organisations and undertakings. The following are among the texts included: the Statutes of the IAEA, NEA, Eurochemic; the Euratom Treaty; the Tlatelolco Treaty; the co-operation agreement between Portugal and the United States on the peaceful uses of nuclear energy. (NEA) [fr

  2. Review of the KBS II plan for handling and final storage of unreprocessed spent nuclear fuel

    International Nuclear Information System (INIS)

    1980-01-01

    The Swedish utilities programme for disposal of spent nuclear fuel elements (KBS II) is summarized. Comments and criticism to the programme are given by experts from several foreign or international institutions. (L.E.)

  3. Current Status of World Nuclear Fuel Cycle Technology (II): Japan

    International Nuclear Information System (INIS)

    Jeong, Chang Joon; Ko, Won Il

    2007-06-01

    Japan needs to import around 80% of its energy requirements. In 1966, the first nuclear power plant began operation, nuclear energy has been a national strategic priority since 1973. Currently, 55 reactors provide around 30% of the country's electricity. Japanese energy policy has been conducted by the energy security and minimization of dependence of energy imports. The main factors regarding nuclear power are: - Continue to have nuclear power as a main factor of electricity production. - Recycle uranium and plutonium, and start domestic reprocessing from 2005. - Continue to develop fast breeder reactors to increase uranium utilization. - Promote the nuclear transparency to the public, emphasizing safety and non-proliferation. Also, the prospects of Asia's nuclear energy growth has been reviewed

  4. Aseismic design of the Heysham II Nuclear Power Station

    International Nuclear Information System (INIS)

    Day, J.W.

    1983-01-01

    A brief description of the seismic criteria established for use with the Steam Generating Heavy Water Reactor (SCHWR) and taken for the Heysham II Project is given. The qualification strategy adopted for Heysham II is described, and a brief overview is given of some of the more important design changes required for seismic purposes on that station

  5. Is nuclear energy safe for workers and the public. II

    International Nuclear Information System (INIS)

    Rasmussen, N.

    1976-01-01

    Dr. Rasmussen first examines the safety and economics records of nuclear power and finds them unassailable. Since these two items could not cause dissension, he observes that dissension evolves from safety of the power plant itself, the issue of diversion to use nuclear materials to make weapons, and from the disposal of wastes. He reviews how radioactivity can be released to cause an accident, and then cites figures comparing accidents each year in the U.S. with those projected for nuclear accidents. To date, not one person has been killed with the operation of a nuclear power plant. Dr. Rasmussen believes that taking advantage of nuclear power far outweighs its risks since ''we have seen what unemployment at the level of 9 or 10 percent has done in this country...to go without power that our industrial society needs to provide livelihood for our citizens will almost surely result in something much more dramatic than that.''

  6. Development and Application of MCNP5 and KENO-VI Monte Carlo Models for the Atucha-2 PHWR Analysis

    Directory of Open Access Journals (Sweden)

    M. Pecchia

    2011-01-01

    Full Text Available The geometrical complexity and the peculiarities of Atucha-2 PHWR require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Core models of Atucha-2 PHWR were developed using both MCNP5 and KENO-VI codes. The developed models were applied for calculating reactor criticality states at beginning of life, reactor cell constants, and control rods volumes. The last two applications were relevant for performing successive three dimensional neutron kinetic analyses since it was necessary to correctly evaluate the effect of each oblique control rod in each cell discretizing the reactor. These corrective factors were then applied to the cell cross sections calculated by the two-dimensional deterministic lattice physics code HELIOS. These results were implemented in the RELAP-3D model to perform safety analyses for the licensing process.

  7. Nuclear legislation analytical study. Regulatory and institutional framework for nuclear activities in OECD member countries. Volume II

    International Nuclear Information System (INIS)

    1984-01-01

    This study is part of a series of analytical studies of the major aspects of nuclear legislation in OECD Member countries and is published in two volumes. This volume II of the study is a revision and an expansion of a 1969 study concerning the organisation and general regime governing nuclear activities. The national studies were prepared, to the extent possible, following a standard plan for all countries to facilitate information retrieval and comparison. This volume also contains tables of international conventions of relevance to the nuclear field. (NEA) [fr

  8. Joining of Tungsten Cermet Nuclear Fuel, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — Nuclear Thermal Propulsion (NTP) has been identified as a critical technology needed for human missions to Mars and beyond due to its increased specific impulse...

  9. Construction and cost experience regarding the 2nd pool house for spent fuel storage facility in the Atucha Power Station

    International Nuclear Information System (INIS)

    Barbosa, C.A.

    1980-01-01

    The Atucha I second pool house storage for spent fuel is designed as an extension of the Atucha I power station. The two are linked by civil structure, controlling circuits, electrical and compressed air and water supplies, low level wastes disposal, ventilation under pressure maintenance, and, most important, the ability to transfer spent and new fuel in both directions. Because the second pool house is, by location and design, an extension of the existing installation, and since there is no design departure, regarding storage and transfer of fuel from that of the original installation, the rules and regulations applied for its construction were the same as those valid for the Atucha I construction. The requirement not to exceed a four-year period for construction and commissioning was determined by the need to have storage room for the Atucha I fuel. Argentina will meet the 1982 target by having the installation available during the second half of 1981. The second pool house is a wet storage location with a capacity of 1000 tons metallic uranium. It was designed by the Kraftwerk Union of West Germany along the same lines as the 440-ton storage location originally built with the station. The Atomic Energy Commission of Argentina has managed the construction and participated in project and design stages. As in the original pool, the 6 m long assemblies are stacked in double tiers. The cost figures which are mentioned differ from previously released figures and are not the final ones. With civil construction almost finished and mechanical erection started, the present estimates should not differ by more than 10% from the final figures. The installation has an investment cost of 61 million dollars, (1980), and, depending on the amortization time span considered, a total yearly cost per kg of capacity of metallic uranium, ranging between 5.5 and 9.3 dollars per kg

  10. Progress on Chinese evaluated nuclear parameter library (CENPL) (II)

    International Nuclear Information System (INIS)

    Su Zhongdi; Ge Zhigang; Zhou Chunmei

    1993-01-01

    CENPL collected, evaluated and compiled nuclear basic constants and model parameters. CENPL-1 contain six sub-libraries, they are: (1) Atomic masses and characteristic constants for nuclear ground states; (2) discrete level schemes and branch ratios of γ decay; (3) level density parameters; (4) giant dipole resonance parameters for γ-ray strength function (5) fission barrier parameter; (6) optical model parameters. Their progresses are introduced

  11. Reassessment and suspension of the nuclear power plant design requirement of the constraint of collective dose per unit of practice. (Requirement 6 (b), Standard AR 3.1.2)

    International Nuclear Information System (INIS)

    Amado, Valeria A.; Canoba, Analia C.; Curti, Adriana R.; Biaggio, Alfredo L.

    2009-01-01

    By the middle of 2005, the Nuclear Regulatory Authority (ARN) decided to re-assess the basis of a design requirement applicable to the limitation of nuclear power reactor radioactive discharges. Such requirement, aimed at restricting the discharge of globally dispersed long-lived radionuclides, was in force in Argentina since 1979 and was expressed as a limitation of the collective dose commitment per unit of electrical energy generated. The practical result of such regulatory action was the need to retain C-14 in the Atucha II power reactor under construction as well as in future heavy water reactors to be built in the country, and, later on, to manage it as to assure its isolation from the biosphere during an appropriate period of time. For the above-mentioned reassessment, an ad hoc task group was created and an internal report was presented to the Board of Directors by the middle of 2007. Because of such report the ARN decided to suspend the application of the requirement (i.e. it is not more mandatory, even for Atucha II). The present work presents the main aspects of that report. In particular, it explains the basis of the design requirement and the most important assumptions that triggered it. The differences between the assumptions made at that time and the reality of nuclear power generation at the beginning of the 21st Century, as well as their implications in relation to the requirement are described, including the Suess effect and its impact in the total dose due to C-14. Finally, after explaining in detail the facts that made no longer reasonable to keep in force the above mentioned requirement, the work presents the conclusions that lead the ARN to the suspension of this requirement. (author) [es

  12. Nuclear-Recoil Energy Scale in CDMS II Silicon Dark-Matter Detectors

    Energy Technology Data Exchange (ETDEWEB)

    Agnese, R.; et al.

    2018-03-07

    The Cryogenic Dark Matter Search (CDMS II) experiment aims to detect dark matter particles that elastically scatter from nuclei in semiconductor detectors. The resulting nuclear-recoil energy depositions are detected by ionization and phonon sensors. Neutrons produce a similar spectrum of low-energy nuclear recoils in such detectors, while most other backgrounds produce electron recoils. The absolute energy scale for nuclear recoils is necessary to interpret results correctly. The energy scale can be determined in CDMS II silicon detectors using neutrons incident from a broad-spectrum $^{252}$Cf source, taking advantage of a prominent resonance in the neutron elastic scattering cross section of silicon at a recoil (neutron) energy near 20 (182) keV. Results indicate that the phonon collection efficiency for nuclear recoils is $4.8^{+0.7}_{-0.9}$% lower than for electron recoils of the same energy. Comparisons of the ionization signals for nuclear recoils to those measured previously by other groups at higher electric fields indicate that the ionization collection efficiency for CDMS II silicon detectors operated at $\\sim$4 V/cm is consistent with 100% for nuclear recoils below 20 keV and gradually decreases for larger energies to $\\sim$75% at 100 keV. The impact of these measurements on previously published CDMS II silicon results is small.

  13. Development of methods for measuring materials nuclear characteristics, Phases, I, II, II and IV

    International Nuclear Information System (INIS)

    Maglic, R.

    1963-04-01

    This report contains the following phases of the project 'measurement of nuclear characteristics of reactor materials': nuclear performances of the neutron chopper; method for measuring total effective cross sections by transmission method on the chopper; review of methods for measuring activation cross sections; measurement of neutron spectra of the RA reactor and measurement of total effective cross section of gold by using the chopper

  14. Method for the Calculation of DPA in the Reactor Pressure Vessel of Atucha II

    Directory of Open Access Journals (Sweden)

    J. A. Mascitti

    2011-01-01

    It was determined that the maximum DPA rate in the RPV wall with fresh fuel element (FE is 3.76(3 × 10-12 s-1, it takes place in front of FEs BA42 and BL43, and it is symmetrical about the central channel, LG04, and LH03.

  15. Delays in nuclear power plant construction. Volume II. Final report

    International Nuclear Information System (INIS)

    Mason, G.E.; Larew, R.E.; Borcherding, J.D.; Okes, S.R. Jr.; Rad, P.F.

    1977-01-01

    The report identifies barriers to shortening nuclear power plant construction schedules and recommends research efforts which should minimize or eliminate the identified barriers. The identified barriers include (1) Design and Construction Interfacing Problems; (2) Problems Relating to the Selection and Use of Permanent Materials and Construction Methods; (3) Construction Coordination and Communication Problems; and (4) Problems Associated with Manpower Availability and Productivity

  16. Commercial Nuclear Steam-Electric Power Plants, Part II

    Science.gov (United States)

    Shore, Ferdinand J.

    1974-01-01

    Presents the pros and cons of nuclear power systems. Includes a discussion of the institutional status of the AEC, AEC regulatory record, routine low-level radiation hazards, transport of radioactive materials, storage of wastes, and uranium resources and economics of supply. (GS)

  17. Monte Carlo simulation of nuclear energy study (II). Annual report on Nuclear Code Evaluation Committee

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-01-01

    In the report, research results discussed in 1999 fiscal year at Nuclear Code Evaluation Committee of Nuclear Code Research Committee were summarized. Present status of Monte Carlo simulation on nuclear energy study was described. Especially, besides of criticality, shielding and core analyses, present status of applications to risk and radiation damage analyses, high energy transport and nuclear theory calculations of Monte Carlo Method was described. The 18 papers are indexed individually. (J.P.N.)

  18. Performances on nuclear activation analysis by TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Capannesi, G.; Rosada, A.

    1986-01-01

    Progresses in methodological research and connected applications in the field of activation analysis are introduced. Some peculiar characteristics on the TRIGA MARK II reactor have enabled the possibility of obtaining interesting results. The particular, the rotating radiation device Lazy Susan, with a capability of 40 positionings, permits homogeneity in neutron flux and energy spectrum stability within 15%. High level of precision and accuracy are obtained in analytic. Applications of major interest have been: - reference material certification; - forensic applications; - electrolytic cell productivity evaluation. The TRIGA MARK II reactor is equipped with a thermal column throughout a D 2 O diaphragm with a thickness of 70 cm. The available neutron flux has no fast and epithermal components. Via this facility a method has been tested for the instrumental determination of Al in Si metal of solar and electronic degree. (author)

  19. Nuclear pulse. II - Ensuring delivery of the doomsday signal

    Science.gov (United States)

    Broad, W. J.

    1981-06-01

    The ability of the communications systems on which U.S. strategic forces depend to survive the electromagnetic pulse (EMP) effects of a nuclear blast in the upper atmosphere is examined. It is shown that the Bell system telephone network, Autovon, on which much military communication presently depends, is especially vulnerable to EMP; while satellite and microwave communications networks are expected to be more resistant to attack. Satellites are, though, vulnerable to killer-satellite attack. Much promise is seen in the conversion of ground communications links to fiber-optic form, which is inherently highly resistant to EMP. A nuclear bomb detonated 200 miles above Nebraska would affect communications equipment throughout the contiguous U.S. with peak fields of 500,000 volts/meter.

  20. Report of the Secretary of Defense Task Force on DoD Nuclear Weapons Management. Phase II: Review of the DoD Nuclear Mission

    National Research Council Canada - National Science Library

    Schlesinger, James R; Carns, Michael P; Crouch, II, J. D; Gansler, Jacques S; Giambastiani, Jr., Edmund P; Hamre, John J; Miller, Franklin C; Williams, Christopher A; Blackwell, Jr, James A

    2008-01-01

    ...). This report covers Phase II findings and recommendations. In Phase II, the Task Force found that the lack of interest in and attention to the nuclear mission and nuclear deterrence, as discussed in our Phase I report, go well beyond the Air...

  1. Analysis of some nuclear waste management options. Volume II. Appendices

    International Nuclear Information System (INIS)

    Berman, L.E.; Ensminger, D.A.; Giuffre, M.S.; Koplik, C.M.; Oston, S.G.; Pollak, G.D.; Ross, B.I.

    1978-01-01

    This report describes risk analyses performed on that portion of a nuclear fuel cycle which begins following solidification of high-level waste. Risks associated with handling, interim storage and transportation of the waste are assessed, as well as the long term implications of disposal in deep mined cavities. The risk is expressed in terms of expected dose to the general population and peak dose to individuals in the population. This volume consists of appendices which provide technical details of the work performed

  2. Analysis of some nuclear waste management options. Volume II. Appendices

    Energy Technology Data Exchange (ETDEWEB)

    Berman, L.E.; Ensminger, D.A.; Giuffre, M.S.; Koplik, C.M.; Oston, S.G.; Pollak, G.D.; Ross, B.I.

    1978-10-10

    This report describes risk analyses performed on that portion of a nuclear fuel cycle which begins following solidification of high-level waste. Risks associated with handling, interim storage and transportation of the waste are assessed, as well as the long term implications of disposal in deep mined cavities. The risk is expressed in terms of expected dose to the general population and peak dose to individuals in the population. This volume consists of appendices which provide technical details of the work performed.

  3. Recent applications of nuclear medicine in diagnostics: II part

    Directory of Open Access Journals (Sweden)

    Giorgio Treglia

    2013-04-01

    Full Text Available Introduction: Positron-emission tomography (PET and single photon emission computed tomography (SPECT are effective diagnostic imaging tools in several clinical settings. The aim of this article (the second of a 2-part series is to examine some of the more recent applications of nuclear medicine imaging techniques, particularly in the fields of neurology, cardiology, and infection/inflammation. Discussion: A review of the literature reveals that in the field of neurology nuclear medicine techniques are most widely used to investigate cognitive deficits and dementia (particularly those associated with Alzheimer disease, epilepsy, and movement disorders. In cardiology, SPECT and PET also play important roles in the work-up of patients with coronary artery disease, providing accurate information on the state of the myocardium (perfusion, metabolism, and innervation. White blood cell scintigraphy and FDG-PET are widely used to investigate many infectious/inflammatory processes. In each of these areas, the review discusses the use of recently developed radiopharmaceuticals, the growth of tomographic nuclear medicine techniques, and the ways in which these advances are improving molecular imaging of biologic processes at the cellular level.

  4. CAREM-25: a low-risk nuclear option

    International Nuclear Information System (INIS)

    Baron, Jorge H.; Nunez Mac Leod, J.E.; Rivera, S.S.

    2000-01-01

    The future use of nuclear energy for electricity production is assumed as a viable alternative at present, mainly taking into account the high environmental impact of the fossil fuel alternatives (greenhouse effect, acid rain). In the worldwide context, however, it is desirable that the next generation of nuclear power stations to be safer than the present ones. To demonstrate the safety level of a particular nuclear installation, the Risk Analysis (or Probabilistic Safety Assessment) is the most appropriate tool. Quantitative risk estimations can be performed with PSA. The risk can be split as the product of two factors: the first one takes into account the occurrence probability of accidental sequences that involve the release of radioactive material, and the second takes into account the magnitude and consequences of such a release. In the present work, the reduction of both factors is analyzed. The probability is reduced by the use of simpler and more reliable systems to perform the safety functions, and the consequence by the use of small power production units, provided with passive mitigation systems and long response times. The work is illustrated with a risk comparison for electricity production with CAREM-25 units, towards classic production units (Atucha II). The results are based on PSAs performed for both plants. The conclusions show an effective risk reduction (both in probability and in consequence) for the innovative CAREM-25 plant, coming to doses so low as to prevent any acute effect in the nearby population. (author)

  5. Optimization in the nuclear fuel cycle II: Surface contamination

    International Nuclear Information System (INIS)

    Pereira, W.S.; Silva, A.X.; Lopes, J.M.; Carmo, A.S.; Fernandes, T.S.; Mello, C.R.; Kelecom, A.

    2017-01-01

    Optimization is one of the bases of radioprotection and aims to move doses away from the dose limit that is the borderline of acceptable radiological risk. This work aims to use the monitoring of surface contamination as a tool of the optimization process. 53 surface contamination points were analyzed at a nuclear fuel cycle facility. Three sampling points were identified with monthly mean values of contamination higher than 1 Bq ∙ cm -2 , points 28, 42 and 47. These points were indicated for the beginning of the optimization process

  6. Cardiac nuclear medicine, part II: diagnosis of coronary artery diseas

    International Nuclear Information System (INIS)

    Polak, J.F.; Holman, B.L.

    1981-01-01

    Diagnosing coronary artery disease is difficult and requires careful consideration of the roles and limitations of the tests used. Standard ECG tests are not reliable indicators of the presence of disease in asymptomatic patients. Thallium stress testing to assess ischemia and exercise ventriculography to assess functional status of the heart are limited in sensitivity and specificity. This is the second of a three-part series on cardiac nuclear medicine. Part I (Med. Instrum., May-June, 1981) focused on the commonly used examinations in cardiac physiology and pathophysiology. Part III will focus on myocardial infarction and other cardiac diseases

  7. Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

    Energy Technology Data Exchange (ETDEWEB)

    1984-01-01

    This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces.

  8. Proceedings of the 1984 DOE nuclear reactor and facility safety conference. Volume II

    International Nuclear Information System (INIS)

    1984-01-01

    This report is a collection of papers on reactor safety. The report takes the form of proceedings from the 1984 DOE Nuclear Reactor and Facility Safety Conference, Volume II of two. These proceedings cover Safety, Accidents, Training, Task/Job Analysis, Robotics and the Engineering Aspects of Man/Safety interfaces

  9. Titanium-II: an evaluated nuclear data file

    International Nuclear Information System (INIS)

    Philis, C.; Howerton, R.; Smith, A.B.

    1977-06-01

    A comprehensive evaluated nuclear data file for elemental titanium is outlined including definition of the data base, the evaluation procedures and judgments, and the final evaluated results. The file describes all significant neutron-induced reactions with elemental titanium and the associated photon-production processes to incident neutron energies of 20.0 MeV. In addition, isotopic-reaction files, consistent with the elemental file, are separately defined for those processes which are important to applied considerations of material-damage and neutron-dosimetry. The file is formulated in the ENDF format. This report formally documents the evaluation and, together with the numerical file, is submitted for consideration as a part of the ENDF/B-V evaluated file system. 20 figures, 9 tables

  10. Fission distribution measurements of Atucha's fuel pellets with solid state track detectors

    International Nuclear Information System (INIS)

    Ricabarra, M.D. Bovisio de; Waisman, Dina.

    1979-08-01

    Distribution of fissions in a UO 2 rod has been measured by means of solid state detectors. Mica muscovite and Makrofol-N detectors were used in the experiment. The merits of mica muscovite relative to the Makrofol-N for the detection of fission fragments have been verified. However both fission track detectors closely agree (0,5%) in the final fission distribution of the UO 2 rod. Sensitivity of the detectors shows to be linear in the range between 50.000and 360.000 fission tracks per square centimeter. Due to the high spatial resolution this method is better than any other technique. Determination were made in UO 2 pellets similar to the fuel element of the Atucha reactor. The average fission rate in the rod has been measured within 0,8% error, and provides an accurate determination for the distribution of fissions in the rod wich is needed for the determination of energy liberated per fission in the natural uranium rod.(author) [es

  11. The nuclear data collecting system designed with ARM and μC/OS-II

    International Nuclear Information System (INIS)

    Wang Chunsheng; Ma Yingjie; Han Feng

    2008-01-01

    Introduce a kind of nuclear data collecting system regarding ARM-μC/OS-II as the platform, gathering the GPS receiver in it. It was detailed to expatiated the key techniques of the multi-channel pulse amplitude analyzer, the interface design of LPC2148, a controller in ARM, and how to expand the RTOS and design applications on μC/OS-II. This system can communicate with the GPS-OEM module by the UART interface, collecting the GPS information synchronously as well as nuclear data. And then save and display them or pass them to the host computer by the USB interface. The embedded and Real-Time system, μC/OS-II build up the real-time and stability of the system and advance the integration. (authors)

  12. Nuclear reactions with radioactive and stable beams (Part II)

    International Nuclear Information System (INIS)

    Aguilera R, E.F.; Martinez Q, E.; Gomez C, A.; Lizcano, D.

    2005-12-01

    At the present time there is a great interest at world level in experiments, with accelerated nuclei of short half life. The dispersion, fusion, transfer and break processes in the interaction of weakly light projectiles bounded with targets of Z great its have been object of intense recent investigation, at world level. Our group, in collaboration with the University of Notre Dame, it has measured and analyzed these processes for weakly bound systems as: 6 He + 209 Bi, 8 Li + 208 Pb, 10 Be + 208 Pb. On the other hand a research line that has wakened up great interest, it is that of studies of resonant reactions using the Inverse Kinematics technique with thick targets. The use of this technique allows to measure an entire excitation function with a single bombardment. Our group has carried out, in the ININ, preliminary bombardments for the system 12 C + 4 He. This allowed to establish the feasibility of implementing this technique in our Laboratory. The application of this and other techniques to different systems like 18 O + 4 He, 12 C + 12 C, 12 C + 16 O, 16 O + 16 O, it opens the possibility to measure the fusion of these systems at very low energy and to deepen in the knowledge of the nuclear structure and the nuclear astrophysics. In this technical report, the activities carried out by our group during the second stage of this project, considered for 2005 are described. Also in that year, our group carries out a research stay in the University of Notre Dame, during this stay, the angular distribution of the projectiles of 8 B dispersed in an enriched target of 58 Ni was measured. The same as in the previous experiments, in this occasion it was also possible to measure those angular distributions of the projectiles of 7 Be and 6 Li dispersed in this same target. In this same one our stay group participates in other three experiments proposed by collaborators of other institutions (University of Notre Dame, University of Sao Paulo), where the products of the

  13. Multiparticle Production in Particle and Nuclear Collisions. II

    Science.gov (United States)

    Kanki, T.; Kinoshita, K.; Sumiyoshi, H.; Takagi, F.

    The dominant phenomenon in high-energy particle and nuclear collisions is multiple production of hadrons. This had attracted may physicists in 1950's, the period of the first remarkable development of particle physics. Multiparticle production was already observed in cosmic-ray experiments and expected to be explained as a natural consequence of the strong Yukawa interaction. Statistical and hydrodynamical models were then proposed by Fermi, Landau and others. These theories are still surviving even today as a prototype of modern ``fire-ball'' models. After twenty years, a golden age came in this field of physics. It was closely related to the rapid development of accelerator facilities, especially, the invention of colliding-beam machines which yield high enough center-of-mass energies for studying reactions with high multiplicity. Abundant data on final states of multiparticle production have been accumulated mainly by measuring inclusive cross sections and multiplicity distributions. In super high-energy bar{p}p collisions at CERN S pmacr pS Collider, we confirmed the increasing total cross section and found violations of many scaling laws which seemed to be valid at lower energies. This suggests a fundamental complexity of the multiparticle phenomena and offers new materials for further development of theoretical investigations. In the same period, studies of constituent (quark-gluon) structure of hadrons had also been develped. Nowadays, pysicists believe that the quantum chromodynamics (QCD) is the fundamental law of the hadronic world. Multiparticle dynamics should also be described by QCD. We have known that the hard-jet phenomena are well explained by the perturbative QCD. On the other hand, the soft processes are considered to be non-perturbative phenomena which have not yet been solved, and related to the mechanism of the color confinement and formation of strings or color-flux tubes. Multiparticle production would offer useful information on this

  14. Seismic fragility of nuclear power plant components (Phase II)

    International Nuclear Information System (INIS)

    Bandyopadhyay, K.K.; Hofmayer, C.H.; Kassir, M.K.; Pepper, S.E.

    1990-02-01

    As part of the Component Fragility Program which was initiated in FY 1985, three additional equipment classes have been evaluated. This report contains the fragility results and discussions on these equipment classes which are switchgear, I and C panels and relays. Both low and medium voltage switchgear assemblies have been considered and a separate fragility estimate for each type is provided. Test data on cabinets from the nuclear instrumentation/neutron monitoring system, plant/process protection system, solid state protective system and engineered safeguards test system comprise the BNL data base for I and C panels (NSSS). Fragility levels have been determined for various failure modes of switchgear and I ampersand C panels, and the deterministic results are presented in terms of test response spectra. In addition, the test data have been evaluated for estimating the respective probabilistic fragility levels which are expressed in terms of a median value, an uncertainty coefficient, a randomness coefficient and an HCLPF value. Due to a wide variation of relay design and the fragility level, a generic fragility level cannot be established for relays. 7 refs., 13 figs., 12 tabs

  15. Atucha I nuclear power plant: Probabilistic safety study. Loss-of-coolant accidents

    International Nuclear Information System (INIS)

    Perez, S.S.

    1987-01-01

    The plant response to the group of events 'large coolant loss' in order to evaluate the associated risk is analyzed. The event that covers all events of similar sequence due to its evolution features, being also the most demanded, is selected as starting event. The representative event is the 'guillotine type rupture of cold primary branch'. An annual occurrence frequency of 10/year is assumed for this event. The safety systems, when the event occurs, must assure the reactor shutdown and the core cooling, creating a heat sink to remove the decay heat. The annual frequency of core meltdown due to great loss of coolant is obtained multiplying the annual frequency of the starting event by the probability of failure of involved safety systems. By means of failure trees, the following is obtained: a) probability of failure to demand of the boron injection shutdown system = 4 x 10 -2 ; b) probability of failure to demand of the high pressure safety injection = 3 x 10 -3 ; c) probability of emergency cooling system failure = 4.4 x 10 -2 . Therefore, the three possible sequences of core meltdown have the following frequencies: λ 1 = 4 x 10 -6 /year λ 2 = 3 x 10 -7 /year λ 3 = 4.4 x 10 -6 /year. (Author)

  16. Cutting and dismantling of the South West ladder of the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Anasco, Roberto

    2006-01-01

    The metallic ladder built in stainless steel was used originally to check the welding of the reactor pressure vessel. It was located between the thermal insulation and reactor pressure vessel. Because of a failure in the mechanism, which let the ladder runs around the vessel, it had to be removed. A special tool remotely operated was designed to make different cuts in the bottom of the structure in a very high radioactive location [es

  17. Modifications in secondary circuit chemistry of Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Iglesias, Alberto M.; Jimenez Rebagliatti, Raul; Gentilli, Nelida; Raffo Calderon, Maria C.; Gemini, Hector

    1999-01-01

    The CNA I secondary circuit presents, by design, some materials which are difficult to be compatible from the corrosion point of view. The presence of Cu alloys limits the use of ammonia (or products that for decomposition generates it) as pH regulator substance which would be convenient to minimize the corrosion processes. The pH limit value in agreement with the operative experience is 9.2. This value is below the one required to minimize the effects of corrosion on the carbon steel, which is present inside the secondary circuit with a considerable exposed area and under hydrodynamic and hydrothermal conditions that favor those processes. This corrosive effect diminishes below certain limits, i. d. if the pH value is increased. The realization of this study involves three stages at least: a)- Independent measurements and description of the circuit current state; b)- Laboratory experiences of the possible alternatives into replace NH3 as alkaline agent and to provide better control of the corrosion process, on Cu alloys as well as steel alloys; c)- Plant implementations of the actions that are feasible from the point of view of power station operation, in such a way that in the secondary circuit it minimizes the presence of ammonia in the vapor phase and at the same time, the possibility of increasing the pH of the liquid phase, to diminish the corrosion phenomena of carbon steel. (author)

  18. The modernization of the nuclear power plants of Asco and Vandellos II; Modernizacion de las centrales nucleares de Asco y Vandellos II

    Energy Technology Data Exchange (ETDEWEB)

    Martinez Anton, L.

    2011-07-01

    Since the beginning of their commercial operation, the nuclear power plants of Asco and Vandellos have made design modifications aimed at improving the safety, reliability and operation of the plants. From the moment the management of the three plants was brought together within the Association Nuclear Asco-Vandellos II, and within the assets management process, joint strategic modernisation and improvement plans have been developed on the basis of the status of equipment, the evaluation of their ageing, obsolesce, degradations, manufacturers recommendations and/or the application of new regulations. The article lists the mos important actions already carried out or in the project phase for the 3 plants in the primary and secondary system, electrical and instrumentation systems and auxiliary systems, highlighting the problems and the solutions adopted in the most relevant modifications. (Author)

  19. Nuclear knowledge management at the IAEA

    International Nuclear Information System (INIS)

    Yanev, Y.

    2004-01-01

    Nuclear Knowledge Management as a part of the IAEA mission and its aim to help organizations to achieve competitive advantage; costs reduction; accelerated time to market in companies and large private sector organisations; innovation, supports error free decision making are discussed. The most important outputs such as nuclear knowledge management methodology; identifying endangered areas of nuclear science and technology; developing knowledge repositories; knowledge preservation technology; dedicated projects with Member States, (Atucha, Angra, KNK2, ) are presented. A brief review of the currently implemented with Agency's assistance project ANENT (Asian Network for Education in Nuclear Technology) is also given

  20. Independent Safety Assessment of the TOPAZ-II space nuclear reactor power system (Revised)

    International Nuclear Information System (INIS)

    1993-09-01

    The Independent Safety Assessment described in this study report was performed to assess the safety of the design and launch plans anticipated by the U.S. Department of Defense (DOD) in 1993 for a Russian-built, U.S.-modified, TOPAZ-II space nuclear reactor power system. Its conclusions, and the bases for them, were intended to provide guidance for the U.S. Department of Energy (DOE) management in the event that the DOD requested authorization under section 91b. of the Atomic Energy Act of 1954, as amended, for possession and use (including ground testing and launch) of a nuclear-fueled, modified TOPAZ-II. The scientists and engineers who were engaged to perform this assessment are nationally-known nuclear safety experts in various disciplines. They met with participants in the TOPAZ-II program during the spring and summer of 1993 and produced a report based on their analysis of the proposed TOPAZ-II mission. Their conclusions were confined to the potential impact on public safety and did not include budgetary, reliability, or risk-benefit analyses

  1. WSPEEDI-II system user's manual for a nuclear or radiological emergency

    International Nuclear Information System (INIS)

    Nakanishi, Chika; Sato, Sohei; Muto, Shigeo; Furuno, Akiko; Terada, Hiroaki; Nagai, Haruyasu

    2011-03-01

    Nuclear Emergency Assistance and Training Center (NEAT) has developed the response system to evaluate the radiological consequences of an accident on a nuclear power plant or nuclear weapons testing around Japan and to support prediction of radioactive material distributions by using an atmospheric dispersion model on the framework of the Response Assistance Network (RANET) which is established by the International Atomic Energy Agency (IAEA). For the enhancement of assistance capability to external organizations at a nuclear or radiological emergency, NEAT will introduce a computer-based emergency response system, 'Worldwide version of System for Prediction of Environmental Emergency Dose Information: WSPEEDI 2nd version (WSPEEDI-II)' developed by Division of Environmental and Radiation Sciences. This manual covers the overview of the system and configuration parameters as the basic knowledge needed for operating the systems. (author)

  2. Nuclear reactions with radioactive and stable beams (Part II); Reacciones nucleares con haces radiactivos y estables (Parte II)

    Energy Technology Data Exchange (ETDEWEB)

    Aguilera R, E.F.; Martinez Q, E.; Gomez C, A.; Lizcano, D. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2005-12-15

    At the present time there is a great interest at world level in experiments, with accelerated nuclei of short half life. The dispersion, fusion, transfer and break processes in the interaction of weakly light projectiles bounded with targets of Z great its have been object of intense recent investigation, at world level. Our group, in collaboration with the University of Notre Dame, it has measured and analyzed these processes for weakly bound systems as: {sup 6}He + {sup 209}Bi, {sup 8}Li + {sup 208}Pb, {sup 10}Be + {sup 208}Pb. On the other hand a research line that has wakened up great interest, it is that of studies of resonant reactions using the Inverse Kinematics technique with thick targets. The use of this technique allows to measure an entire excitation function with a single bombardment. Our group has carried out, in the ININ, preliminary bombardments for the system {sup 12}C + {sup 4}He. This allowed to establish the feasibility of implementing this technique in our Laboratory. The application of this and other techniques to different systems like {sup 18}O + {sup 4}He, {sup 12}C + {sup 12}C, {sup 12}C + {sup 16}O, {sup 16}O + {sup 16}O, it opens the possibility to measure the fusion of these systems at very low energy and to deepen in the knowledge of the nuclear structure and the nuclear astrophysics. In this technical report, the activities carried out by our group during the second stage of this project, considered for 2005 are described. Also in that year, our group carries out a research stay in the University of Notre Dame, during this stay, the angular distribution of the projectiles of {sup 8}B dispersed in an enriched target of {sup 58}Ni was measured. The same as in the previous experiments, in this occasion it was also possible to measure those angular distributions of the projectiles of {sup 7}Be and {sup 6}Li dispersed in this same target. In this same one our stay group participates in other three experiments proposed by collaborators of

  3. Nuclear desalination in the Arab world - Part II: Advanced inherent and passive safe nuclear reactors

    International Nuclear Information System (INIS)

    Karameldin, A.; Samer S. Mekhemar

    2004-01-01

    Rapid increases in population levels have led to greater demands for fresh water and electricity in the Arab World. Different types of energies are needed to contribute to bridging the gap between increased demand and production. Increased levels of safeguards in nuclear power plants have became reliable due to their large operational experience, which now exceeds 11,000 years of operation. Thus, the nuclear power industry should be attracting greater attention. World electricity production from nuclear power has risen from 1.7% in 1970 to 17%-20% today. This ratio had increased in June 2002 to reach more than 30%, 33% and 42% in Europe, Japan, and South Korea respectively. In the Arab World, both the public acceptance and economic viability of nuclear power as a major source of energy are greatly dependent on the achievement of a high level of safety and environmental protection. An assessment of the recent generation of advanced reactor safety criteria requirements has been carried out. The promising reactor designs adapted for the Arab world and other similar developing countries are those that profit from the enhanced and passive safety features of the new generation of reactors, with a stronger focus on the effective use of intrinsic characteristics, simplified plant design, and easy construction, operation and maintenance. In addition, selected advanced reactors with a full spectrum from small to large capacities, and from evolutionary to radical types, which have inherent and passive safety features, are discussed. The relevant economic assessment of these reactors adapted for water/electricity cogeneration have been carried out and compared with non-nuclear desalination methods. This assessment indicates that, water/electricity cogeneration by the nuclear method with advanced inherent and passive safe nuclear power plants, is viable and competitive. (author)

  4. ENC 94 International Nuclear Congress - Atoms for Energy. Transactions Vol.II: Poster Papers

    International Nuclear Information System (INIS)

    1995-01-01

    The transactions have been published in 2 volumes. Volume II contains the papers, which were orally presented in 4 sessions. In Session 1 'The need for nuclear energy in different parts of the world' was discussed in 17 contributions from the US, Korea, Turkey, Yugoslavia, Finland, the Netherlands, Poland, Russia, Croatia, Belgium and Germany. The other 3 sessions covered: Safety of operating nuclear plants (54 posters); Back-end of the fuel cycle (35 posters); Do we need new reactors to improve safety and economics ? (32 Posters)

  5. Human Reliability analysis for digitized nuclear power plants: Case study on the LingAo II nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Zou, Yan Hua; Zhang, Li [Institute of Human Factors Engineering AND Safety Management, Hunan Institute of Technology, Hengyang (China); Dai, Cao; Li, Peng Cheng; Qing, Tao [Human Factors Institute, University of South China, Hengyang (China)

    2017-03-15

    The main control room (MCR) in advanced nuclear power plants (NPPs) has changed from analog to digital control system (DCS). Operation and control have become more automated, centralized, and accurate due to the digitalization of NPPs, which has improved the efficiency and security of the system. New issues associated with human reliability inevitably arise due to the adoption of new accident procedures and digitalization of main control rooms in NPPs. The LingAo II NPP is the first digital NPP in China to apply the state-oriented procedure. In order to address issues related to human reliability analysis for DCS and DCS + state-oriented procedure, the Hunan Institute of Technology conducted a research project based on a cooperative agreement with the LingDong Nuclear Power Co. Ltd. This paper is a brief introduction to the project.

  6. Human Reliability Analysis for Digitized Nuclear Power Plants: Case Study on the LingAo II Nuclear Power Plant

    Directory of Open Access Journals (Sweden)

    Yanhua Zou

    2017-03-01

    Full Text Available The main control room (MCR in advanced nuclear power plants (NPPs has changed from analog to digital control system (DCS. Operation and control have become more automated, centralized, and accurate due to the digitalization of NPPs, which has improved the efficiency and security of the system. New issues associated with human reliability inevitably arise due to the adoption of new accident procedures and digitalization of main control rooms in NPPs. The LingAo II NPP is the first digital NPP in China to apply the state-oriented procedure. In order to address issues related to human reliability analysis for DCS and DCS + state-oriented procedure, the Hunan Institute of Technology conducted a research project based on a cooperative agreement with the LingDong Nuclear Power Co. Ltd. This paper is a brief introduction to the project.

  7. Lists I and II, nuclear medical diagnostics. As of January 18, 1983

    International Nuclear Information System (INIS)

    1983-01-01

    The information booklet presents the guidelines of the Federal Association of Panel Doctors, concerning the minimum equipment required for nuclear medical diagnostics practices (nuclear medical equipment guidelines), in the amended version of May 18, 1981; it also contains the list I (modern commercially available equipment) and the list II (older types of equipment). The devices specified in these lists are products of firms that are members of the ZVEI, and are in compliance with the guidelines of the Panel Doctors' Association. Combinations of older computer equipment/cameras with up-to-date equipment, also come up to the standards given in the guidelines if specifically mentioned therein. The list of manufacturers gives addresses of the manufacturers of the equipment stated in list I and II. An appendix up-dates the information to the date of October 1, 1986. (orig./HP) [de

  8. Utilization of noise analysis technique for mechanical vibrations estimation in the ATUCHA1 and Embalse Argentine NPP

    International Nuclear Information System (INIS)

    Lescano, V.H.; Wentzeis, L.M.; Guevara, M.; Moreno, C.; Pineyro, J.

    1996-01-01

    In Argentine, comprehensive noise measurements have been performed with the reactor instrumentation of the PHWR power plant Atucha I and Embalse. The Embalse reactor is a CANDU-600 (600 Mwe) type pressurized heavy water reactor. It's a heavy water moderator and heavy water cooled natural uranium fueled pressure tube system. Signal of vanadium and platinum type in core-self power neutron detectors of ex-core ion chambers and of a moderator pressure sensor have been recorded and analysed. The vibration of reactor internals as vertical and horizontal in-core neutron flux detectors units and the coolant channels systems, consisting of calandria and pressure tubes with fuel bundles, have been identified and monitored during normal reactor operation. Atucha I, is a PHWR reactor natural uranium fueled, and heavy water moderated and cooled. Neutron noise techniques using of ex-core ionization chambers and in-core Vanadium SPND's were implemented, among others, in order to produce early detection of anomalous vibrations in the reactor internals. Noise analysis was successfully performed to identify normal and peculiar vibrations in particular reactor internals. (author)

  9. Temperature variation on the Mediterranean Sea by the exploitation of the Vandellos II Nuclear Power Plant

    International Nuclear Information System (INIS)

    Villarreal Romero, M.; Ribes Hernandez, G.; Esparza Martin, J. L.

    2010-01-01

    The aim of this study is to verify the compliance with the Resolution of 7th February MAH/285/2007 Departament de Medi Ambient I Habitatge de la Generalitat de Catalunya establishing discharges limits to the Mediterranean Sea and, in particular, the section that references the thermal rise. The study area include about 1.5 km coastline, which is located in the vicinity of the Vandellos II Nuclear Power Plant.

  10. Development of methods for measuring materials nuclear characteristics, Phases, I, II, II and IV; Razvijanje metoda merenja nuklearnih karakteristika materijala, I, II, II i VI faza

    Energy Technology Data Exchange (ETDEWEB)

    Maglic, R [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

    1963-04-15

    This report contains the following phases of the project 'measurement of nuclear characteristics of reactor materials': nuclear performances of the neutron chopper; method for measuring total effective cross sections by transmission method on the chopper; review of methods for measuring activation cross sections; measurement of neutron spectra of the RA reactor and measurement of total effective cross section of gold by using the chopper.

  11. 14N nuclear quadrupole interaction in Cu(II) doped L-alanine

    International Nuclear Information System (INIS)

    Murgich, J.; Calvo, R.; Oseroff, S.B.; Instituto Venezolano de Investigaciones Cientificas, Caracas. Dept. de Quimica)

    1980-01-01

    The 14 N nuclear quadrupole interaction tensor Psub(N) measured by ENDOR in Cu(II) doped L-alanine is analyzed in terms of the Townes and Daily theory assuming a tetra-hedrally bonded N atom. The results of this analysis are compared with those for the 14 N in pure L-alanine and it is found that the principal directions of the Psub(N) tensor are drastically changed upon metal complexation as a consequence of the higher electron affinity of Cu(II) with respect to C and H. Comparison of the corresponding bond populations in pure and Cu(II) doped L-alanine indicates that the Cu draws 0.11 more electron from the N than the substituted H atom. (orig.)

  12. Nuclear power production costs

    International Nuclear Information System (INIS)

    Erramuspe, H.J.

    1988-01-01

    The economic competitiveness of nuclear power in different highly developed countries is shown, by reviewing various international studies made on the subject. Generation costs (historical values) of Atucha I and Embalse Nuclear Power Plants, which are of the type used in those countries, are also included. The results of an international study on the economic aspects of the back end of the nuclear fuel cycle are also reviewed. This study shows its relatively low incidence in the generation costs. The conclusion is that if in Argentina the same principles of economic racionality were followed, nuclear energy would be economically competitive in the future, as it is today. This is of great importance in view of its almost unavoidable character of alternative source of energy, and specially since we have to expect an important growth in the consumption of electricity, due to its low share in the total consumption of energy, and the low energy consumption per capita in Argentina. (Author) [es

  13. COMURHEX II, a 610 million euro investment to meet tomorrow's nuclear power needs

    International Nuclear Information System (INIS)

    2007-01-01

    The worldwide development of nuclear power will lead to increasing demands for uranium. To meet its customers' growing requirements, AREVA has chosen to invest in COMURHEX II to renew and modernize its industrial conversion tool. An entirely new plant is to be built on the Tricastin site to the north of the existing COMURHEX plant. The plant, together with other large-scale investment projects such as the future Georges Besse II enrichment plant - also located on the Tricastin site - and the modernization of the FBFC fuel fabrication facilities in the south of France, will enable AREVA to strengthen its position as a long-standing and fully integrated player at the Front End of the nuclear fuel cycle. These major investments confirm the group's strong commitment to the global development of nuclear power. Converting uranium ore into uranium hexafluoride (UF 6 ) is a key stage before the enrichment and fabrication of nuclear fuel. AREVA is gearing up for market changes, increasing its uranium production from 15,000 tons per year to 21,000 tons per year to match market needs. Today the conversion units of the different industrial operators are showing their age. They will need replacing in the medium term to increase production capacity and keep abreast of the economic, regulatory and environmental conditions of tomorrow's market. Its euros 610 million investment in the Narbonne and Pierrelatte sites in southern France will make AREVA the first uranium converter to overhaul its industrial tool. Thus indicating the group's intention of remaining world leader in UF 6 conversion. The COMURHEX II project will involve the modernization and upgrading of our installations on the basis of tried-and-tested processes and techniques, while incorporating technological innovations that will improve the production performance, reinforce nuclear safety in the facilities, while further reducing the environmental impact of their activities. The first industrial production on the

  14. Survey II of public and leadership attitudes toward nuclear power development in the United States

    International Nuclear Information System (INIS)

    Anon.

    1976-01-01

    In August 1975, Ebasco Services Incorporated released results of a survey conducted by Louis Harris and Associates, Inc. to determine attitudes of the American public and its leaders toward nuclear power development in the U.S. Results showed, among other things, that the public favored building nuclear power plants; that they believed we have an energy shortage that will not go away soon; that they were not willing to make environmental sacrifices; and that, while favoring nuclear power development, they also had concerns about some aspects of nuclear power. Except for the environmental group, the leadership group felt the same way the public does. A follow-up survey was made in July 1976 to measure any shifts in attitudes. Survey II showed that one of the real worries that remains with the American public is the shortage of energy; additionally, the public and the leaders are concerned about the U.S. dependence on imported oil. With exception of the environmentalists, the public and its leaders support a host of measures to build energy sources, including: solar and oil shale development; speeding up the Alaskan pipeline; speeding up off-shore drilling; and building nuclear power plants. The public continues to be unwilling to sacrifice the environment. There is less conviction on the part of the public that electric power will be in short supply over the next decade. The public believes the days of heavy dependence on oil or hydroelectric power are coming to an end. By a margin of 3 to 1, the public favors building more nuclear power plants in the U.S., but some concerns about the risks have not dissipated. Even though the public is worried about radioactivity escaping into the atmosphere, they consider nuclear power generation more safe than unsafe

  15. The research on corrosion condition and anticorrosion methods of SEP system pipelines in Qinshan Nuclear Power Plant Phase II

    International Nuclear Information System (INIS)

    Zhang Wei; Cao Feng; Wang Jianjun

    2010-01-01

    SEP system in Qinshan nuclear power plant phase II provides drinking water and firefight water for nuclear island, conventional island, inner and outer of BOP structures. Many corrosion perforations in the SEP pipeline were found during operation. This article analysis the corrosion reasons and presents some reasonable treatment and surveillance methods. (authors)

  16. Analysis of extreme hydrometric values in the nuclear power plants siting

    International Nuclear Information System (INIS)

    Gomez, H.R.; Maggio, G.E.; Tripoli, C.R.

    1983-01-01

    The atucha nuclear power plants are located on the right shore of the Parana de las Palmas river in the entire transition between a fluvial regime and other of tides. Since the disponibility of cooling water is one of the factors to take into account when choosing the nuclear power plant site, it is essential to perform a probabilistic study of extreme hydrometric values. Deterministic and historical analysis should be done to complete the studies already mentioned, in order to establish the values of probable maximum floods. From the application of these methods, it is concluded that the site of the Atucha nuclear power plants constitutes a hydrometric singularity, so that, an optimization has been obtained from that point of view. (Author) [es

  17. METHUSELAH II - A Fortran program and nuclear data library for the physics assessment of liquid-moderated reactors

    International Nuclear Information System (INIS)

    Brinkworth, M.J.; Griffiths, J.A.

    1966-03-01

    METHUSELAH II is a Fortran program with a nuclear data library, used to calculate cell reactivity and burn-up in liquid-moderated reactors. It has been developed from METHUSELAH I by revising the nuclear data library, and by introducing into the program improvements relating to nuclear data, improvements in efficiency and accuracy, and additional facilities which include a neutron balance edit, specialised outputs, fuel cycling, and fuel costing. These developments are described and information is given on the coding and usage of versions of METHUSELAH II for the IBM 7030 (STRETCH), IBM 7090, and KDF9 computers. (author)

  18. Potential safety enhancements to nuclear plant control: proof testing at EBR-II

    International Nuclear Information System (INIS)

    Lindsay, R.W.; Chisholm, G.H.

    1984-01-01

    Future changes in nuclear plant control and protective systems will reflect an evolutionary improvement through increased use of computers coupled with a better integration of man and machine. Before improvements can be accepted into the licensed commercial plant environment, significant testing must be accomplished to answer safety questions and to prove the worth of new ideas. The Experimental Breeder Reactor-II (EBR-II) is being used as a test-bed for both in-house development and testing for others in a DOE sponsored Man-Machine Integration program. The ultimate result of the development and testing would be a control system for which safety credit could be taken in the licensing process

  19. Combined Electrical, Optical and Nuclear Investigations of Impurities and Defects in II-VI Semiconductors

    CERN Multimedia

    2002-01-01

    % IS325 \\\\ \\\\ To achieve well controlled bipolar conductivity in II-VI semiconductors represents a fundamental problem in semiconductor physics. The doping problems are controversely discussed, either in terms of self compensation or of compensation and passivation by unintentionally introduced impurities. \\\\ \\\\It is the goal of our experiments at the new ISOLDE facility, to shed new light on these problems and to look for ways to circumvent it. For this aim the investigation of impurities and native defects and the interaction between each other shall be investigated. The use of radioactive ion beams opens the access to controlled site selective doping of only one sublattice via nuclear transmutation. The compensating and passivating mechanisms will be studied by combining nuclear, electrical and optical methods like Perturbed Angular Correlation~(PAC), Hall Effect~(HE), Deep Level Transient Spectroscopy~(DLTS), Photoluminescence Spectroscopy~(PL) and electron paramagnetic resonance (EPR). \\\\ \\\\We intend to ...

  20. The modernization of the nuclear power plants of Asco and Vandellos II

    International Nuclear Information System (INIS)

    Martinez Anton, L.

    2011-01-01

    Since the beginning of their commercial operation, the nuclear power plants of Asco and Vandellos have made design modifications aimed at improving the safety, reliability and operation of the plants. From the moment the management of the three plants was brought together within the Association Nuclear Asco-Vandellos II, and within the assets management process, joint strategic modernisation and improvement plans have been developed on the basis of the status of equipment, the evaluation of their ageing, obsolesce, degradations, manufacturers recommendations and/or the application of new regulations. The article lists the mos important actions already carried out or in the project phase for the 3 plants in the primary and secondary system, electrical and instrumentation systems and auxiliary systems, highlighting the problems and the solutions adopted in the most relevant modifications. (Author)

  1. Study and application of ANISN and DOT-II nuclear cores in reactor physics problems

    International Nuclear Information System (INIS)

    Dias, Artur Flavio

    1980-01-01

    To solve time-independent neutrons and/or gamma rays transport problems in nuclear reactors, two codes available at IPEN were studied and applied to solve benchmark problems. The ANISN code solves the one-dimensional Boltzmann transport equation for neutrons or gamma rays, in plane, spherical, or cylindrical geometries. The DOT-II code solves the same equation in two-dimensional space for plane, cylindrical and circular geometries. General anisotropic scattering allowed in both codes. Moreover, pointwise convergence criteria, and alternate step function difference equations are also used in order to remove the oscillating flux distributions, sometimes found in discrete ordinates solutions. Basic theories and numerical techniques used in these codes are studied and summarized. Benchmark problems have been solved using these codes. Comparisons of the results show that both codes can be used with confidence in the analysis of nuclear problems. (author)

  2. Exxon nuclear power distribution control for pressurized water reactors: Phase II

    International Nuclear Information System (INIS)

    Holm, J.S.; Burnside, R.J.

    1978-01-01

    The power distribution control procedure, denoted PDC-II, described in this report enables nuclear plants to manage core power distributions such that Technical Specification Limits on F/sub Q//sup T/ are not violated during normal operation and limits on MDNBR are not violated during steady-state, load-follow, and anticipated transients. The PDC-II data base described provides the means for predicting the maximum F/sub Q//sup T/(z) distribution anticipated during operation under the PDC-II procedure taking into account the incore measured equilibrium power distribution data for the reactor in question. A comparison of this distribution with the Technical Specification limit curve determines whether the Technical Specification limit can be protected by PDC-II procedure. If such protection can be confirmed for a given operating cycle interval, APDMS monitoring is not necessary over this interval and the excore monitored constant axial offset limits will protect the Technical Specification F/sub Q//sup T/ limits. This document describes the maximum possible variation in F/sub Q//sup T/(z) which can occur during operation when following the PDC-II procedures. This bounding variation in F/sub Q//sup T/(z) is referred to as V(z). This V(z) distribution represents the maximun variation in F/sub Q//sup T/(z) when the axial offset is maintained within the range defined in this report [+- 5% at full power condition

  3. Compact sodium cooled nuclear power plant with fast core (KNK II- Karlsruhe), Safety Report

    International Nuclear Information System (INIS)

    1977-09-01

    After the operation of the KNK plant with a thermal core (KNK I), the installation of a fast core (KNK II) had been realized. The planning of the core and the necessary reconstruction work was done by INTERATOM. Owner and customer was the Nuclear Research Center Karlsruhe (KfK), while the operating company was the Kernkraftwerk-Betriebsgesellschaft mbH (KBG) Karlsruhe. The main goals of the KNK II project and its special experimental test program were to gather experience for the construction, the licensing and operation of future larger plants, to develop and to test fuel and absorber assemblies and to further develop the sodium technology and the associated components. The present safety report consists of three parts. Part 1 contains the description of the nuclear plant. Hereby, the reactor and its components, the handling facilities, the instrumentation with the plant protection, the design of the plant including the reactor core and the nominal operation processes are described. Part 2 contains the safety related investigation and measures. This concerns the reactivity accidents, local cooling perturbations, radiological consequences with the surveillance measures and the justification of the choice of structural materials. Part three finally is the appendix with the figures, showing the different buildings, the reactor and its components, the heat transfer systems and the different auxiliary facilities [de

  4. Measurement of Nuclear Recoils in the CDMS II Dark Matter Search

    Science.gov (United States)

    Fallows, Scott M.

    The Cryogenic Dark Matter Search (CDMS) experiment is designed to directly detect elastic scatters of weakly-interacting massive dark matter particles (WIMPs), on target nuclei in semiconductor crystals composed of Si and Ge. These scatters would occur very rarely, in an overwhelming background composed primarily of electron recoils from photons and electrons, as well as a smaller but non-negligible background of WIMP-like nuclear recoils from neutrons. The CDMS~II generation of detectors simultaneously measure ionization and athermal phonon signals from each scatter, allowing discrimination against virtually all electron recoils in the detector bulk. Pulse-shape timing analysis allows discrimination against nearly all remaining electron recoils taking place near detector surfaces. Along with carefully limited neutron backgrounds, this experimental program allowed for "background-free'' operation of CDMS~II at Soudan, with less than one background event expected in each WIMP-search analysis. As a result, exclusionary upper-limits on WIMP-nucleon interaction cross section were placed over a wide range of candidate WIMP masses, ruling out large new regions of parameter space. These results, like any others, are subject to a variety of systematic effects that may alter their final interpretations. A primary focus of this dissertation will be difficulties in precisely calibrating the energy scale for nuclear recoil events like those from WIMPs. Nuclear recoils have suppressed ionization signals relative to electron recoils of the same recoil energy, so the response of the detectors is calibrated differently for each recoil type. The overall normalization and linearity of the energy scale for electron recoils in CDMS~II detectors is clearly established by peaks of known gamma energy in the ionization spectrum of calibration data from a 133Ba source. This electron-equivalent keVee) energy scale enables calibration of the total phonon signal (keVt) by enforcing unity

  5. Generation IV nuclear energy systems: road map and concepts. 2. Generation II Measurement Systems for Generation IV Nuclear Power Plants

    International Nuclear Information System (INIS)

    Miller, Don W.

    2001-01-01

    need for substantial research. As we consider I and C systems in Generation IV reactors, we have the opportunity to take a much less 'timid' design philosophy than was taken in the design of I and C systems in the ALWRs. We need to make use of advanced technology to design an I and C system for the Generation IV multi-unit plant designs currently being considered. Such a design should accomplish the following: 1. provides for multi-unit control; 2. contributes to a plant design objective of a very low core damage frequency; 3. maximizes plant thermal efficiency (>50%); 4. maximizes plant capacity factor (>90%); 5. optimizes operability; 6. maximizes maintainability; 7. provides for on-line monitoring, calibration, and diagnostics; 8. provides optimum response to disturbances; 9. provides excellent load-following capability. When we consider the current situation in operating Generation I and II nuclear power plants and even Generation III ALWR design, we conclude that Generation IV reactors should employ at least Generation II measurement systems. Let us first consider data transmission, which is a form of communication, and ask the question: Do new communication-transferring methods by electrons flow in copper wires? The obvious answer is no. Virtually all new communication systems are using some electromagnetic method, such as light, microwaves, HF or VHF radio signals, and virtually no copper wires. When we envision Generation IV nuclear power plants, we should minimize the use of copper wires for data transmission. We should transmit data primarily by fiber optics and various wireless methods, some of which can penetrate thick barriers. Now let us consider sensors. If we use light for data transmission, then we should also use optical-based sensors. We should also take advantage of microprocessors, which provide opportunities to embed 'intelligence' in the sensor that can be used to increase accuracy, stability, and tolerance to external stressors (i.e., radiation

  6. Molecular characterization of a nuclear topoisomerase II from Nicotiana tabacum that functionally complements a temperature-sensitive topoisomerase II yeast mutant.

    Science.gov (United States)

    Singh, B N; Mudgil, Yashwanti; Sopory, S K; Reddy, M K

    2003-07-01

    We have successfully expressed enzymatically active plant topoisomerase II in Escherichia coli for the first time, which has enabled its biochemical characterization. Using a PCR-based strategy, we obtained a full-length cDNA and the corresponding genomic clone of tobacco topoisomerase II. The genomic clone has 18 exons interrupted by 17 introns. Most of the 5' and 3' splice junctions follow the typical canonical consensus dinucleotide sequence GU-AG present in other plant introns. The position of introns and phasing with respect to primary amino acid sequence in tobacco TopII and Arabidopsis TopII are highly conserved, suggesting that the two genes are evolved from the common ancestral type II topoisomerase gene. The cDNA encodes a polypeptide of 1482 amino acids. The primary amino acid sequence shows a striking sequence similarity, preserving all the structural domains that are conserved among eukaryotic type II topoisomerases in an identical spatial order. We have expressed the full-length polypeptide in E. coli and purified the recombinant protein to homogeneity. The full-length polypeptide relaxed supercoiled DNA and decatenated the catenated DNA in a Mg(2+)- and ATP-dependent manner, and this activity was inhibited by 4'-(9-acridinylamino)-3'-methoxymethanesulfonanilide (m-AMSA). The immunofluorescence and confocal microscopic studies, with antibodies developed against the N-terminal region of tobacco recombinant topoisomerase II, established the nuclear localization of topoisomerase II in tobacco BY2 cells. The regulated expression of tobacco topoisomerase II gene under the GAL1 promoter functionally complemented a temperature-sensitive TopII(ts) yeast mutant.

  7. Hyperfine structure in the Gd II spectrum and the nuclear electric quadrupole moment of 157Gd

    International Nuclear Information System (INIS)

    Clieves, H.P.; Steudel, A.

    1979-01-01

    The hyperfine structure of 157 Gd was investigated in 20 Gd II lines by means of a photoelectric recording Fabry-Perot interferometer with digital data processing. The hyperfine splitting factors, A and B, were obtained by computer fits to the observed line structures. Using a multiconfigurational set of wave functions in intermediate coupling derived by Wyart, mono-electronic parameters were deduced by a parametric treatment. The nuclear electric quadrupole moment of 157 Gd was evaluated from the quadrupole interaction of the 5d electron in 4f 7 5d6s, the 5d electron in 4f 7 5d6p, and the 6p electron in 4f 7 5d6p. The three values obtained for the quadrupole moment agree very well. The final result, corrected for Sternheimer shielding, is Q( 157 Gd) = 1.34(7) x 10 -24 cm 2 . (orig.) [de

  8. KNK II, Compact Sodium-Cooled Reactor in the Nuclear Research Center Karlsruhe

    International Nuclear Information System (INIS)

    1978-01-01

    The report gives an overview of the project of the sodium-cooled fast reactor KNK II in the nuclear research center KfK in Karlsruhe. This test reactor was the preparatory stage of the prototype plant SNR 300 and had several goals: to train operating personal, to practice the licensing procedures in Germany, to get experience with the sodium technology and to serve as a test bed for fast breeder core components. The report contains contributions of KfK as the owner and project managing organization, of INTERATOM as the design and construction company and of the KBG as the plant operating organization. Experience with and results of relevant aspects of the project are tackled: project management, reactor core and component design, safety questions and licensing, plant design and test programs [de

  9. Measurement of Nuclear Recoils in the CDMS II Dark Matter Search

    Energy Technology Data Exchange (ETDEWEB)

    Fallows, Scott Mathew [Univ. of Minnesota, Minneapolis, MN (United States)

    2014-12-01

    The Cryogenic Dark Matter Search (CDMS) experiment is designed to directly detect elastic scatters of weakly-interacting massive dark matter particles (WIMPs), on target nuclei in semiconductor crystals composed of Si and Ge. These scatters would occur very rarely, in an overwhelming background composed primarily of electron recoils from photons and electrons, as well as a smaller but non-negligible background of WIMP-like nuclear recoils from neutrons. The CDMS II generation of detectors simultaneously measure ionization and athermal phonon signals from each scatter, allowing discrimination against virtually all electron recoils in the detector bulk. Pulse-shape timing analysis allows discrimination against nearly all remaining electron recoils taking place near detector surfaces. Along with carefully limited neutron backgrounds, this experimental program allowed for \\background- free" operation of CDMS II at Soudan, with less than one background event expected in each WIMP-search analysis. As a result, exclusionary upper-limits on WIMP-nucleon interaction cross section were placed over a wide range of candidate WIMP masses, ruling out large new regions of parameter space.

  10. Nuclear proliferation and civilian nuclear power. Report of the Nonproliferation Alternative Systems Assessment Program. Volume II. Proliferation resistance

    International Nuclear Information System (INIS)

    1980-06-01

    The purpose of this volume is limited to an assessment of the relative effects that particular choices of nuclear-power systems, for whatever reasons, may have on the possible spread of nuclear-weapons capabilities. This volume addresses the concern that non-nuclear-weapons states may be able to initiate efforts to acquire or to improve nuclear-weapons capabilities through civilian nuclear-power programs; it also addresses the concern that subnational groups may obtain and abuse the nuclear materials or facilities of such programs, whether in nuclear-weapons states (NWS's) or nonnuclear-weapons states (NNW's). Accordingly, this volume emphasizes one important factor in such decisions, the resistance of nuclear-power systems to the proliferation of nuclear-weapons capabilities

  11. Role of the chemistry in the occupational dose control in nuclear power plants

    International Nuclear Information System (INIS)

    Blesa, M.A.

    1988-01-01

    The safety and radioprotection problem in nuclear power plants does not only concern the plants in operation, but it also includes the design, building and decommissioning stages. The factors that determine the radiation field development and the possibility of diminishing them when they reach critical values are presented. Here are considered pressure vessel-heavy water reactors and particularly the radionucleides coming from the products of structural material corrosion. These products are removed by decontaminating compounds and primary circuit systems in general. In accordance with ALARA criterium, the factors that make the decontamination process advisable are analyzed. Firstly, the number of collective doses is discused. In case of heavily contaminated components there is another limitation to the ALARA criterion: the limits of individual dose in a fixed period of time (year, trimester, etc). Among the various decontamination processes-physical or chemical - the stages to follow just in chemical procedures are stated. As Atucha I and II Power Plants are uniques it is necessary to be ready to solve problems. The research and development programs of National Atomic Energy Commission have produced very valuable results such as the 'in situ' activation model and the HERO (high efficient electrochemical removal of oxides) decontamination procedure. (M.E.L.) [es

  12. Safety Evaluation Report related to the final design approval of the GESSAR II BWR/6 Nuclear Island Design (Docket No. 50-447). Supplement No. 3

    International Nuclear Information System (INIS)

    1985-01-01

    Supplement 3 to the Safety Evaluation Report (SER) for the application filed by General Electric Company for the final design approval for the GE BWR/6 nuclear island design has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. This report supplements the GESSAR II SER (NUREG-0979), issued in April 1983, summarizing the results of the staff's safety review of the GESSAR II BWR/6 nuclear island design. Subject to favorable resolution of the items discussed in this supplement, the staff concludes that the GESSAR II design satisfactorily addresses the severe-accident concerns described in draft NUREG-1070

  13. Safety Evaluation Report related to the final design approval of the GESSAR II BWR/6 Nuclear Island design, Docket No. 50-447

    International Nuclear Information System (INIS)

    1983-04-01

    The Safety Evaluation Report for the application filed by General Electric Company for the Final Design Approval for the General Electric Standard Safety Analysis Report (GESSAR II FSAR) has been prepared by the Office of Nuclear Reactor Regulation of the Nuclear Regulatory Commission. This report summarizes the results of the staff's safety review of the GESSAR II BWR/6 Nuclear Island Design. Subject to favorable resolution of items discussed in the Safety Evaluation Report, the staff concludes that the facilities referencing GESSAR II, subject to approval of the balance-of-plant design, can conform with the provisions of the Act and the regulations of the Nuclear Regulatory Commission

  14. Beznau II nuclear power plant: Expertise on NOK's request for the removal of the time limitation for the operation licence

    International Nuclear Information System (INIS)

    2004-03-01

    The Federal Agency for the Safety of Nuclear Installations (HSK) is the Swiss authority responsible for nuclear safety and protection against radioactivity in nuclear power plants. It has to examine the request of the North-East Swiss Power Corporation (NOK) concerning the removal of the operational time limitation for the Beznau-II reactor (KKB-II). In the present report HSK reviews the enterprise management and the safety of KKB-II on the basis of the results of the Periodic Safety Review. The Beznau nuclear power plant exhibits a very high degree of technical and organisational safety. During the past 10 years the plant has been operated in a safe manner. At the same time the plant has been improved and this guarantees that the mechanisms of ageing degradation are systematically identified and that measures can be taken that are possibly necessary. Under such conditions, the safety of KKB-II can be guarantied at all times. As a result of the management of quality, environmental and working safety conditions, the correct application and the continuous improvement of all processes important to safety are ensured. With these measures KKB has shown that safety is given priority over and against all other working goals. The examination by HSK of the Periodic Safety Review has shown that, in the past, KKB has applied modernisation measures independent of the licensing situation of the two reactor blocks. These modernisation measures largely contribute to the fact that the HSK examination did not reveal any significant safety deficiencies. Other improvement measures allow risk reduction or can bee seen as an adaptation to experience gained and to the state of the technological art. In conclusion, HSK states that no safety-relevant facts have been found which could prevent the removal of the time limitation on the operational licence for KKB-II. From the point of view of HSK, KKB-II fulfils the conditions for the safe continuation of operation

  15. A Procedure to Address the Fuel Rod Failures during LB-LOCA Transient in Atucha-2 NPP

    Directory of Open Access Journals (Sweden)

    Martina Adorni

    2011-01-01

    Full Text Available Depending on the specific event scenario and on the purpose of the analysis, the availability of calculation methods that are not implemented in the standard system thermal hydraulic codes might be required. This may imply the use of a dedicated fuel rod thermomechanical 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 thermalhydraulics, 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 are provided by core physics and three-dimensional neutron kinetic coupled thermal-hydraulic system codes calculations. The procedure is completed by the sensitivity calculations and the application of the probabilistic method, which are outside the scope of the current paper.

  16. Aging of snubbers in nuclear service: Phase I study results and Phase II plans

    International Nuclear Information System (INIS)

    Goodman, R.L.; Bush, S.H.; Page, R.E.

    1988-01-01

    Two major research areas were investigated in the Phase I snubber aging studies. The first area involved a preliminary evaluation of the effects of various aging mechanisms on snubber operation; failure modes of mechanisms were identified and their contributions to aging degradation were assessed relative to other failure modes. The second area involved estimating the efficacy of existing tests and examinations that are intended to determine the effects of aging and degradation. Available data on snubber behavior and operating experience were reviewed, using licensee event reports and other historical data for the 10-year period from 1973 through 1983. Value-impact was considered in terms of (1) exposure of workers to radioactive environments for examination/testing and (2) the cost for expansion of the snubber testing program due to failed snubbers. Results from the Phase I studies identified the need to modify or improve examination and testing procedures to enhance snubber reliability. Based on the results of the Phase I snubber studies, the seals and fluids were identified as the two principal elements affected by aging degradation in hydraulic snubbers. Phase II work, which was initiated in FY 1987, will develop cooperative activities between PNL and operating utilities through the Snubber Utility Group (SNUG), who will work to establish a strong data and experience base for both hydraulic and mechanical snubbers based on actual operating and maintenance history at nuclear power plants. Application guidelines for snubbers will be recommended based on the study results

  17. The Gpn3 Q279* cancer-associated mutant inhibits Gpn1 nuclear export and is deficient in RNA polymerase II nuclear targeting.

    Science.gov (United States)

    Barbosa-Camacho, Angel A; Méndez-Hernández, Lucía E; Lara-Chacón, Bárbara; Peña-Gómez, Sonia G; Romero, Violeta; González-González, Rogelio; Guerra-Moreno, José A; Robledo-Rivera, Angélica Y; Sánchez-Olea, Roberto; Calera, Mónica R

    2017-11-01

    Gpn3 is required for RNA polymerase II (RNAPII) nuclear targeting. Here, we investigated the effect of a cancer-associated Q279* nonsense mutation in Gpn3 cellular function. Employing RNAi, we replaced endogenous Gpn3 by wt or Q279* RNAi-resistant Gpn3R in epithelial model cells. RNAPII nuclear accumulation and transcriptional activity were markedly decreased in cells expressing only Gpn3R Q279*. Wild-type Gpn3R localized to the cytoplasm but a fraction of Gpn3R Q279* entered the cell nucleus and inhibited Gpn1-EYFP nuclear export. This property and the transcriptional deficit in Gpn3R Q279*-expressing cells required a PDZ-binding motif generated by the Q279* mutation. We conclude that an acquired PDZ-binding motif in Gpn3 Q279* caused Gpn3 nuclear entry, and inhibited Gpn1 nuclear export and Gpn3-mediated RNAPII nuclear targeting. © 2017 Federation of European Biochemical Societies.

  18. MAP kinase-signaling controls nuclear translocation of tripeptidyl-peptidase II in response to DNA damage and oxidative stress

    Energy Technology Data Exchange (ETDEWEB)

    Preta, Giulio; Klark, Rainier de; Chakraborti, Shankhamala [Center for Molecular Medicine (CMM), Department of Medicine, Karolinska Institutet, Karolinska University Hospital, 171 76 Stockholm (Sweden); Glas, Rickard, E-mail: rickard.glas@ki.se [Center for Molecular Medicine (CMM), Department of Medicine, Karolinska Institutet, Karolinska University Hospital, 171 76 Stockholm (Sweden)

    2010-08-27

    Research highlights: {yields} Nuclear translocation of TPPII occurs in response to different DNA damage inducers. {yields} Nuclear accumulation of TPPII is linked to ROS and anti-oxidant enzyme levels. {yields} MAPKs control nuclear accumulation of TPPII. {yields} Inhibited nuclear accumulation of TPPII decreases DNA damage-induced {gamma}-H2AX expression. -- Abstract: Reactive oxygen species (ROS) are a continuous hazard in eukaroytic cells by their ability to cause damage to biomolecules, in particular to DNA. Previous data indicated that the cytosolic serine peptidase tripeptidyl-peptidase II (TPPII) translocates into the nucleus of most tumor cell lines in response to {gamma}-irradiation and ROS production; an event that promoted p53 expression as well as caspase-activation. We here observed that nuclear translocation of TPPII was dependent on signaling by MAP kinases, including p38MAPK. Further, this was caused by several types of DNA-damaging drugs, a DNA cross-linker (cisplatinum), an inhibitor of topoisomerase II (etoposide), and to some extent also by nucleoside-analogues (5-fluorouracil, hydroxyurea). In the minority of tumor cell lines where TPPII was not translocated into the nucleus in response to DNA damage we observed reduced intracellular ROS levels, and the expression levels of redox defense systems were increased. Further, treatment with the ROS-inducer {gamma}-hexa-chloro-cyclohexane ({gamma}-HCH, lindane), an inhibitor of GAP junctions, restored nuclear translocation of TPPII in these cell lines upon {gamma}-irradiation. Moreover, blocking nuclear translocation of TPPII in etoposide-treated cells, by using a peptide-derived inhibitor (Z-Gly-Leu-Ala-OH), attenuated expression of {gamma}-H2AX in {gamma}-irradiated melanoma cells. Our results indicated a role for TPPII in MAPK-dependent DNA damage signaling.

  19. Validation of finite element code DELFIN by means of the zero power experiences at the nuclear power plant of Atucha I; Convalidacion del codigo DELFIN por medio de las experiencias a potencia cero de la central nuclear Atucha I

    Energy Technology Data Exchange (ETDEWEB)

    Grant, C R [Comision Nacional de Energia Atomica, San Martin (Argentina). Unidad de Actividad Reactores y Centrales Nucleares

    1997-12-31

    Code DELFIN, developed in CNEA, treats the spatial discretization using heterogeneous finite elements, allowing a correct treatment of the continuity of fluxes and currents among elements and a more realistic representation of the hexagonal lattice of the reactor. It can be used for fuel management calculation, Xenon oscillation and spatial kinetics. Using the HUEMUL code for cell calculation (which uses a generalized two dimensional collision probability theory and has the WIMS library incorporated in a data base), the zero power experiences performed in 1974 were calculated. (author). 8 refs., 9 figs., 3 tabs.

  20. The development of the nuclear physics in Latvia II. The building of the Research Nuclear Reactor IRT

    International Nuclear Information System (INIS)

    Ulmanis, U.

    2004-01-01

    Nuclear research reactor IRT of the Academy of Sciences was built near Riga in Salaspils. IRT is pool aqueous - aqueous reactor with nuclear fuel U-235 contained elements, located in the core at a depth of ∼ 7 m under distilled water. Ten horizontal and 10-15 vertical experimental channels are employed in experimental research with the use of neutron fluxes. For the research with gamma rays is constructed radiation loop facility with liquid In-Ga-SN solid solution as intensive gamma-ray sources. Main activities of IRT are to conduct research in nuclear spectroscopy, neutron activation analysis, neutron diffraction and radiation physics, chemistry and biology. (authors)

  1. Preface: II Russian-Spanish Congress on Particle and Nuclear Physics at All Scales, Astroparticle Physics and Cosmology

    OpenAIRE

    Andrianov, Alexander A.; Espriu, D. (Domènec); Andrianov, Vladimir A.; Kolevatov, S.

    2014-01-01

    This publication contains the proceedings of the II Russian-Spanish Congress on Particle and Nuclear Physics at All Scales, Astroparticle Physics and Cosmology, a collection of refereed papers presented in plenary and parallel sessions at a meeting that gathered leading Russian and Spanish Scientists in the above fields in Saint-Petersburg from October 1st through October 4th 2013 (http://hep.phys.spbu.ru/conf/esp-rus2013/).

  2. The ring-stiffened shell of the ISAR II nuclear power plant natural-draught cooling tower

    International Nuclear Information System (INIS)

    Form, J.

    1986-01-01

    The natural-draught cooling tower of the ISAR II nuclear power plant is one of the largest in the world. The bid specifications provided for an unstiffened cooling tower shell. For the execution, however, it was decided to adopt a shell with three additional stiffening rings. The present contribution deals with the static and dynamic calculations of the execution and, in particular, with the working technique employed for the construction of the rings. (author)

  3. Experimental and inspection facilities in post-irradiation of spent fuel pools for the analysis of the behaviour of nuclear fuels in power reactors

    International Nuclear Information System (INIS)

    Ruggirello, G.; Zawerucha, A.

    1992-01-01

    Since the beginning of the Atomic Nuclear Reactors (PHWR) Atucha I and Embalse in Argentine are employed different techniques for the knowing of the fuel bundles performances. It is detailed the facilities on post-irradiation examination. The techniques described are: online measurements, visual inspections, identifications of defective fuels and rods assemblies in spent fuel pools. This controls have made possible the feed-back to the manufactory process and the changes in the manufactory quality controls. (author)

  4. Dual Nuclear/Fluorescence Imaging Potantial of Zinc(II) Phthalocyanine in MIA PaCa-2 Cell Line.

    Science.gov (United States)

    Lambrecht, Fatma Yurt; Ince, Mine; Er, Ozge; Ocakoglu, Kasim; Sarı, Fatma Aslıhan; Kayabasi, Cagla; Gunduz, Cumhur

    2016-01-01

    Pancreatic cancer is very common and difficult to diagnose in early stage. Imaging systems for diagnosing cancer have many disadvantages. However, combining different imaging modalities offers synergistic advantages. Optical imaging is the most multidirectional and widely used imaging modality in both clinical practice and research. In present study, Zinc(II) phthalocyanine [Zn(II)Pc] was synthesized, labeled with iodine- 131 and in vitro study was carried out. The intracellular uptake studies of radiolabeled Zn(II)Pc were performed in WI-38 [ATCC CCL-75™, tissue: human fibroblast lung] and MIA PaCa-2 [ATCC CRL-1420™, tissue: human epithelial pancreas carcinoma] cell lines. The intracellular uptake efficiency of radiolabeled Zn(II)Pc in MIA PaCa-2 cells was determined two times higher than WI-38 cells. Also, fluorescence imaging (FI) efficiency of synthesized Zn(II)Pc was investigated in MIA PaCa-2 cells and significant uptake was observed. Zn(II)Pc might be used as a new agent for dual fluorescence/nuclear imaging for pancreatic cancer. Copyright© Bentham Science Publishers; For any queries, please email at epub@benthamscience.org.

  5. Oxidation behavior analysis of cladding during severe accidents with combined codes for Qinshan Phase II Nuclear Power Plant

    International Nuclear Information System (INIS)

    Shi, Xingwei; Cao, Xinrong; Liu, Zhengzhi

    2013-01-01

    Highlights: • A new verified oxidation model of cladding has been added in Severe Accident Program (SAP). • A coupled analysis method utilizing RELAP5 and SAP codes has been developed and applied to analyze a SA caused by LBLOCA. • Analysis of cladding oxidation under a SA for Qinshan Phase II Nuclear Power Plant (QSP-II NPP) has been performed by SAP. • Estimation of the production of hydrogen has been achieved by coupled codes. - Abstract: Core behavior at a high temperature is extremely complicated during transition from Design Basic Accident (DBA) to the severe accident (SA) in Light Water Reactors (LWRs). The progression of core damage is strongly affected by the behavior of fuel cladding (oxidation, embrittlement and burst). A Severe Accident Program (SAP) is developed to simulate the process of fuel cladding oxidation, rupture and relocation of core debris based on the oxidation models of cladding, candling of melted material and mechanical slumping of core components. Relying on the thermal–hydraulic boundary parameters calculated by RELAP5 code, analysis of a SA caused by the large break loss-of-coolant accident (LBLOCA) without mitigating measures for Qinshan Phase II Nuclear Power Plant (QSP-II NPP) was performed by SAP for finding the key sequences of accidents, estimating the amount of hydrogen generation and oxidation behavior of the cladding

  6. Assessment of RELAP5/MOD2 against a main feedwater turbopump trip transient in the Vandellos II Nuclear Power Plant

    International Nuclear Information System (INIS)

    Llopis, C.; Casals, A.; Perez, J.; Mendizabal, R.

    1993-12-01

    The Consejo de Seguridad Nuclear (CSN) and the Asociacion Nuclear Vandellos (ANV) have developed a model of Vandellos II Nuclear Power Plant. The ANV collaboration consisted in the supply of design and actual data, the cooperation in the simulation of the control systems and other model components, as well as in the results analysis. The obtained model has been assessed against the following transients occurred in plant: A trip from the 100% power level (CSN); a load rejection from 100% to 50% (CSN); a load rejection from 75% to 65% (ANV); and, a feedwater turbopump trip (ANV). This copy is a report of the feedwater turbopump trip transient simulation. This transient actually occurred in the plant on June 19, 1989

  7. Evaluation of Argentinian industrial capacity and of suppliers for nuclear installations

    International Nuclear Information System (INIS)

    Volman de Tanis, S.

    1986-04-01

    This work describes and analyses the Argentinian system of purchasing, the laws and decrees which must be observed by the State enterprises and dependent bodies such as the National Atomic Energy Commission (C.N.E.A.). Within the system there are regulations which must be observed by the State suppliers, and to which the purchasing bodies must conform. Furthermore, there is a thorough analysis of the methods implemented before beginning the construction of the third nuclear plant in Argentina, the Atucha II Nuclear Plant. For this, all the existing antecedents were taken into account, insofar as they were related to the prior projects, and an exhaustive questionnaire was elaborated, in which the examined enterprises described in detail their installed capacity, their antecedents, their human resources, etc. The data thus obtained were evaluated and the task was concluded by drawing up lists of the possible enterprises to supply the most diverse components, services or equipment required both by the electrical plant and for any other large scale work. The evaluation obtained would allow an analysis of the foreign offers concerning the entire project and of the possible participation with each bidder of Argentinian industry and engineering. A description is given of the advanced method used to assess bidders with respect to such participation, to assess the replies to the questionnaire, which were analysed in detail, and the weighting factors applied to each item. Also described are some major contracts concluded with enterprises in the country. These contracts relate to the external acquisition of technology for nuclear plants and the essential points are outlined. Also analysed are the results obtained during the execution of the programmes of technology transfer and training, both for parts of different enterprises selected to produce the components, and for the State, through C.N.E.A. and ENACE S.A. 6 refs, 5 tabs

  8. CINCH-II project. Next step in the coordination of education in nuclear- and radiochemistry in Europe

    International Nuclear Information System (INIS)

    John, Jan; Cuba, Vaclav; Nemec, Mojmir

    2013-01-01

    Any of the potential options for the nuclear power – both the renaissance, if any, or the phase out – will require significant numbers of the respective specialists, amongst others the nuclear and/or radiochemists. In parallel, a significant demand exists for these specialists in non-energy fields, such as environmental protection, radiopharmacy, nuclear medicine, biology, authorities, etc. Since the numbers of staff in teaching and the number of univerzities with facilities licensed for the work with open sources of ionizing radiation has decreased on or sometimes even below the critical level, coordination and collaboration are required to maintain the necessary teaching and training capabilities. The CINCH-II project, aiming at the Coordination of education and training In Nuclear CHemistry in Europe, will be a direct continuation of the CINCH-I project which, among others, identified the EuroMaster in Nuclear Chemistry quality label recognized and guaranteed by the European Chemistry Thematic Network Association as an optimum common mutual recognition system in the field of education in Nuclear Chemistry in Europe, surveyed the status of Nuclear Chemistry in industry / the needs of the end-users, developed an efficient system of education/training compact modular courses, or developed and tested two electronic tools as a basis of a future efficient distance learning system. In the first part of this paper, the achievements of the CINCH-I project will be described. This description will cover both the status review and the development activities of this Collaboration. In the status review field, the results of a detailed survey of the universities and curricula in nuclear- and radiochemistry in Europe and Russia will be presented. Another survey mapped the nuclear- and radiochemistry in industry – specifically the training and education needs of the end users. In the development activities field, the main achievements of the CINCH-project will be presented

  9. Safety analysis of Atucha 1 reactor pressure vessel for a typical transient

    International Nuclear Information System (INIS)

    Chomik, E.; Jinchuk, D.

    1994-01-01

    As a consequence of disturbances on the CNA I external electric grid some incidents were produced in a 6 minutes lapse, causing a sudden cooling of the primary system, while pressure was maintained nearly constant. On the basis of this event, a safety analysis based on the LInear Elastic Fracture Mechanics was carried out. This paper presents an alternative method for the calculation of transients; the Finite Element Method, particularly, the OCA-II FEM code. By using this method it was possible to demonstrate, for this event, a safe operating condition for the end of life of the RPV, with regard to brittle fracture risk. 6 refs, 11 figs, 1 tab

  10. EMPIRE-II 2.18, Comprehensive Nuclear Model Code, Nucleons, Ions Induced Cross-Sections

    International Nuclear Information System (INIS)

    Herman, Michal Wladyslaw; Panini, Gian Carlo

    2003-01-01

    1 - Description of program or function: EMPIRE-II is a flexible code for calculation of nuclear reactions in the frame of combined optical, Multi-step Direct (TUL), Multi-step Compound (NVWY) and statistical (Hauser-Feshbach) models. Incident particle can be a nucleon or any nucleus(Heavy Ion). Isomer ratios, residue production cross sections and emission spectra for neutrons, protons, alpha-particles, gamma-rays, and one type of Light Ion can be calculated. The energy range starts just above the resonance region for neutron induced reactions and extends up to several hundreds of MeV for the Heavy Ion induced reactions. IAEA1169/06: This version corrects an error in the Absoft compile procedure. 2 - Method of solution: For projectiles with A<5 EMPIRE calculates fusion cross section using spherical optical model transmission coefficients. In the case of Heavy Ion induced reactions the fusion cross section can be determined using various approaches including simplified coupled channels method (code CCFUS). Pre-equilibrium emission is treated in terms of quantum-mechanical theories (TUL-MSD and NVWY-MSC). MSC contribution to the gamma emission is taken into account. These calculations are followed by statistical decay with arbitrary number of subsequent particle emissions. Gamma-ray competition is considered in detail for every decaying compound nucleus. Different options for level densities are available including dynamical approach with collective effects taken into account. EMPIRE contains following third party codes converted into subroutines: - SCAT2 by O. Bersillon, - ORION and TRISTAN by H. Lenske and H. Wolter, - CCFUS by C.H. Dasso and S. Landowne, - BARMOM by A. Sierk. 3 - Restrictions on the complexity of the problem: The code can be easily adjusted to the problem by changing dimensions in the dimensions.h file. The actual limits are set by the available memory. In the current formulation up to 4 ejectiles plus gamma are allowed. This limit can be relaxed

  11. Studies on the radioactive contamination due to nuclear detonations II. Preliminary findings on the radioactive fallout due to nuclear detonations

    Energy Technology Data Exchange (ETDEWEB)

    Nishiwaki, Yasushi [Nuclear Reactor Laboratory, Tokyo Institute of Technology, Tokyo (Japan); Nuclear Reactor Laboratoroy, Kinki University, Fuse City, Osaka Precture (Japan)

    1961-11-25

    Since we have detected a considerable amount of artificial radioactivity in the rain in spring 1954, it has become one of the most important items, from the health physics point of view, to continue measurements of radioactivity in the rain and in the atmosphere. To watch out the radioactive contamination of our environment due to repeated nuclear weapons testings in other countries was also considered to be important from the nuclear engineering point of view, in the sense that the permissible allowances of the radioactivity for the peaceful uses of atomic energy might be lowered if the degree of radioactive contamination due to nuclear testings should continue to increase gradually and indefinitely. If the permissible level were lowered, the cost for radiation protection may be expected to increase at the peaceful uses of atomic energy and should the radioactive contamination increase seriously in the future, it was anticipated that we may have to face a very difficult situation in designing the atomic energy facilities for peaceful purposes in our country. From these points of views, we have been continuing measurements of the radioactivity in the rain in Osaka, Japan since the spring of 1954. Some of the preliminary findings are introduced in this paper.

  12. Studies on the radioactive contamination due to nuclear detonations II. Preliminary findings on the radioactive fallout due to nuclear detonations

    International Nuclear Information System (INIS)

    Nishiwaki, Yasushi

    1961-01-01

    Since we have detected a considerable amount of artificial radioactivity in the rain in spring 1954, it has become one of the most important items, from the health physics point of view, to continue measurements of radioactivity in the rain and in the atmosphere. To watch out the radioactive contamination of our environment due to repeated nuclear weapons testings in other countries was also considered to be important from the nuclear engineering point of view, in the sense that the permissible allowances of the radioactivity for the peaceful uses of atomic energy might be lowered if the degree of radioactive contamination due to nuclear testings should continue to increase gradually and indefinitely. If the permissible level were lowered, the cost for radiation protection may be expected to increase at the peaceful uses of atomic energy and should the radioactive contamination increase seriously in the future, it was anticipated that we may have to face a very difficult situation in designing the atomic energy facilities for peaceful purposes in our country. From these points of views, we have been continuing measurements of the radioactivity in the rain in Osaka, Japan since the spring of 1954. Some of the preliminary findings are introduced in this paper

  13. In service inspection of the reactor pressure vessel coolant and moderator nozzles at Atucha 1. 1998/1999 outages

    International Nuclear Information System (INIS)

    Antonaccio, Carlos; Conde, Alberto; Fittipaldi, Andres H.; Maniotti, Jorge; Moliterno, Gabriel E.

    2000-01-01

    During the August 1998 and the August 1999 Atucha 1 outages, two areas were inspected on the Reactor Pressure Vessel: the nozzle inner radii and the nozzle shell welds on all 3 moderator nozzles and all 4 main coolant nozzles. The inspections themselves were carried out by Mitsui Babcock Energy Limited from Scotland. The coordination, maintenance assistant and mounting of the manipulator devices over the nozzles were carried out by NASA personnel. Although it was not the first time the nozzle shell welds were inspected, due to the technologies advances in the ultrasonic field and in the inspection manipulators (magnetic ones), it was possible to inspect more volume than in previous inspections. In the other hand, it was the first time NASA was able to inspect the inner radii. In this last case the mayor problems to inspect them were the nozzles geometry and the small space available to install manipulators. The result of the inspections were: 1) There were no reportable indications at any of the inner radii inspected; 2) The inspection of nozzle to shell welds in main-coolant nozzles R3 and R4 detected flaws (one in each nozzle) which were reported as exceeding the dimensions specified as the acceptance level under Table IWB 3512-1, Section XI of the ASME code. Subsequent analysis requested by NASA and performed by Mitsui Babcock, demonstrated that the flaws were over dimensioned and could be explained as due to 'point' flaws. The analysis was based on theoretical mathematic model and experimental trials. Therefore their dimension were under the acceptance level of the ASME XI code. Although the Mitsui Babcock analysis, and at the same time it was in progress, it was assumed that the flaws were as they were originally presented (exceeding the acceptance level). NASA asked SIEMENS/KWU, the designer of the plant, to perform the fracture assessment according to ASME XI App. A. The assessment shows that the expected crack growth is negligibly small and the safety

  14. Nuclear dawn F. E. Simon and the race for atomic weapons in World War II

    CERN Document Server

    McRae, Kenneth D

    2014-01-01

    This book provides a rounded biography of Franz (later Sir Francis) Simon, his early life in Germany, his move to Oxford in 1933, and his experimental contributions to low temperature physics approximating absolute zero. After 1939 he switched his research to nuclear physics, and is credited with solving the problem of uranium isotope separation by gaseous diffusion for the British nuclear programme Tube Alloys. The volume is distinctive for its inclusion of source materials not available to previous researchers, such as Simon's diary and his correspondence with his wife, and for a fresh, well-informed insider voice on the five-power nuclear rivalry of the war years. The work also draws on a relatively mature nuclear literature to attempt a comparison and evaluation of the five nuclear rivals in wider political and military context, and to identify the factors, or groups of factors, that can explain the results.

  15. Experience in the chemistry field from the operating cycle of Grohnde and Philippsburg II nuclear power stations

    International Nuclear Information System (INIS)

    Jacobi, G.; Ruehle, W.

    1987-01-01

    Experience from the primary section of the plants in relation to the activity pattern of corrosion products, indicates primarily that cobalt-free materials have been used throughout in Philippsburg II nuclear power station, which was no longer economically possible at Grohnde because of the advanced stages of manufacture and installation. Consequently, the activity concentration for Co-60 in Philippsburg was lower from the outset than at a comparable time at Grohnde. The second part of the paper discusses experience from the secondary section of the plants, based on the AVT (all volatile treatment) method of operation and its effect on the deposits in the steam generators. The chemical control is described and a comparison is made between the sampling points at Grohnde and Philippsburg II. (orig.) [de

  16. Genetic association analysis of 13 nuclear-encoded mitochondrial candidate genes with type II diabetes mellitus: The DAMAGE study

    DEFF Research Database (Denmark)

    Reiling, Erwin; van Vliet-Ostaptchouk, Jana V; van 't Riet, Esther

    2009-01-01

    ). After a meta-analysis, only one SNP in SIRT4 (rs2522138) remained significant (P=0.01). Extending the second stage with samples from the Danish Steno Study (n=1220 participants) resulted in a common odds ratio (OR) of 0.92 (0.85-1.00), P=0.06. Moreover, in a large meta-analysis of three genome......Mitochondria play an important role in many processes, like glucose metabolism, fatty acid oxidation and ATP synthesis. In this study, we aimed to identify association of common polymorphisms in nuclear-encoded genes involved in mitochondrial protein synthesis and biogenesis with type II diabetes...

  17. TIBER II/ETR [Engineering Test Reactor] nuclear shielding and optional tritium breeding system: An overview

    International Nuclear Information System (INIS)

    Lee, J.D.; Sawan, M.

    1987-01-01

    TIBER II, the Tokamak Ignition/Burn Experimental Reactor II, is a design concept developed as the US candidate for an International Engineering Test Reactor (ETR). An important objective of this design is to minimize cost by minimizing major radius while providing a wall loading greater than 1.0 MW/m2 and a total fluence greater than 3.0 MWY/m2 needed for blanket module testing. The shielding required for the superconducting TF coils is an important element in setting TIBER II's 3.0m major radius. 6 refs., 1 fig., 1 tab

  18. Review of first line supervisory positions in nuclear power plants - Phase II

    Energy Technology Data Exchange (ETDEWEB)

    Mackenzie, C W; Huntley, M [Hickling Corp., Ottawa, ON (Canada)

    1995-10-01

    This report provides an overview of first line supervisory activities at Ontario Hydro nuclear generating stations (Pickering `A` and Bruce `B`) and the Point Lepreau nuclear generating station in New Brunswick. Activity profiles describing the range of first line supervisory roles and responsibilities for nuclear operators have been developed from survey data and flowcharting methods. These activity profiles have then been compared with formal job responsibilities as identified in job descriptions, supervisory training provided and assessment criteria used to evaluate supervisors. Finally, this report relates the findings of supervisory practices in the group under study with the findings in the current literature relating to supervisory functioning. (author). 32 tabs., 2 figs.

  19. Review of first line supervisory positions in nuclear power plants - Phase II

    International Nuclear Information System (INIS)

    Mackenzie, C.W.; Huntley, M.

    1995-10-01

    This report provides an overview of first line supervisory activities at Ontario Hydro nuclear generating stations (Pickering 'A' and Bruce 'B') and the Point Lepreau nuclear generating station in New Brunswick. Activity profiles describing the range of first line supervisory roles and responsibilities for nuclear operators have been developed from survey data and flowcharting methods. These activity profiles have then been compared with formal job responsibilities as identified in job descriptions, supervisory training provided and assessment criteria used to evaluate supervisors. Finally, this report relates the findings of supervisory practices in the group under study with the findings in the current literature relating to supervisory functioning. (author). 32 tabs., 2 figs

  20. Nuclear Magnetic Resonance Spectrometer Console Upgrade for a Type II Quantum Computer

    National Research Council Canada - National Science Library

    Cory, David

    2003-01-01

    ...) spectrometer to enable an improved implementation of type II quantum computers (TTQC). This upgrade is fully functional and has permitted our NMR studies to be moved to higher strength magnetic fields for better sensitivity and spectral dispersion...

  1. The nuclear engineering programmes at the Royal Military College of Canada. Part II

    Energy Technology Data Exchange (ETDEWEB)

    Bonin, H.W. [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

    2002-08-01

    The coverage of the activities within the nuclear science and engineering programmes at RMC reveals the dynamism of the College which is still growing at a fast rate. Being the only completely bilingual university in Canada and a true national institution gathering students and staff from all parts of the country. RMC continues in its mission to support the Canadian Forces, the Department of National Defence, the people of Canada and Canadian Industry that includes the nuclear sector. It is in this spirit that the staff has been actively involved with organizations such as the Canadian Nuclear Society and the Canadian Nuclear Association, having hosted four of the Student conferences and three major topical conferences of the CNS.

  2. High-Speed Neutron and Gamma Flux Sensor for Monitoring Surface Nuclear Reactors, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA needs compact nuclear reactors to power future bases on the moon and/or Mars. These reactors require robust automatic control systems using low mass, rapid...

  3. Hydrogen Wave Heater for Nuclear Thermal Propulsion Component Testing, Phase II

    Data.gov (United States)

    National Aeronautics and Space Administration — NASA has identified Nuclear Thermal Propulsion (NTP) as an approach that can provide the fastest trip times to Mars and as the preferred concept for human space...

  4. GTP-dependent binding and nuclear transport of RNA polymerase II by Npa3 protein

    DEFF Research Database (Denmark)

    Staresincic, Lidija; Walker, Jane; Dirac-Svejstrup, A Barbara

    2011-01-01

    in yeast extracts. Indeed, Npa3 depletion in vivo affects nuclear localization of RNAPII; the polymerase accumulates in the cytoplasm. Npa3 is a member of the GPN-LOOP family of GTPases. Npa3 mutants that either cannot bind GTP or that bind but cannot hydrolyze it are inviable and unable to support nuclear...... transport of RNAPII. Surprisingly, we were unable to detect interactions between Npa3 and proteins in the classical importin a/ß pathway for nuclear import. Interestingly, Npa3-RNAPII binding is significantly increased by the addition of GTP or its slowly hydrolyzable analogue guanosine 5'-3-O......-(thio)triphosphate (GTP¿S). Moreover, the Npa3 mutant that binds GTP, but cannot hydrolyze it, binds RNAPII even in the absence of added GTP, whereas the mutant that cannot bind GTP is unable to bind the polymerase. Together, our data suggest that Npa3 defines an unconventional pathway for nuclear import of RNAPII, which...

  5. The nuclear engineering programmes at the Royal Military College of Canada. Part II

    International Nuclear Information System (INIS)

    Bonin, H.W.

    2002-01-01

    The coverage of the activities within the nuclear science and engineering programmes at RMC reveals the dynamism of the College which is still growing at a fast rate. Being the only completely bilingual university in Canada and a true national institution gathering students and staff from all parts of the country. RMC continues in its mission to support the Canadian Forces, the Department of National Defence, the people of Canada and Canadian Industry that includes the nuclear sector. It is in this spirit that the staff has been actively involved with organizations such as the Canadian Nuclear Society and the Canadian Nuclear Association, having hosted four of the Student conferences and three major topical conferences of the CNS

  6. Proceeding of the National Seminar on Research and Management of Nuclear Equipment: Book II

    International Nuclear Information System (INIS)

    Tjipto Sujitno; Syarip; Agus Taftazani; Elisabeth Supriyatni; Prayitno; MV Purwani; Budi Setiawan; Prajitno; Rany Saptaaji; Bambang Siswanto; Eko Priyono; Jumari

    2013-09-01

    The scientific meeting and presentation on accelerator technology and its applications was held by PTAPB BATAN on 11 September 2013. This meeting aims to promote the technology and its applications to accelerator scientists, academics, researchers and technology users as well as accelerator-based accelerator research that have been conducted by researchers in and outside BATAN. This proceeding contains 49 papers about , chemistry, physics, accelerator, nuclear instrument and nuclear reactor, etc. (PPIKSN)

  7. Laser-enhanced chemical reactions and the liquid state. II. Possible applications to nuclear fuel reprocessing

    International Nuclear Information System (INIS)

    DePoorter, G.L.; Rofer-DePoorter, C.K.

    1976-01-01

    Laser photochemistry is surveyed as a possible improvement upon the Purex process for reprocessing spent nuclear fuel. Most of the components of spent nuclear fuel are photochemically active, and lasers can be used to selectively excite individual chemical species. The great variety of chemical species present and the degree of separation that must be achieved present difficulties in reprocessing. Lasers may be able to improve the necessary separations by photochemical reaction or effects on rates and equilibria of reactions

  8. Nuclear Thermal Rocket Element Environmental Simulator (NTREES) Phase II Upgrade Activities

    Science.gov (United States)

    Emrich, William J.; Moran, Robert P.; Pearson, J. Bose

    2013-01-01

    To support the on-going nuclear thermal propulsion effort, a state-of-the-art non nuclear experimental test setup has been constructed to evaluate the performance characteristics of candidate fuel element materials and geometries in representative environments. The facility to perform this testing is referred to as the Nuclear Thermal Rocket Element Environment Simulator (NTREES). This device can simulate the environmental conditions (minus the radiation) to which nuclear rocket fuel components will be subjected during reactor operation. Test articles mounted in the simulator are inductively heated in such a manner so as to accurately reproduce the temperatures and heat fluxes which would normally occur as a result of nuclear fission and would be exposed to flowing hydrogen. Initial testing of a somewhat prototypical fuel element has been successfully performed in NTREES and the facility has now been shutdown to allow for an extensive reconfiguration of the facility which will result in a significant upgrade in its capabilities. Keywords: Nuclear Thermal Propulsion, Simulator

  9. Analysis of occupational doses in radioactive and nuclear facilities

    International Nuclear Information System (INIS)

    Curti, A.; Gomez P, I.; Pardo, G.; Thomasz, E.

    1996-01-01

    Occupational doses were analyzed in the most important nuclear and radioactive facilities in Argentina, on the period 1988-1994. The areas associated with uranium mining and milling, and medical uses of radiation facilities were excluded from this analysis. The ICRP publication 60 recommendations, adopted in 1990, and enforced in Argentine in 1994, keep the basic criteria of dose limitation system and recommend a substantial reduction in the dose limits. The reduction of the dose limits will affect the individual dose distributions, principally in those installations with occupational doses close to 50 mSv. It were analyzed Occupational doses, principally in the following facilities: Atucha-I and Embalse Nuclear Power Plants, radioisotope production plants, research reactors and radioactive waste management plants. The highest doses were identified in each facility, as well as the task associated with them. Trends in the individual dose distribution and collective and average doses were analyzed. It is concluded, that no relevant difficulties should appear in accomplishing with the basic standards for radiological safety, except for the Atucha-I Nuclear Power Plant. In this NPP a significant effort for the optimization of radiological safety procedures in order to diminish the occupational doses, and a change of the fuel channels by new ones free of cobalt are being carried out. (authors). 4 refs., 3 figs., 3 tabs

  10. A role for nuclear translocation of tripeptidyl-peptidase II in reactive oxygen species-dependent DNA damage responses

    Energy Technology Data Exchange (ETDEWEB)

    Preta, Giulio; Klark, Rainier de [Center for Molecular Medicine (CMM), Department of Medicine, Karolinska Institutet, Karolinska University Hospital, 171 76 Stockholm (Sweden); Glas, Rickard, E-mail: rickard.glas@ki.se [Center for Molecular Medicine (CMM), Department of Medicine, Karolinska Institutet, Karolinska University Hospital, 171 76 Stockholm (Sweden)

    2009-11-27

    Responses to DNA damage are influenced by cellular metabolism through the continuous production of reactive oxygen species (ROS), of which most are by-products of mitochondrial respiration. ROS have a strong influence on signaling pathways during responses to DNA damage, by relatively unclear mechanisms. Previous reports have shown conflicting data on a possible role for tripeptidyl-peptidase II (TPPII), a large cytosolic peptidase, within the DNA damage response. Here we show that TPPII translocated into the nucleus in a p160-ROCK-dependent fashion in response to {gamma}-irradiation, and that nuclear expression of TPPII was present in most {gamma}-irradiated transformed cell lines. We used a panel of nine cell lines of diverse tissue origin, including four lymphoma cell lines (T, B and Hodgkins lymphoma), a melanoma, a sarcoma, a colon and two breast carcinomas, where seven out of nine cell lines showed nuclear TPPII expression after {gamma}-irradiation. Further, this required cellular production of ROS; treatment with either N-acetyl-Cysteine (anti-oxidant) or Rotenone (inhibitor of mitochondrial respiration) inhibited nuclear accumulation of TPPII. The local density of cells was important for nuclear accumulation of TPPII at early time-points following {gamma}-irradiation (at 1-4 h), indicating a bystander effect. Further, we showed that the peptide-based inhibitor Z-Gly-Leu-Ala-OH, but not its analogue Z-Gly-(D)-Leu-Ala-OH, excluded TPPII from the nucleus. This correlated with reduced nuclear expression of p53 as well as caspase-3 and -9 activation in {gamma}-irradiated lymphoma cells. Our data suggest a role for TPPII in ROS-dependent DNA damage responses, through alteration of its localization from the cytosol into the nucleus.

  11. Cost estimation of hydrogen and DME produced by nuclear heat utilization system II

    International Nuclear Information System (INIS)

    Shiina, Yasuaki; Nishihara, Tetsuo

    2004-09-01

    Utilization and production of hydrogen has been studied in order to spread utilization of the hydrogen energy in 2020 or 2030. It will take, however, many years for the hydrogen energy to be used very easily like gasoline, diesel oil and city gas in the world. During the periods, low CO 2 release liquid fuels would be used together with hydrogen. Recently, di-methyl-ether (DME). has been noticed as one of the substitute liquid fuels of petroleum. Such liquid fuels can be produced from the mixed gas such as hydrogen and carbon oxide which are produced from natural gas by steam reforming. Therefore, the system would become one of the candidates of future system of nuclear heat utilization. Following the study in 2002, we performed economic evaluation of the hydrogen and DME production by nuclear heat utilization plant where heat generated by HTGR is completely consumed for the production. The results show that hydrogen price produced by nuclear was about 17% cheaper than the commercial price by increase in recovery rate of high purity hydrogen with increased in PSA process. Price of DME in indirect method produced by nuclear heat was also about 17% cheaper than the commercial price by producing high purity hydrogen in the DME producing process. As for the DME, since price of DME produced near oil land in petroleum exporting countries is cheaper than production in Japan, production of DME by nuclear heat in Japan has disadvantage economically in this time. Trial study to estimate DME price produced by direct method was performed. From the present estimation, utilization of nuclear heat for the production of hydrogen would be more effective with coupled consideration of reduction effect of CO 2 release. (author)

  12. Work Analysis of the nuclear power plant control room operators (II): The classes of situation

    International Nuclear Information System (INIS)

    Alengry, P.

    1989-03-01

    This report presents a work analysis of nuclear power plant control room operators focused on the classes of situation they can meet during their job. Each class of situation is first described in terms of the process variables states. We then describe the goals of the operators and the variables they process in each class of situation. We report some of the most representative difficulties encountered by the operators in each class of situation. Finally, we conclude on different topics: the nature of the mental representations, the temporal dimension, the monitoring activity, and the role of the context in the work of controlling a nuclear power plant [fr

  13. Nuclear data for reactors. Proceedings of the second international conference. Vol. II

    International Nuclear Information System (INIS)

    1970-01-01

    The Second International Conference on Nuclear Data for Reactors, held in Helsinki at the invitation of the Finnish Government, was convened by the International Atomic Energy Agency from 15 to 19 June 1970. The Conference, held as a result of recommendations made by the International Nuclear Data Committee, was attended by 163 participants from 28 countries and four international organizations, and 21 invited and 98 contributed papers were presented. This Conference was the second held by the IAEA on Nuclear Data for Reactors. Almost four years have elapsed since the first was held in Paris in 1966. During these years gratifying progress has been made by reactor, nuclear and evaluation physicists, whose collaboration has been greatly enhanced. As a result, many laboratories have concentrated their efforts on items of particular importance for reactor research and development, and many measurements are now available. The main purpose of this Conference was to provide an opportunity to review results of recent basic neutron-physics investigations against a background need for basic information, especially concerning reactors. The Conference itself, together with the preparatory meetings of IAEA experts in Studsvik on the status of α( 239 Pu) and the ν-bar-values for fissionable nuclei, showed an emphasis on the nuclear data aspects most important for nuclear technology. Most contributors dealt with the measurement and analysis of neutron cross-sections. This extensive new cross-section information can be attributed to several factors, the most important being the development and systematic exploitation of high-intensity neutron sources, such as modern linear accelerators, modern cyclotrons and underground nuclear explosions, improvements in instrumentation and in sample preparation techniques, and other technical improvements. Compared with the first IAEA Conference on Nuclear Data for Reactors this one has many more contributions on neutron data evaluation. Many

  14. Long-term consequences of and prospects for recovery from nuclear war: Two views. View II

    International Nuclear Information System (INIS)

    Anspaugh, L.R.

    1986-01-01

    The author comments on the information presented in this volume and speculates on the long-term consequences of nuclear war and the prospects for recovery. In order to do that, it might be useful to define long term. To him this means time frames of years to perhaps even hundreds of years in terms of the ultimate response and recovery of large-scale ecosystems. Such long time frames may seem excessive, but if some of the speculated efforts of nuclear war are actually realized, it may indeed take centuries before native ecosystems restabilize. Also, when referring to long-term effects of the magnitude required to have a major impact on entire ecosystems, it is clear that the driving force would not be the direct effects of nuclear war. Of potentially greater significance would be the secondary effects mediated by the intermediate-term impacts on global climate. Specifically, he refers to the speculative impacts of major decreases in the heat and light fluxes reaching the Earth's surface. Such changes are commonly referred to as ''nuclear winter.''

  15. Single Particle Potential of a Σ Hyperon in Nuclear Matter. II Rearrangement Effects

    International Nuclear Information System (INIS)

    Dabrowski, J.

    2000-01-01

    The rearrangement contribution to the real part of the single particle potential of a Σ hyperon in nuclear matter, U Σ , is investigated. The isospin and spin dependent parts of U Σ are considered. Results obtained for four models of the Nijmegen baryon-baryon interaction are presented and discussed. (author)

  16. Neuroradiology in the ocular motility disorders : II. nuclear and infranuclear pathway

    International Nuclear Information System (INIS)

    Kim, Hyung Jin; Kim, Jae Hyoung; Ha, Choong Gun; Lim, Myung Kwan; Cho, Young Kuk; Suh, Chang Hae

    1999-01-01

    The nuclear and infranuclear pathway of eye movement begins from the ocular motor nuclei situated in the brain stem, where the axons originate and form three ocular motor nerves. Although each of the ocular motor nerves follows a distinct route to reach the end organ, the extraocular muscles, they also have common housings in the cavernous sinus and at the orbital apex, where part or all of them are frequently and simultaneously affected by a common disease process. Since the fine details of normal and diseased structures can frequently be seen on radiologic imaging, especially magnetic resonance (MR) imaging, a knowledge of the basic anatomy involved in nuclear and infranuclear eye movement is important. In this description, in addition to the normal nuclear and infranuclear pathway of eye movement, we have noted the radiologic findings of typical diseases involving each segment of the nuclear and infranuclear pathway, particularly as seen on magnetic resonance images. Brief comments on ocular motor pseudopalsy, which mimics ocular motor palsy, are also included

  17. Final environmental statement. Final addendum to Part II: Manufacture of floating nuclear power plants by Offshore Power Systems. DOCKET-STN--50-437

    International Nuclear Information System (INIS)

    1978-06-01

    This Addendum to Part II of the Final Environmental Statement related to manufacture of floating nuclear power plants by Offshore Power Systems (OPS), NUREG-0056, issued September 1976, was prepared by the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Reactor Regulation. The staff's basic evaluation is presented in NUREG-0056. The current Addendum provides further consideration of a number of topics discussed in NUREG-0056, particularly additional consideration of shore zone siting at estuarine and ocean regions. This Summary and Conclusions recapitulates and is cumulative for Part II of the FES and the current Addendum. Augmentations to the Summary and Conclusions presented in Part II of the FES and arising from the evaluations contained in this Addendum are italicized

  18. Visual examination program of the TRIGA Mark II reactor Vienna with the nuclear underwater telescope

    International Nuclear Information System (INIS)

    Boeck, H.; Hammer, J.; Varga, K.

    1985-12-01

    The visual inspection programm carried out during a three month shut-period at the TRIGA Mark II reactor Vienna is described. Optical inspection of all welds inside the reactor tank was carried out with an underwater telescope developed by the Central Research Institute of Physics, Budapest, Hungary. It is shown that even after 23 years of reactor operation all tank internals were found to be in good condition and minor defects can be easily repaired by remote handling tools. (Author)

  19. The rising of the land at the nuclear power plant Asco II

    International Nuclear Information System (INIS)

    Garcia Sanchez, J.; Ubalde, L.

    1997-01-01

    The rising of the land on which Asco II is settled is a geological process of slow and falling evolution in the time, studied and followed by means of geophysical models of the underground and continuous auscultation. Their influence on the structures, equipment, components and systems is contemplated in the bases of the design of the power station, whose behaviour towards this phenomenon evolves satisfactorily in accordance with control parameters. (Author)

  20. Relocalization of nuclear DNA helicase II during the growth period of bovine oocytes

    Czech Academy of Sciences Publication Activity Database

    Baran, V.; Kovářová, Hana; Klíma, Jiří; Hozák, Pavel; Motlík, Jan

    2006-01-01

    Roč. 125, 1-2 (2006), s. 155-164 ISSN 0948-6143 R&D Projects: GA ČR GA523/03/0857 Grant - others:Slovenská Akademie věd(SK) VEGA 2/3065/23 Institutional research plan: CEZ:AV0Z50450515; CEZ:AV0Z50390512 Keywords : DNA helicase II * fibroblasts * oocytes Subject RIV: EB - Genetics ; Molecular Biology Impact factor : 3.220, year: 2006

  1. Analytical mass formula and nuclear surface properties in the ETF approximation. Part II: asymmetric nuclei

    Science.gov (United States)

    Aymard, François; Gulminelli, Francesca; Margueron, Jérôme

    2016-08-01

    We have recently addressed the problem of the determination of the nuclear surface energy for symmetric nuclei in the framework of the extended Thomas-Fermi (ETF) approximation using Skyrme functionals. We presently extend this formalism to the case of asymmetric nuclei and the question of the surface symmetry energy. We propose an approximate expression for the diffuseness and the surface energy. These quantities are analytically related to the parameters of the energy functional. In particular, the influence of the different equation of state parameters can be explicitly quantified. Detailed analyses of the different energy components (local/non-local, isoscalar/isovector, surface/curvature and higher order) are also performed. Our analytical solution of the ETF integral improves previous models and leads to a precision of better than 200 keV per nucleon in the determination of the nuclear binding energy for dripline nuclei.

  2. Health effects models for off-site radiological consequence analysis on nuclear reactor accidents (II)

    Energy Technology Data Exchange (ETDEWEB)

    Homma, Toshimitsu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Takahashi, Tomoyuki [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst; Yonehara, Hidenori [National Inst. of Radiological Sciences, Chiba (Japan)] [eds.

    2000-12-01

    This report is a revision of JAERI-M 91-005, 'Health Effects Models for Off-Site Radiological Consequence Analysis of Nuclear Reactor Accidents'. This revision provides a review of two revisions of NUREG/CR-4214 reports by the U.S. Nuclear Regulatory Commission which is the basis of the JAERI health effects models and other several recent reports that may impact the health effects models by international organizations. The major changes to the first version of the JAERI health effects models and the recommended parameters in this report are for late somatic effects. These changes reflect recent changes in cancer risk factors that have come from longer followup and revised dosimetry in major studies on the Japanese A-bomb survivors. This report also provides suggestions about future revisions of computational aspects on health effects models. (author)

  3. Health effects models for off-site radiological consequence analysis on nuclear reactor accidents (II)

    International Nuclear Information System (INIS)

    Homma, Toshimitsu

    2000-12-01

    This report is a revision of JAERI-M 91-005, 'Health Effects Models for Off-Site Radiological Consequence Analysis of Nuclear Reactor Accidents'. This revision provides a review of two revisions of NUREG/CR-4214 reports by the U.S. Nuclear Regulatory Commission which is the basis of the JAERI health effects models and other several recent reports that may impact the health effects models by international organizations. The major changes to the first version of the JAERI health effects models and the recommended parameters in this report are for late somatic effects. These changes reflect recent changes in cancer risk factors that have come from longer followup and revised dosimetry in major studies on the Japanese A-bomb survivors. This report also provides suggestions about future revisions of computational aspects on health effects models. (author)

  4. Technical improvement of ATE system of Ling'ao Nuclear Power Plant Phase II

    International Nuclear Information System (INIS)

    Zhu Xingbao; Xiong Jingchuan; Liang Qiaohong

    2009-01-01

    In order to solve the problem that the content of SO 4 2- in Steam Generator significantly increased beyond the criteria after the use of the condensate treatment (ATE) system in Daya Bay Nuclear Power Plant and Ling'ao Nuclear Power Plant Phase I, technical improvement have been conducted on the sizes of the fore cation bed and the mixed bed, water distributing devices, ion exchange resins and separation facility. The effectiveness for the ion exchange of the mixed bed is improved, the resolved substance of cation resin is decreased; it is more impossible for fragments and powder which would lead high SO 4 2- content in Steam Generator. Finally, the quality of the steam-water could be improved and ensured. (authors)

  5. Underwater Nuclear Fuel Disassembly and Rod Storage Process and Equipment Description. Volume II

    International Nuclear Information System (INIS)

    Viebrock, J.M.

    1981-09-01

    The process, equipment, and the demonstration of the Underwater Nuclear Fuel Disassembly and Rod Storage System are presented. The process was shown to be a viable means of increasing spent fuel pool storage density by taking apart fuel assemblies and storing the fuel rods in a denser fashion than in the original storage racks. The assembly's nonfuel-bearing waste is compacted and containerized. The report documents design criteria and analysis, fabrication, demonstration program results, and proposed enhancements to the system

  6. Planning and execution of jobs in nuclear power plants according to IWRS II. Pt. 1

    International Nuclear Information System (INIS)

    Hilmer, R.; Hauck, W.

    1986-01-01

    This paper deals with various efforts taken for a reduction of the radiation exposure of nuclear power plant staff. Health physics considerations range from the design of plants to include components requiring less time for maintenance and repair, to plant operation and optimization of job processes with special consideration of requirements made by radiation protection and maintenance. The paper presents in detail legal provisions and their translation into job process organization, job assignment and job schedulling. (HAG) [de

  7. II. congress of Czechoslovak Society of Nuclear Medicine and Radiation Hygiene

    International Nuclear Information System (INIS)

    1986-01-01

    The proceedings contain 165 abstracts of papers covering all areas of the application of nuclear medicine, such as osteology, cardiology, immunology, neurology, oncology, etc. The topics include the examination of the skeleton with radioisotopes, various immunology methods, scintiscanning of body organs, tumor monitoring, radiopharmacology aspects, biological radiation effects, cytogenetic changes following irradiation, and studies of radiation effects on DNA repair. Separate volumes are devoted to education of technicians, processing of radioisotope examination data, radiation protection and decontamination. (M.D.)

  8. Nuclear energy and public safety (Part II): a bibliography of technical resources

    International Nuclear Information System (INIS)

    Gabriel, M.R.

    1982-01-01

    Part 2 of the bibliography focuses on technical information of interest to those concerned with the operation of nuclear power plants and the subjects of safety and accidents. A subject index included after the bibliography provides a breakdown of the references into seven categories. There is also an author index. The material cited is available through the National Technical Information Service (NTIS) in Springfield, Virginia

  9. Meson dynamics and the nuclear many-body problem. II. Finite density Hartree-Fock

    International Nuclear Information System (INIS)

    Wilets, L.; Puff, R.D.; Chiang, D.; Nutt, W.T.

    1976-01-01

    The field-theoretic many-nucleon problem is formulated, and an analysis which sums all ''uncrossed meson line'' diagrams is investigated in detail. The calculation of energy per nucleon, after proper identification of infinite mass renormalization terms, exhibits effects of nuclear recoil, relativistic kinematics, and retardation. Numerical results are presented for π and ω mesons, and the nucleon interaction energies obtained are compared with the traditional static limit of infinite nucleon mass

  10. Nuclear Energy Center study. Phase II. Site suitability analysis. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Fellows, W.S.; Sharp, J.M.; Benator, B.I.

    1978-06-01

    A site screening study was conducted to identify a site or sites for detailed, site-specific study as a nuclear energy center. Using technical criteria of water requirements, geotechnical constraints, and projected load center and transmission considerations as well as environmental and institutional considerations, five potential study sites in the State of South Carolina were identified, evaluated against established criteria, and ranked according to their acceptability as potential nuclear energy center study sites. Consideration of what is ''representative'' of a site as well as the ranking score was factored into site recommendations, since the site deemed easiest to license and permit may not be the most desirable site for future study of the technical and institutional feasibility and practicality of a specific site. The sites near Lake Hartwell and the Savannah River Plant (SRP) of the Department of Energy were selected as potential study sites after consideration of the above criteria. Because the Lake Hartwell site offers the opportunity to consider institutional issues which may be more representative of other possible NEC sites, it is recommended that the Lake Hartwell site be studied to establish the feasibility and practicality of the nuclear energy concept on a site-specific basis.

  11. Probabilistic risk assessment on maritime spent nuclear fuel transportation (Part II: Ship collision probability)

    International Nuclear Information System (INIS)

    Christian, Robby; Kang, Hyun Gook

    2017-01-01

    This paper proposes a methodology to assess and reduce risks of maritime spent nuclear fuel transportation with a probabilistic approach. Event trees detailing the progression of collisions leading to transport casks’ damage were constructed. Parallel and crossing collision probabilities were formulated based on the Poisson distribution. Automatic Identification System (AIS) data were processed with the Hough Transform algorithm to estimate possible intersections between the shipment route and the marine traffic. Monte Carlo simulations were done to compute collision probabilities and impact energies at each intersection. Possible safety improvement measures through a proper selection of operational transport parameters were investigated. These parameters include shipment routes, ship's cruise velocity, number of transport casks carried in a shipment, the casks’ stowage configuration and loading order on board the ship. A shipment case study is presented. Waters with high collision probabilities were identified. Effective range of cruising velocity to reduce collision risks were discovered. The number of casks in a shipment and their stowage method which gave low cask damage frequencies were obtained. The proposed methodology was successful in quantifying ship collision and cask damage frequency. It was effective in assisting decision making processes to minimize risks in maritime spent nuclear fuel transportation. - Highlights: • Proposes a probabilistic framework on the safety of spent nuclear fuel transportation by sea. • Developed a marine traffic simulation model using Generalized Hough Transform (GHT) algorithm. • A transportation case study on South Korean waters is presented. • Single-vessel risk reduction method is outlined by optimizing transport parameters.

  12. The downstream side of the nuclear fuel cycle. Tome II: Electricity generating costs

    International Nuclear Information System (INIS)

    Bataille, Ch.; Galley, R.

    1999-01-01

    As part of the Office's continuing work in the nuclear field, Mr. Christian Bataille and Mr. Robert Galley, Members of Parliament for the Nord and Aube departements respectively, published in June 1998 the first part of their investigation into the downstream side of the nuclear fuel cycle, focusing on the work done in application of the law of 30 December 1991 concerning research into radioactive waste management. This document supplements that initial technical approach with a technical and economic study of the costs of generating electricity. To begin with, the performance of existing nuclear generating plant is examined, in particular the past, present and future contributions of this plant to the growth and competitiveness of the French economy. Secondly, the competitiveness of the different generating systems is analysed with a view to the construction of new facilities, using the method of discounted average costs which is at present the standard approach governing investment decisions, and identifying the different ways in which the said systems are dealt with as regards the cost categories considered. The potential contributions of external factor analysis and the calculation of external costs are then reviewed in order to evaluate the advantages and drawbacks of the different electricity generating systems on a more global basis. The report includes more than a hundred tables of data and cost curves upon which the Rapporteurs base their comments, conclusions and recommendations

  13. Nuclear Energy Center study. Phase II. Site suitability analysis. Final report

    International Nuclear Information System (INIS)

    Fellows, W.S.; Sharp, J.M.; Benator, B.I.

    1978-06-01

    A site screening study was conducted to identify a site or sites for detailed, site-specific study as a nuclear energy center. Using technical criteria of water requirements, geotechnical constraints, and projected load center and transmission considerations as well as environmental and institutional considerations, five potential study sites in the State of South Carolina were identified, evaluated against established criteria, and ranked according to their acceptability as potential nuclear energy center study sites. Consideration of what is ''representative'' of a site as well as the ranking score was factored into site recommendations, since the site deemed easiest to license and permit may not be the most desirable site for future study of the technical and institutional feasibility and practicality of a specific site. The sites near Lake Hartwell and the Savannah River Plant (SRP) of the Department of Energy were selected as potential study sites after consideration of the above criteria. Because the Lake Hartwell site offers the opportunity to consider institutional issues which may be more representative of other possible NEC sites, it is recommended that the Lake Hartwell site be studied to establish the feasibility and practicality of the nuclear energy concept on a site-specific basis

  14. Qinshan phase II extension nuclear power project thermal stratification and fatigue stress analysis for pressurizer surge line

    International Nuclear Information System (INIS)

    Yu Xiaofei; Zhang Yixiong; Ai Honglei

    2010-01-01

    Thermal stratification of pressurizer surge line induced by the inside fluid brings on global bending moments, local thermal stresses, unexpected displacements and support loadings of the pipe system. In order to avoid a costly three-dimensional computation, a combined 1D/2D technique has been developed and implemented to analyze the thermal stratification and fatigue stress of pressurize surge line of QINSHAN Phase II Extension Nuclear Power Project in this paper, using the computer codes SYSTUS and ROCOCO. According to the mechanical analysis results of stratification, the maximum stress and cumulative usage factor, the loadings at connections of surge line to main pipe and RCP and the displacements of surge line at supports are obtained. (authors)

  15. Geological disposal of nuclear waste: II. From laboratory data to the safety analysis – Addressing societal concerns

    International Nuclear Information System (INIS)

    Grambow, Bernd; Bretesché, Sophie

    2014-01-01

    Highlights: • Models for repository safety can only partly be validated. • Long term risks need to be translated in the context of societal temporalities. • Social sciences need to be more strongly involved into safety assessment. - Abstract: After more than 30 years of international research and development, there is a broad technical consensus that geologic disposal of highly-radioactive waste will provide for the safety of humankind and the environment, now, and far into the future. Safety analyses have demonstrated that the risk, as measured by exposure to radiation, will be of little consequence. Still, there is not yet an operating geologic repository for highly-radioactive waste, and there remains substantial public concern about the long-term safety of geologic disposal. In these two linked papers, we argue for a stronger connection between the scientific data (paper I, Grambow et al., 2014) and the safety analysis, particularly in the context of societal expectations (paper II). In this paper (II), we assess the meaning of the technical results and derived models (paper I) for the determination of the long-term safety of a repository. We consider issues of model validity and their credibility in the context of a much broader historical, epistemological and societal context. Safety analysis is treated in its social and temporal dimensions. This perspective provides new insights into the societal dimension of scenarios and risk analysis. Surprisingly, there is certainly no direct link between increased scientific understanding and a public position for or against different strategies of nuclear waste disposal. This is not due to the public being poorly informed, but rather due to cultural cognition of expertise and historical and cultural perception of hazards to regions selected to host a geologic repository. The societal and cultural dimension does not diminish the role of science, as scientific results become even more important in distinguishing

  16. Computational analysis of neutronic parameters for TRIGA Mark-II research reactor using evaluated nuclear data libraries

    International Nuclear Information System (INIS)

    Uddin, M.N.; Sarker, M.M.; Khan, M.J.H.; Islam, S.M.A.

    2010-01-01

    The aim of this study is to analyze the neutronic parameters of TRIGA Mark-II research reactor using the chain of NJOY-WIMS-CITATION computer codes based on evaluated nuclear data libraries CENDL-2.2 and JEFF-3.1.1. The nuclear data processing code NJOY99.0 has been employed to generate the 69 group WIMS library for the isotopes of TRIGA core. The cell code WIMSD-5B was used to generate the cross sections in CITATION format and then 3-dimensional diffusion code CITTATION was used to calculate the neutronic parameters of the TRIGA Mark-II research reactor. All the analyses were performed using the 7-group macroscopic cross section library. The CITATION test-runs using different cross section sets based on different models applied in WIMS calculations have shown a strong influence of those models on the final integral parameters. Some of the cells were specially treated with PRIZE options available in WIMSD-5B to take into account the fine structure of the flux gradient in the fuel-reflector interface region. It was observed that two basic parameters, the effective multiplication factor, k eff and the thermal neutron flux, were in good agreement among the calculated results with each other as well as the measured values. The maximum power densities at the hot spot were 1.0446E02 W/cc and 1.0426E02 W/cc for the libraries CENDL-2.2 and JEFF-3.1.1 respectively. The calculated total peaking factors 5.793 and 5.745 were compared to the original SAR value of 5.6325 as well as MCNP result. Consequently, this analysis will be helpful to enhance the neutronic calculations and also be used for the further thermal-hydraulics study of the TRIGA core.

  17. Poisson regression analysis of the mortality among a cohort of World War II nuclear industry workers

    International Nuclear Information System (INIS)

    Frome, E.L.; Cragle, D.L.; McLain, R.W.

    1990-01-01

    A historical cohort mortality study was conducted among 28,008 white male employees who had worked for at least 1 month in Oak Ridge, Tennessee, during World War II. The workers were employed at two plants that were producing enriched uranium and a research and development laboratory. Vital status was ascertained through 1980 for 98.1% of the cohort members and death certificates were obtained for 96.8% of the 11,671 decedents. A modified version of the traditional standardized mortality ratio (SMR) analysis was used to compare the cause-specific mortality experience of the World War II workers with the U.S. white male population. An SMR and a trend statistic were computed for each cause-of-death category for the 30-year interval from 1950 to 1980. The SMR for all causes was 1.11, and there was a significant upward trend of 0.74% per year. The excess mortality was primarily due to lung cancer and diseases of the respiratory system. Poisson regression methods were used to evaluate the influence of duration of employment, facility of employment, socioeconomic status, birth year, period of follow-up, and radiation exposure on cause-specific mortality. Maximum likelihood estimates of the parameters in a main-effects model were obtained to describe the joint effects of these six factors on cause-specific mortality of the World War II workers. We show that these multivariate regression techniques provide a useful extension of conventional SMR analysis and illustrate their effective use in a large occupational cohort study

  18. Communication and surrounding of Asco and Vandellos II nuclear power plants

    International Nuclear Information System (INIS)

    2004-01-01

    The Asco and Vandellos-II power plants have always been integrated into the regions where they are located, and they take an active part in the development of surrounding towns through quality employment provided by our facilities, social and cultural support and aid to development promoted by regional councils. Communication to media is a corporate priority defined in our strategic plant, to ensure openness, rigour and punctuality. We also attend to the visitors who want to learn more about our facilities in the visitor center, and we have agreements with agrarian institutions in the area so that students can de practical training in the farms we own for agricultural production. (Author)

  19. Beznau II nuclear power plant: Expertise on NOK's request for the removal of the time limitation for the operation licence; KKW Beznau II: Gutachten zum Gesuch der NOK um Aufhebung der Befristung der Betriebsbewilligung

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-03-15

    The Federal Agency for the Safety of Nuclear Installations (HSK) is the Swiss authority responsible for nuclear safety and protection against radioactivity in nuclear power plants. It has to examine the request of the North-East Swiss Power Corporation (NOK) concerning the removal of the operational time limitation for the Beznau-II reactor (KKB-II). In the present report HSK reviews the enterprise management and the safety of KKB-II on the basis of the results of the Periodic Safety Review. The Beznau nuclear power plant exhibits a very high degree of technical and organisational safety. During the past 10 years the plant has been operated in a safe manner. At the same time the plant has been improved and this guarantees that the mechanisms of ageing degradation are systematically identified and that measures can be taken that are possibly necessary. Under such conditions, the safety of KKB-II can be guarantied at all times. As a result of the management of quality, environmental and working safety conditions, the correct application and the continuous improvement of all processes important to safety are ensured. With these measures KKB has shown that safety is given priority over and against all other working goals. The examination by HSK of the Periodic Safety Review has shown that, in the past, KKB has applied modernisation measures independent of the licensing situation of the two reactor blocks. These modernisation measures largely contribute to the fact that the HSK examination did not reveal any significant safety deficiencies. Other improvement measures allow risk reduction or can bee seen as an adaptation to experience gained and to the state of the technological art. In conclusion, HSK states that no safety-relevant facts have been found which could prevent the removal of the time limitation on the operational licence for KKB-II. From the point of view of HSK, KKB-II fulfils the conditions for the safe continuation of operation

  20. Recommended criteria for the evaluation of on-site nuclear power plant emergency plans, volume II: criteria

    International Nuclear Information System (INIS)

    1997-01-01

    A critical review of existing Canadian and international nuclear power plant (NPP) emergency plans, evaluation criteria, and approaches has been conducted to provide AECB staff with information which can be used to assess the adequacy of NPP on-site emergency response plans. The results of this work are published in two volumes. Volume I, Basis Document, provides the reasons why certain requirements are in place. It also gives comprehensive references to various standards.Volume II, Criteria, contains the criteria which relate to on-site actions and their integration with control room activities and the roles of off-site responsible organizations. The recommended criteria provide information on what is required, and not on how to accomplish the requirements. The licensees are given the latitude to decide on the methods and processes needed to meet the requirements. The documents do not address NPP off-site plans and response capability, or the control room emergency operating procedures and response capability. This report contains only Volume II: Criteria. 55 refs., 2 tabs., 1 fig

  1. The Nuclear Receptor, Nor-1, Markedly Increases Type II Oxidative Muscle Fibers and Resistance to Fatigue

    OpenAIRE

    Pearen, Michael A.; Eriksson, Natalie A.; Fitzsimmons, Rebecca L.; Goode, Joel M.; Martel, Nick; Andrikopoulos, Sofianos; Muscat, George E. O.

    2012-01-01

    Nuclear hormone receptors (NR) have been implicated as regulators of lipid and carbohydrate metabolism. The orphan NR4A subgroup has emerged as regulators of metabolic function. Targeted silencing of neuron-derived orphan receptor 1 (Nor-1)/NR4A3 in skeletal muscle cells suggested that this NR was necessary for oxidative metabolism in vitro. To investigate the in vivo role of Nor-1, we have developed a mouse model with preferential expression of activated Nor-1 in skeletal muscle. In skeletal...

  2. AIRDOS-II computer code for estimating radiation dose to man from airborne radionuclides in areas surrouding nuclear facilities

    International Nuclear Information System (INIS)

    Moore, R.E.

    1977-04-01

    The AIRDOS-II computer code estimates individual and population doses resulting from the simultaneous atmospheric release of as many as 36 radionuclides from a nuclear facility. This report describes the meteorological and environmental models used is the code, their computer implementation, and the applicability of the code to assessments of radiological impact. Atmospheric dispersion and surface deposition of released radionuclides are estimated as a function of direction and distance from a nuclear power plant or fuel-cycle facility, and doses to man through inhalation, air immersion, exposure to contaminated ground, food ingestion, and water immersion are estimated in the surrounding area. Annual doses are estimated for total body, GI tract, bone, thyroid, lungs, muscle, kidneys, liver, spleen, testes, and ovaries. Either the annual population doses (man-rems/year) or the highest annual individual doses in the assessment area (rems/year), whichever are applicable, are summarized in output tables in several ways--by nuclides, modes of exposure, and organs. The location of the highest individual doses for each reference organ estimated for the area is specified in the output data

  3. Synthetic seismograms - II. Synthesis of amplitude spectra and seismograms of P waves from underground nuclear explosions

    International Nuclear Information System (INIS)

    Banghar, A.R.

    1980-01-01

    As a part of programme of seismic detection of underground nuclear explosions, step by step variations in the amplitude spectra and waveforms of P wave signal, as it propagates from source to receiver region, are investigated. Influences on the amplitude spectra and waveforms of teleseismic p waves due to : (1) variation in the shape of reduced displacement potential, (2) variation of mantle Q values, (3) change in depth, (4) various yields, (5) spalling, and (6) variation of crustal structure at source as well as at receiver are studied. The results show that for a yield of 85 kilotons, the time structure of seismograms is nearly same for four types of reduced displacement potentials considered here. The duration of waveforms is affected both by crustal structure at source as well as due to spalling. In general, effect of receiver crust on seismograms is found to be minor. Synthesized and observed P wave seismograms for Longshot, Milrow and Cannikin underground nuclear explosions are computed at various seismometer array stations of the UKAEA. Computed seismograms compare well with the recorded ones. It is seen that: (1) overburden P wave velocity inferred from seismograms is less as compared to its value obtained from on-site measurements, and (2) the source function, the source crust transfer function, the mantle transfer function and the spalling function are the most important factors that influence shaping of spectra and seismograms. (M.G.B.)

  4. Saturne II: a 3 GeV proton synchrotron for nuclear physics

    International Nuclear Information System (INIS)

    Faure, J.; Penicaud, J.P.

    1978-01-01

    A 3 GeV proton Synchrotron is now under completion at the Saclay Nuclear Research Center in France. This machine replaces the former Saturne Synchrotron built in 1958. The lattice type of the new machine is a strong focusing one, and the structure of the magnetic ring is made up of 16 bending magnets and 24 quadrupolar lenses. Due to the small injection energy (20 MeV), it has been necessary to design large aperture magnets. The two accelerating R.F. cavities need a wide range of tuning by ferrites from 0,86 to 8,3 MHz with a peak voltage of 18 kV. The performances of the new machine are better adaptated to the needs of Nuclear Physics. The main features of the extracted protons beam are an intensity of 2.10 12 protons per second at a variable energy from 0,5 to 3 GeV, an energy spread of a few 10 -4 and a small emittance (horizontal approximately 6 π mm.mrd, vertical 25 π mm.mrd). Heavy ions up to N 7+ and polarized particles (H + and D + ) will be accelerated too, around 10 9 per pulse on the target. On the experimental areas nine lines are fully equipped and four spectrometers will be set up. The first accelerated beam is expected in October 1978, and the physics experiments should start at the end of this year

  5. Saturne II: A 3 GeV proton synchrotron for nuclear physics

    Energy Technology Data Exchange (ETDEWEB)

    Faure, J; Penicaud, J P [Centre detude nucleaire de Saclay, Gif sur Yvette (France)

    1978-07-01

    A 3 GeV proton Synchrotron is now under completion at the Saclay Nuclear Research Center in France. This machine replaces the former Saturne Synchrotron built in 1958. The lattice type of the new machine is a strong focusing one, and the structure of the magnetic ring is made up to 16 bending magnets and 24 quadrupolar lenses. Due to the small injection energy (20 MeV), it has been necessary to design large aperture magnets. The two accelerating R.F. cavities need a wide range of tuning by ferrites from 0.86 to 8.3 MHz with a peak voltage 18 kV. The performances of the new machine are better adapted to the needs of Nuclear Physics. The main features of the extracted protons beam are an intensity of 2.10{sup 12} protons per second at a variable energy from 0.5 to 3 GeV, an energy spread of a few 10{sup -4} and a small emittance (horizontal {approx_equal} 6 {pi} mm.mrd, vertical 25 {pi} mm.mrad). Heavy ions up to N{sup 7+} and polarized particles (H{sup +} and D{sup +}) will be accelerated too, around 10{sup 9} per pulse on the target. On the experimental areas nine lines are fully equipped and four spectrometers will be set up. The first accelerated beam is expected in October 1978, and the physics experiments should start at the end of this year. (author)

  6. Nuclear Energy Center Site Survey, 1975. Part II. The U.S. electric power system and the potential role of nuclear energy centers

    International Nuclear Information System (INIS)

    1976-01-01

    Information related to Nuclear Energy Centers (NEC) in the U.S. is presented concerning the U.S. electric power system today; electricity demand history and forecasts; history and forecasts of the electric utility industry; regional notes; the status, history, and forecasts of the nuclear role; power plant siting problems and practices; nuclear facilities siting problems and practices; origin and evolution of the nuclear energy center concept; conceptualized description of nuclear energy centers; potential role of nuclear energy centers; assumptions, criteria, and bases; typical evolution of a nuclear energy center; and the nuclear fuel cycle

  7. In the matter of the application of the Westinghouse Electric Corporation for the export of pressurized water reactor to Asociacion Nuclear ASCO II, Barcelona, Spain

    International Nuclear Information System (INIS)

    Rowden, M.A.; Mason, E.A.; Gilinsky, V.; Kennedy, R.T.

    1976-01-01

    The paper contains the text of a decision of the US NRC that the export of the ASCO nuclear power unit II to Spain would not be inimical to the common defense and security of the United States, so that there are no objections to issue the license to Westinghouse Electric Corporation. Furthermore the paper contains the dissenting opinion of Commissioner Gilinsky. (HP) [de

  8. A probabilistic seismic risk assessment procedure for nuclear power plants: (II) Application

    Science.gov (United States)

    Huang, Y.-N.; Whittaker, A.S.; Luco, N.

    2011-01-01

    This paper presents the procedures and results of intensity- and time-based seismic risk assessments of a sample nuclear power plant (NPP) to demonstrate the risk-assessment methodology proposed in its companion paper. The intensity-based assessments include three sets of sensitivity studies to identify the impact of the following factors on the seismic vulnerability of the sample NPP, namely: (1) the description of fragility curves for primary and secondary components of NPPs, (2) the number of simulations of NPP response required for risk assessment, and (3) the correlation in responses between NPP components. The time-based assessment is performed as a series of intensity-based assessments. The studies illustrate the utility of the response-based fragility curves and the inclusion of the correlation in the responses of NPP components directly in the risk computation. ?? 2011 Published by Elsevier B.V.

  9. Multifragmentation of a very heavy nuclear system (II): bulk properties and spinodal decomposition

    Energy Technology Data Exchange (ETDEWEB)

    Frankland, J.D.; Rivet, M.F.; Borderie, B. [Paris-11 Univ., Inst. de Physique Nucleaire, 91 - Orsay (France)] [and others

    2000-07-01

    The properties of fragments and light charged particles emitted in multifragmentation of single sources formed in central 36 A.MeV Gd+U collisions are reviewed. Most of the products are isotropically distributed in the reaction c.m. Fragment kinetic energies reveal the onset of radial collective energy. A bulk effect is experimentally evidenced from the similarity of the charge distribution with that from the lighter 32 A.MeV Xe+Sn system. Spinodal decomposition of finite nuclear matter exhibits the same property in simulated central collisions for the two systems, and appears therefore as a possible mechanism at the origin of multifragmentation in this incident energy domain. (authors)

  10. Optimization in the nuclear fuel cycle II: Concentration of alpha emitters in the air

    International Nuclear Information System (INIS)

    Pereira, W.S.; Silva, A.X.; Lopes, J.M.; Carmo, A.S.; Mello, C.R.; Fernandes, T.S.; Kelecom, A.

    2017-01-01

    Optimization is one of the bases of radioprotection and aims to move doses away from the dose limit that is the borderline of acceptable radiological risk. The work aims to use the monitoring of the concentration of alpha emitters in the air as a tool of the optimization process. We analyzed 27 sampling points of airborne alpha concentration in a nuclear fuel cycle facility. The monthly averages were considered statistically different, the highest in the month of February and the lowest in the month of August. All other months were found to have identical mean activity concentration values. Regarding the sampling points, the points with the highest averages were points 12, 15 and 9. These points were indicated for the beginning of the optimization process. Analysis of the production of the facility should be performed to verify possible correlations between production and concentration of alpha emitters in the air

  11. Multifragmentation of a very heavy nuclear system (II): bulk properties and spinodal decomposition

    International Nuclear Information System (INIS)

    Frankland, J.D.; Rivet, M.F.; Borderie, B.

    2000-01-01

    The properties of fragments and light charged particles emitted in multifragmentation of single sources formed in central 36 A.MeV Gd+U collisions are reviewed. Most of the products are isotropically distributed in the reaction c.m. Fragment kinetic energies reveal the onset of radial collective energy. A bulk effect is experimentally evidenced from the similarity of the charge distribution with that from the lighter 32 A.MeV Xe+Sn system. Spinodal decomposition of finite nuclear matter exhibits the same property in simulated central collisions for the two systems, and appears therefore as a possible mechanism at the origin of multifragmentation in this incident energy domain. (authors)

  12. Reactor dynamics experiment of nuclear ship Mutsu using pseudo random signal (II). The second experiment

    International Nuclear Information System (INIS)

    Hayashi, Koji; Shimazaki, Junya; Nabeshima, Kunihiko; Ochiai, Masaaki; Shinohara, Yoshikuni; Inoue, Kimihiko.

    1995-01-01

    In order to investigate dynamics of the reactor plant of the nuclear ship Mutsu, the second reactor noise experiment using pseudo random binary sequences (PRBS) was performed on August 30, 1991 in the third experimental navigation. The experiments using both reactivity and load disturbances were performed at 50% of reactor power and under a quiet sea condition. Each PRBS was applied by manual operation of the control rod or the main steam valve. Various signals of the plant responses and of the acceleration of ship motion were measured. Furthermore, natural reactor noise signals were measured after each PRBS experiment in order to evaluate the effects of the PRBS disturbances. This paper summarizes the planning of the experiment, the instruction for the experiment and logs, the data recording conditions, recorded signal wave forms and the results of power spectral analysis. (author)

  13. In situ leaching of a nuclear rubblized copper ore body. Volume II

    International Nuclear Information System (INIS)

    1975-06-01

    This volume contains detailed descriptions of technical and economical evaluations undertaken for the feasibility study. A summary of these results can be found in Vol. 1 along with the conclusions derived from the feasibility study and the recommendations tendered for future work. The sections of this study are presented in process order, and each section is complete in itself. The form of the presentation, hopefully, is logical and in a manner suitable for design purposes. As a further aid, each section has its own table of contents. The sections presented include method of attack, reference case, description of concept, nuclear rubblization, blasting plan, underground plumbing, fluid circulation, leaching technology, wellhead plant and pipeline, process plant, material and heat balance, hydrology, radioactivity, seismic, economics, sensitivity analysis, guide for environmental studies, exploration, and recommended experimental program. (U.S.)

  14. De "átomos para la paz" a los reactores de potencia: Tecnología y política nuclear en la Argentina (1955-1976)

    OpenAIRE

    Hurtado de Mendoza, Diego

    2005-01-01

    Durante el período 1955-76, el programa nuclear argentino se integró a la arena internacional; su Comisión Nacional de Energía Atómica construyó cuatro reactores de investigación, adquirió a una empresa alemana y puso en marcha el primer reactor de potencia Atucha I, y compró a una empresa canadiense un segundo reactor de potencia. En este artículo se examinan estos desarrollos en relación con el contexto político local y con el panorama nuclear internacional. En particular, se analizan la po...

  15. Enhancement of organizational resilience in light of the Fukushima Dai-ichi Nuclear Power Plant accident (4). Consideration of nurturing attitude to achieve the safety-II

    International Nuclear Information System (INIS)

    Oba, Kyoko; Yoshizawa, Atsufumi; Kitamura, Masaharu

    2015-01-01

    The Fukushima nuclear plant accident has been examined aiming at clarifying the factors influencing responding which is one of the four cornerstones of resilience engineering. Among the causal factors of responding, such as attitude, skill, health and environment, particular attention has been paid on the role of attitude. In addition, the case of Tokai Dai-ni nuclear plant, which was the success case despite tsunami attack, and the case of Fukushima Dai-ichi nuclear plant have been examined focusing on preparatory actions taken prior to the tsunami attack. Through the comparative examinations, attitudes of several kinds have been identified as key factors contributing to enhance organizational resilience. Moreover, the importance of safety-II concept proposed in conjunction with the methodology of resilience engineering has been clearly exemplified. As a whole, it can be concluded that the methodology of resilience engineering and the concept of safety-II are quite effective when utilized with the structured model. (author)

  16. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    International Nuclear Information System (INIS)

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy's (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher's workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead

  17. The European Research on Severe Accidents in Generation-II and -III Nuclear Power Plants

    Directory of Open Access Journals (Sweden)

    Jean-Pierre Van Dorsselaere

    2012-01-01

    Full Text Available Forty-three organisations from 22 countries network their capacities of research in SARNET (Severe Accident Research NETwork of excellence to resolve the most important remaining uncertainties and safety issues on severe accidents in existing and future water-cooled nuclear power plants (NPP. After a first project in the 6th Framework Programme (FP6 of the European Commission, the SARNET2 project, coordinated by IRSN, started in April 2009 for 4 years in the FP7 frame. After 2,5 years, some main outcomes of joint research (modelling and experiments by the network members on the highest priority issues are presented: in-vessel degraded core coolability, molten-corium-concrete-interaction, containment phenomena (water spray, hydrogen combustion…, source term issues (mainly iodine behaviour. The ASTEC integral computer code, jointly developed by IRSN and GRS to predict the NPP SA behaviour, capitalizes in terms of models the knowledge produced in the network: a few validation results are presented. For dissemination of knowledge, an educational 1-week course was organized for young researchers or students in January 2011, and a two-day course is planned mid-2012 for senior staff. Mobility of young researchers or students between the European partners is being promoted. The ERMSAR conference is becoming the major worldwide conference on SA research.

  18. The Bologna complete sample of nearby radio sources. II. Phase referenced observations of faint nuclear sources

    Science.gov (United States)

    Liuzzo, E.; Giovannini, G.; Giroletti, M.; Taylor, G. B.

    2009-10-01

    Aims: To study statistical properties of different classes of sources, it is necessary to observe a sample that is free of selection effects. To do this, we initiated a project to observe a complete sample of radio galaxies selected from the B2 Catalogue of Radio Sources and the Third Cambridge Revised Catalogue (3CR), with no selection constraint on the nuclear properties. We named this sample “the Bologna Complete Sample” (BCS). Methods: We present new VLBI observations at 5 and 1.6 GHz for 33 sources drawn from a sample not biased toward orientation. By combining these data with those in the literature, information on the parsec-scale morphology is available for a total of 76 of 94 radio sources with a range in radio power and kiloparsec-scale morphologies. Results: The fraction of two-sided sources at milliarcsecond resolution is high (30%), compared to the fraction found in VLBI surveys selected at centimeter wavelengths, as expected from the predictions of unified models. The parsec-scale jets are generally found to be straight and to line up with the kiloparsec-scale jets. A few peculiar sources are discussed in detail. Tables 1-4 are only available in electronic form at http://www.aanda.org

  19. TRIGA Mark II nuclear reactor facility. Final report, 1 July 1980--30 June 1995

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, B.C.

    1997-05-01

    This report is a final culmination of activities funded through the Department of Energy`s (DOE) University Reactor Sharing Program, Grant DE-FG02-80ER10273, during the period 1 July 1980 through 30 June 1995. Progress reports have been periodically issued to the DOE, namely the Reactor Facility Annual Reports C00-2082/2219-7 through C00-2082/10723-21, which are contained as an appendix to this report. Due to the extent of time covered by this grant, summary tables are presented. Table 1 lists the fiscal year financial obligations of the grant. As listed in the original grant proposals, the DOE grant financed 70% of project costs, namely the total amount spent of these projects minus materials costs and technical support. Thus the bulk of funds was spent directly on reactor operations. With the exception of a few years, spending was in excess of the grant amount. As shown in Tables 2 and 3, the Reactor Sharing grant funded a immense number of research projects in nuclear engineering, geology, animal science, chemistry, anthropology, veterinary medicine, and many other fields. A list of these users is provided. Out of the average 3000 visitors per year, some groups participated in classes involving the reactor such as Boy Scout Merit Badge classes, teacher`s workshops, and summer internships. A large number of these projects met the requirements for the Reactor Sharing grant, but were funded by the University instead.

  20. Radiation effects in concrete for nuclear power plants, Part II: Perspective from micromechanical modeling

    Energy Technology Data Exchange (ETDEWEB)

    Le Pape, Y., E-mail: lepapeym@ornl.gov; Field, K.G.; Remec, I.

    2015-02-15

    Highlights: • A micromechanical model for irradiated concrete is proposed. • Confrontation with literature data is successful. • Neutron radiation-induced volumetric expansion is a predominant degradation mode. • The nature of the aggregate alters the severity of damage to irradiated concrete. - Abstract: The need to understand and characterize the effects of neutron irradiation on concrete has become urgent because of the possible extension of service life of many nuclear power generating stations. Current knowledge is primarily based on a collection of data obtained in test reactors. These data are inherently difficult to interpret because materials and testing conditions are inconsistent. A micromechanical approach based on the Hashin composite sphere model is presented to derive a first-order separation of the effects of radiation on cement paste and aggregate, and, also, on their interaction. Although the scarcity of available data limits the validation of the model, it appears that, without negating a possible gamma-ray induced effect, the neutron-induced damage and swelling of aggregate plays a predominant role on the overall concrete expansion and the damage of the cement paste. The radiation-induced volumetric expansion (RIVE) effects can also be aided by temperature elevation and shrinkage in the cement paste.

  1. Radiation effects in concrete for nuclear power plants, Part II: Perspective from micromechanical modeling

    International Nuclear Information System (INIS)

    Le Pape, Y.; Field, K.G.; Remec, I.

    2015-01-01

    Highlights: • A micromechanical model for irradiated concrete is proposed. • Confrontation with literature data is successful. • Neutron radiation-induced volumetric expansion is a predominant degradation mode. • The nature of the aggregate alters the severity of damage to irradiated concrete. - Abstract: The need to understand and characterize the effects of neutron irradiation on concrete has become urgent because of the possible extension of service life of many nuclear power generating stations. Current knowledge is primarily based on a collection of data obtained in test reactors. These data are inherently difficult to interpret because materials and testing conditions are inconsistent. A micromechanical approach based on the Hashin composite sphere model is presented to derive a first-order separation of the effects of radiation on cement paste and aggregate, and, also, on their interaction. Although the scarcity of available data limits the validation of the model, it appears that, without negating a possible gamma-ray induced effect, the neutron-induced damage and swelling of aggregate plays a predominant role on the overall concrete expansion and the damage of the cement paste. The radiation-induced volumetric expansion (RIVE) effects can also be aided by temperature elevation and shrinkage in the cement paste

  2. Study on the establishment of retrospective dosimetry system for nuclear radiation accident(II)

    Energy Technology Data Exchange (ETDEWEB)

    Jun, Jae Shik; Chai, Ha Seok; Lee, Jong Ok [Chungnam National Univ., Taejon (Korea, Republic of)

    1999-03-15

    This study was driven forward centering around physical techniques in retrospective dosimetry system for encountering nuclear radiation accident. The results obtained through this study are summarized as follow : the minimal facilities based on physical techniques should be assured at KINS for appropriate operation and establishment of retrospective accident dosimetry system, the necessary apparatus and man power for retrospective dose assessment by physical techniques might be operated flexibly, however, CL and TL/OSL readers should be equipped with the highest priority, a series of comparative examination of several physical techniques for retrospective dose assessment revealed that most of the irradiated materials around accident sites are usable for the dose assessment, if a priori study on the dosimetrical characteristics of those materials is preceded in accordance with the species of the collectable samples, the results of the study on the CL-dose response and radiation energy dependence of sugar and sorbitol, showed the nonlinearity in CL-dose relationship at the range of low dose(less than 5 Gy), and it led us to perform a study on the correction of the nonlinearity, and in the later study, CL output showed heavy dependence on radiation energy in the energy below around 100 keV and accordingly, a study on the correction for the energy dependence was also carried out, ve were able to obtain good results as a first attempt to carry out such corrections.

  3. Development of advanced nuclear core analysis system applicable to various reactor types (II)

    International Nuclear Information System (INIS)

    Kaneko, Kunio

    2003-03-01

    A 900 group cross section library based on the specification determined last year was produced for 27 nuclei of the fast reactor benchmark problem evaluated in nuclear data file JENDL-3.2. In addition, the new SLAROM code, which has been developed as an advanced detail analysis system, was revised so as to make cell calculations effectively with the above 900 group library. Furthermore, new functions were added to the SLAROM so that the SLAROM evaluates assembly parameters using effective cross sections derived by the SLAROM and produces any condensed effective cross section set for core performance analysis. With the 900 group cross section library and the revised SALROM, three cell calculations for fast and medium neutron speed reactors having different neutron spectrum were performed, and the results were compared with those calculated by the continuos energy Monte Carlo code MVP. By the comparisons, it is concluded that the newly revised SLAROM and a 900 group cross section library give accuracy comparable to MVP for predicting core performances. (author)

  4. Nuclear Activities in Argentina. A short Review. Part 2

    International Nuclear Information System (INIS)

    Coll, Jorge A.; Radicella, Renato

    2002-01-01

    The second part of this historical review covers the 'industrial' period of nuclear energy in Argentina. The National Atomic Energy Commission (CNEA), after a feasibility study carried out by argentine experts, in 1968 signed a contract to build a nuclear power plant. This PHWR plant, Atucha 1, of 310 MWe was inaugurated in 1973 and it is still operating. The same year the CNEA signed a new contract to build a CANDU type plant of 600 MWe, Embalse, that was finally inaugurated in 1983. The construction of a third plant, Atucha 2 of 745 MWe also a PHWR, was started in 1980 but was arrested in 1994, when more than 80% was completed, and it is still waiting a political decision to reach completion. Within the development of the nuclear power program, a fuel element production plant for the Argentine power reactors was built by the CNEA and a heavy water production plant of 250 tons/year was inaugurated in 1993 in the southern province of Neuquen. A pilot spent fuel reprocessing plant was designed but its construction was not completed. At the same time, a pilot gaseous diffusion plant was constructed in order to produce enriched uranium for research reactors. The activities in the field of radioisotope and radiation applications were also intensive, mainly in nuclear medicine and food preservation. A facility to fabricate sealed sources was built to process the Co 60 produced by the Embalse power plant. Argentina was active in the export of nuclear facilities: CNEA built a complete nuclear research center in Peru, and the Argentine company INVAP built research reactors in Algeria and Egypt. The same company is now building a research reactor in Australia. (author)

  5. Spent fuel management of NPPs in Argentina

    International Nuclear Information System (INIS)

    Alvarez, D.E.; Lee Gonzalez, H.M.

    2010-01-01

    There are two Nuclear Power Plants in operation in Argentina: 'Atucha I' (unique PHWR design) in operation since 1974, and 'Embalse' (typical Candu reactor) which started operation in 1984. Both NPPs are operated by 'Nucleoelectrica Argentina S.A' which is responsible for the management and interim storage of spent fuel till the end of the operative life of the plants. A third NPP, 'Atucha II' is under construction, with a similar design of Atucha I. The legislative framework establishes that after final shutdown of a NPP the spent fuel will be transferred to the 'National Atomic Energy Commission', which is also responsible for the decommissioning of the Plants. In Atucha I, the spent fuel is stored underwater, until another option is implemented meanwhile in Embalse the spent fuel is stored during six years in pools and then it is moved to a dry storage. A decision about the fuel cycle back-end strategy will be taken before year 2030. (authors)

  6. SPLOSH II: A dynamics programme for nuclear - thermal - hydrodynamic behaviour of water-cooled reactors

    International Nuclear Information System (INIS)

    Moxon, D.

    1966-01-01

    A dynamics code is described that solves the two-group neutron diffusion equations simultaneously with the thermal and the hydraulic equations for an average channel of a water-cooled reactor. Other reactor channels can be represented as 'slaves', which have no feedback to the average channel. The fission power at any axial station in a slave channel is related to that in the average by prescribed time-dependent factors, and the hydraulic flow is determined from pressure-drop requirements dictated by the performance of the average channel. A finite difference model of the fuel element and can represents the behaviour of the fuel temperatures and surface heat flux. The representation of the hydraulic circuit has been made sufficiently general that the code is applicable to B.W.R., P.W.R. and pressure tube reactor designs. The code can be used to study transients resulting from imposed time variations in coolant flow, inlet enthalpy, system pressure, electrical torque supplied to the circulating pumps, (or alternatively, the angular velocity of the pump rotors,) moderator height, frictional resistances simulating blockages and control rod and fuel element insertions. The harmonic response can be obtained by injecting sinusoidal time variations until the starting transient has been damped out. Output includes axial distributions of the neutron fluxes, heat flux, coolant density and temperature, burn-but margin, and the fuel and can temperatures in both the average and the slave channels. The code was originally written in FORTRAN II for use on the IBM 7090. Computing times vary greatly with the problem and the desired accuracy but experience has shown that a computing time which is slower than real time by a factor thirty is adequate for a wide range of cases. The code has recently been converted to S2 and EGTRAN for use on the IBM 7030 and the English Electric Leo Marconi KDF 9 computers. (author)

  7. DJ-1 Modulates Nuclear Erythroid 2-Related Factor-2-Mediated Protection in Human Primary Alveolar Type II Cells in Smokers.

    Science.gov (United States)

    Bahmed, Karim; Messier, Elise M; Zhou, Wenbo; Tuder, Rubin M; Freed, Curt R; Chu, Hong Wei; Kelsen, Steven G; Bowler, Russell P; Mason, Robert J; Kosmider, Beata

    2016-09-01

    Cigarette smoke (CS) is a main source of oxidative stress and a key risk factor for emphysema, which consists of alveolar wall destruction. Alveolar type (AT) II cells are in the gas exchange regions of the lung. We isolated primary ATII cells from deidentified organ donors whose lungs were not suitable for transplantation. We analyzed the cell injury obtained from nonsmokers, moderate smokers, and heavy smokers. DJ-1 protects cells from oxidative stress and induces nuclear erythroid 2-related factor-2 (Nrf2) expression, which activates the antioxidant defense system. In ATII cells isolated from moderate smokers, we found DJ-1 expression by RT-PCR, and Nrf2 and heme oxygenase (HO)-1 translocation by Western blotting and immunocytofluorescence. In ATII cells isolated from heavy smokers, we detected Nrf2 and HO-1 cytoplasmic localization. Moreover, we found high oxidative stress, as detected by 4-hydroxynonenal (4-HNE) (immunoblotting), inflammation by IL-8 and IL-6 levels by ELISA, and apoptosis by terminal deoxynucleotidyl transferase dUTP nick end labeling (TUNEL) assay in ATII cells obtained from heavy smokers. Furthermore, we detected early DJ-1 and late Nrf2 expression after ATII cell treatment with CS extract. We also overexpressed DJ-1 by adenovirus construct and found that this restored Nrf2 and HO-1 expression and induced nuclear translocation in heavy smokers. Moreover, DJ-1 overexpression also decreased ATII cell apoptosis caused by CS extract in vitro. Our results indicate that DJ-1 activates the Nrf2-mediated antioxidant defense system. Furthermore, DJ-1 overexpression can restore the impaired Nrf2 pathway, leading to ATII cell protection in heavy smokers. This suggests a potential therapeutic strategy for targeting DJ-1 in CS-related lung diseases.

  8. Nuclear fuel supply view in Argentina

    International Nuclear Information System (INIS)

    Cirimello, R.O.

    1997-01-01

    The Argentine Atomic Energy Commission promoted and participated in a unique achievement in the R and D system in Argentina: the integration of science technology and production based on a central core of knowledge for the control and management of the nuclear fuel cycle technology. CONUAR SA, as a fuel manufacturer, FAE SA, the manufacturer of Zircaloy tubes, CNEA and now DIOXITEC SA producer of Uranium Dioxide, have been supply, in the last ten years, the amount of products required for about 1300 Tn of equivalent U content in fuels. The most promising changes for the fuel cycle economy is the Slight Enriched Uranium project which begun in Atucha I reactor. In 1997 seventy five fuel assemblies, equivalent to 900 Candu fuel bundles, will complete its irradiation. (author)

  9. Nuclear

    International Nuclear Information System (INIS)

    2014-01-01

    This document proposes a presentation and discussion of the main notions, issues, principles, or characteristics related to nuclear energy: radioactivity (presence in the environment, explanation, measurement, periods and activities, low doses, applications), fuel cycle (front end, mining and ore concentration, refining and conversion, fuel fabrication, in the reactor, back end with reprocessing and recycling, transport), the future of the thorium-based fuel cycle (motivations, benefits and drawbacks), nuclear reactors (principles of fission reactors, reactor types, PWR reactors, BWR, heavy-water reactor, high temperature reactor of HTR, future reactors), nuclear wastes (classification, packaging and storage, legal aspects, vitrification, choice of a deep storage option, quantities and costs, foreign practices), radioactive releases of nuclear installations (main released radio-elements, radioactive releases by nuclear reactors and by La Hague plant, gaseous and liquid effluents, impact of releases, regulation), the OSPAR Convention, management and safety of nuclear activities (from control to quality insurance, to quality management and to sustainable development), national safety bodies (mission, means, organisation and activities of ASN, IRSN, HCTISN), international bodies, nuclear and medicine (applications of radioactivity, medical imagery, radiotherapy, doses in nuclear medicine, implementation, the accident in Epinal), nuclear and R and D (past R and D programmes and expenses, main actors in France and present funding, main R and D axis, international cooperation)

  10. Temperature variation on the Mediterranean Sea by the exploitation of the Vandellos II Nuclear Power Plant; Variacion de temperatura en el mar mediterraneo por la explotacion de la C.N. de Vandellos II

    Energy Technology Data Exchange (ETDEWEB)

    Villarreal Romero, M.; Ribes Hernandez, G.; Esparza Martin, J. L.

    2010-07-01

    The aim of this study is to verify the compliance with the Resolution of 7th February MAH/285/2007 Departament de Medi Ambient I Habitatge de la Generalitat de Catalunya establishing discharges limits to the Mediterranean Sea and, in particular, the section that references the thermal rise. The study area include about 1.5 km coastline, which is located in the vicinity of the Vandellos II Nuclear Power Plant.

  11. Nuclear Waste Management under Approaching Disaster: A Comparison of Decommissioning Strategies for the German Repository Asse II.

    Science.gov (United States)

    Ilg, Patrick; Gabbert, Silke; Weikard, Hans-Peter

    2017-07-01

    This article compares different strategies for handling low- and medium-level nuclear waste buried in a retired potassium mine in Germany (Asse II) that faces significant risk of uncontrollable brine intrusion and, hence, long-term groundwater contamination. We survey the policy process that has resulted in the identification of three possible so-called decommissioning options: complete backfilling, relocation of the waste to deeper levels in the mine, and retrieval. The selection of a decommissioning strategy must compare expected investment costs with expected social damage costs (economic, environmental, and health damage costs) caused by flooding and subsequent groundwater contamination. We apply a cost minimization approach that accounts for the uncertainty regarding the stability of the rock formation and the risk of an uncontrollable brine intrusion. Since economic and health impacts stretch out into the far future, we examine the impact of different discounting methods and rates. Due to parameter uncertainty, we conduct a sensitivity analysis concerning key assumptions. We find that retrieval, the currently preferred option by policymakers, has the lowest expected social damage costs for low discount rates. However, this advantage is overcompensated by higher expected investment costs. Considering all costs, backfilling is the best option for all discounting scenarios considered. © 2016 Society for Risk Analysis.

  12. Failure analysis of leakage on titanium tubes within heat exchangers in a nuclear power plant. Part II: Mechanical degradation

    Energy Technology Data Exchange (ETDEWEB)

    Gong, Y.; Yang, Z.G. [Department of Materials Science, Fudan University, Shanghai (China); Yuan, J.Z. [Third Qinshan Nuclear Power Co. Ltd., Haiyan, Zhejiang Province (China)

    2012-01-15

    Serious failure incidents like clogging, quick thinning, and leakage frequently occurred on lots of titanium tubes of heat exchangers in a nuclear power plant in China. In the Part I of the whole failure analysis study with totally two parts, factors mainly involving three kinds of electrochemical corrosions were investigated, including galvanic corrosion, crevice corrosion, and hydrogen-assisted corrosion. In the current Part II, through microscopically analyzing the ruptures on the leaked tubes by scanning electron microscopy (SEM) and energy dispersive spectrometry (EDS), another four causes dominantly lying in the aspect of mechanical degradation were determined - clogging, erosion, mechanical damaging, and fretting. Among them, the erosion effect was the primary one, thus the stresses it exerted on the tube wall were also supplementarily evaluated by finite element method (FEM). Based on the analysis results, the different degradation extents and morphologies by erosion on the tubes when they were clogged by different substances such as seashell, rubber debris, and sediments were compared, and relevant mechanisms were discussed. Finally, countermeasures were put forward as well. (Copyright copyright 2012 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

  13. Report of the Secretary of Defense Task Force on DoD Nuclear Weapons Management. Phase II: Review of the DoD Nuclear Mission

    National Research Council Canada - National Science Library

    Schlesinger, James R; Carns, Michael P; Crouch, II, J. D; Gansler, Jacques S; Giambastiani, Jr., Edmund P; Hamre, John J; Miller, Franklin C; Williams, Christopher A; Blackwell, Jr, James A

    2008-01-01

    Incidents related to the Air Force's mishandling of nuclear weapons and components led to the creation of the Task Force in June 2008 to provide advice on nuclear matters for the Secretary of Defense...

  14. Proceeding of the Scientific Meeting and Presentation on Basic Research in Nuclear Science and Technology part II : Nuclear Chemistry, Process Technology, Radioactive Waste Management and Environment

    International Nuclear Information System (INIS)

    Sukarsono, R.; Ganang Suradjijo

    2002-01-01

    Scientific Meeting and Presentation on Basic Research in Nuclear Science and Technology is a routine activity held by Centre for Research and Development of Advanced Technology, National Nuclear Energy Agency, for monitoring the research activity which achieved in National Nuclear Energy Agency. This proceedings contains a proposal about basic research in nuclear technology which has environment. This proceedings is the second part of the two parts which published in series. There are 57 articles which have separated index. (PPIN)

  15. Inhibition of human T cell leukemia virus type 2 replication by the suppressive action of class II transactivator and nuclear factor Y.

    Science.gov (United States)

    Tosi, Giovanna; Pilotti, Elisabetta; Mortara, Lorenzo; De Lerma Barbaro, Andrea; Casoli, Claudio; Accolla, Roberto S

    2006-08-22

    The master regulator of MHC-II gene transcription, class II transactivator (CIITA), acts as a potent inhibitor of human T cell leukemia virus type 2 (HTLV-2) replication by blocking the activity of the viral Tax-2 transactivator. Here, we show that this inhibitory effect takes place at the nuclear level and maps to the N-terminal 1-321 region of CIITA, where we identified a minimal domain, from positions 64-144, that is strictly required to suppress Tax-2 function. Furthermore, we show that Tax-2 specifically cooperates with cAMP response element binding protein-binding protein (CBP) and p300, but not with p300/CBP-associated factor, to enhance transcription from the viral promoter. This finding represents a unique difference with respect to Tax-1, which uses all three coactivators to transactivate the human T cell leukemia virus type 1 LTR. Direct sequestering of CBP or p300 is not the primary mechanism by which CIITA causes suppression of Tax-2. Interestingly, we found that the transcription factor nuclear factor Y, which interacts with CIITA to increase transcription of MHC-II genes, exerts a negative regulatory action on the Tax-2-mediated HTLV-2 LTR transactivation. Thus, CIITA may inhibit Tax-2 function, at least in part, through nuclear factor Y. These findings demonstrate the dual defensive role of CIITA against pathogens: it increases the antigen-presenting function for viral determinants and suppresses HTLV-2 replication in infected cells.

  16. Proceedings of the Scientific Meeting and Presentation on Basic Researchin Nuclear Science and Technology part II: Nuclear Chemistry, Process Technology, Radioactive Waste Management and Environment

    International Nuclear Information System (INIS)

    Sukarsono, R.; Karmanto, Eko-Edy; Suradjijo, Ganang

    2000-01-01

    Scientific Meeting and Presentation on Basic Research in Nuclear Scienceand Technology is an annual activity held by Centre for Research and Development of Advanced Technology, National Nuclear Energy Agency, for monitoring research activities achieved by the Agency. The papers presented in the meeting were collected into proceedings. These are the second part of the proceedings that contain 71 articles in the fields of nuclear chemistry, process technology, radioactive waste management, and environment (PPIN).

  17. Proceeding of the Scientific Meeting and Presentation on Basic Research in Nuclear of the Scientific and Technology Part II : Nuclear Chemistry; Process Technology and Radioactive Waste Management; Environment

    International Nuclear Information System (INIS)

    Sudjatmoko; Karmanto, Eko Edy; Endang-Supartini

    1996-04-01

    Scientific Meeting and Presentation on Basic Research in Nuclear Science and Technology is a routine activity was held by Yogyakarta Nuclear Research Centre, National Atomic Energy Agency (BATAN) for monitoring the research activity which achieved in BATAN. The Proceeding contains a proposal about basic which has Nuclear Chemistry, Process Technology, Radioactive Waste Management and Environment. This proceeding is the second part from two part which published in series. There are 61 articles which have separated index

  18. Nuclear power plant simulators for operator licensing and training. Part I. The need for plant-reference simulators. Part II. The use of plant-reference simulators

    International Nuclear Information System (INIS)

    Rankin, W.L.; Bolton, P.A.; Shikiar, R.; Saari, L.M.

    1984-05-01

    Part I of this report presents technical justification for the use of plant-reference simulators in the licensing and training of nuclear power plant operators and examines alternatives to the use of plant-reference simulators. The technical rationale is based on research on the use of simulators in other industries, psychological learning and testing principles, expert opinion and user opinion. Part II discusses the central considerations in using plant-reference simulators for licensing examination of nuclear power plant operators and for incorporating simulators into nuclear power plant training programs. Recommendations are presented for the administration of simulator examinations in operator licensing that reflect the goal of maximizing both reliability and validity in the examination process. A series of organizational tasks that promote the acceptance, use, and effectiveness of simulator training as part of the onsite training program is delineated

  19. p53 nuclear accumulation and multiploidy are adverse prognostic factors in surgically resected stage II colorectal cancers independent of fluorouracil-based adjuvant therapy.

    Science.gov (United States)

    Buglioni, S; D'Agnano, I; Vasselli, S; Perrone Donnorso, R; D'Angelo, C; Brenna, A; Benevolo, M; Cosimelli, M; Zupi, G; Mottolese, M

    2001-09-01

    To identify the prognostically highest risk patients, DNA content and p53 nuclear or cytoplasmic accumulation, evaluated by monoclonal antibody DO7 and polyclonal antibody CM1, were determined in 94 surgically resected stage II (Dukes B2) colorectal cancers, treated or not with adjuvant 5-fluorouracil-based chemotherapy. Sixty-one (65%) of the tumors were aneuploid, 16 (17%) of which had a multiploid DNA content; 50 (53%) displayed DO7 nuclear p53 accumulation, and 44 (47%) showed cytoplasmic CM1 positivity. In multivariate analysis, only multiploidy and p53 nuclear positivity emerged as independent prognostic indicators of a poorer outcome. Positivity for p53 was associated with shorter survival in 5-fluorouracil-treated and untreated patients. Therefore, in patients with Dukes B2 colorectal cancer, a biologic profile based on the combined evaluation of DNA multiploidy and p53 status can provide valuable prognostic information, identifying patients to be enrolled in alternative, more aggressive therapeutic trials.

  20. Proceedings of the Scientific Meeting and Presentation on Basic Research in Nuclear of the Science and Technology part II : Nuclear Chemistry and Process Technology

    International Nuclear Information System (INIS)

    Kamsul Abraha; Yateman Arryanto; Sri Jauhari S; Agus Taftazani; Kris Tri Basuki; Djoko Sardjono, Ign.; Sukarsono, R.; Samin; Syarip; Suryadi, MS; Sardjono, Y.; Tri Mardji Atmono; Dwiretnani Sudjoko; Tjipto Sujitno, BA.

    2007-08-01

    The Scientific Meeting and Presentation on Basic Research in Nuclear Science and Technology is a routine activity held by Centre for Accelerator Technology and Material Process, National Nuclear Energy Agency, for monitoring the research activity which achieved in National Nuclear Energy Agency. The Meeting was held in Yogyakarta on July 10, 2007. The proceedings contains papers presented on the meeting about Nuclear Chemistry and Process Technology and there are 47 papers which have separated index. The proceedings is the second part of the three parts which published in series. (PPIN)

  1. Report of the joint seminar on heavy-ion nuclear physics and nuclear chemistry in the energy region of tandem accelerators (II)

    International Nuclear Information System (INIS)

    1986-04-01

    A meeting of the second joint seminar on Heavy-Ion Nuclear Physics and Nuclear Chemistry in the Energy Region of Tandem Accelerators was held after an interval of two years at the Tokai Research Establishment of the JAERI, for three days from January 9 to 11, 1986. In the seminar, about 70 nuclear physicists and nuclear chemists of JAERI and other Institutes participated, and 38 papers were presented. These include general reviews and topical subjects which have been developed intensively in recent years, as well as the new results obtained by using the JAERI tandem accelerator. This report is a collection of the papers presented to the seminar. (author)

  2. Nuclear materials safeguards. Volume II. 1975--March 1976 (a bibliography with abstracts). Report for 1975--Mar 1976

    International Nuclear Information System (INIS)

    Grooms, D.W.

    1976-03-01

    Citations cover the methods of safeguarding nuclear materials through effective management, accountability, nondestructive assays, instrumentation, and automated continuous inventory systems. Problem areas and recommendations for improving the management of nuclear materials are included. (Contains 88 abstracts) See also NTIS/PS-76/0200, Nuclear Materials Safeguards. Vol. 1. 1964-1974 (A title bibliography)

  3. FINESSE: study of the issues, experiments and facilities for fusion nuclear technology research and development. Interim report. Volume II

    International Nuclear Information System (INIS)

    Abdou, M.

    1984-10-01

    The Nuclear Fusion Issues chapter contains a comprehensive list of engineering issues for fusion reactor nuclear components. The list explicitly defines the uncertainties associated with the engineering option of a fusion reactor and addresses the potential consequences resulting from each issue. The next chapter identifies the fusion nuclear technology testing needs up to the engineering demonstration stage

  4. Release procedure according to paragraph 29 StrlSchv on example of the nuclear research reactor TRIGA Heidelberg II; Durchfuehrung von Freigabeverfahren nach paragraph 29 am Beispiel des TRIGA Heidelberg II

    Energy Technology Data Exchange (ETDEWEB)

    Cremer, J. [Siempelkamp Nukleartechnik GmbH (SNT) (Germany); Sold, A. [Deutsches Krebsforschungszentrum Heidelberg (DKFZ) (Germany)

    2005-07-01

    The aim of this lecture is to show the schedule of a release procedure according to paragraph 29 StrlSchV on the example of the decommissioning of the nuclear research reactor TRIGA Heidelberg II. It is shown on the effort done by the radiation protection representative of this plant. Considering this example, starting with planning, application, survey and execution, the complex context of the release procedure is becomes apparent. Thereby the new applied measuring techniques that require a certain practice and the responsibility of the radiation protection representative in the radiation protection law play a relevant role. In such small facilities as the TRIGA Heidelberg II, the radiation protection staff are employed according to the plant's size and work is focussed on radiation protection research and laboratories. The decommissioning process with its wide range of radiation protection requirements represents new challenges which have to be coordinated with the present duties of the radiation protection representative. The supervision and the responsibility for the release procedure according to paragraph 29 are the largest and the most sensitive part of decommissioning of the nuclear research reactor TRIGA Heidelberg II. (orig.)

  5. Chiral Nuclear Dynamics II

    CERN Document Server

    Rho, Mannque

    2008-01-01

    This is the sequel to the first volume to treat in one effective field theory framework the physics of strongly interacting matter under extreme conditions. This is vital for understanding the high temperature phenomena taking place in relativistic heavy ion collisions and in the early Universe, as well as the high-density matter predicted to be present in compact stars. The underlying thesis is that what governs hadronic properties in a heat bath and/or a dense medium is hidden local symmetry which emerges from chiral dynamics of light quark systems and from the duality between QCD in 4D and

  6. Proceeding of the Scientific Meeting and Presentation on Basic Research of Nuclear Science and Technology: Book II. Nuclear Chemistry, Process Technology, and Radioactive Waste Processing and Environment

    International Nuclear Information System (INIS)

    1996-06-01

    The proceeding contains papers presented on Scientific Meeting and Presentation on on Basic Research of Nuclear Science and Technology, held in Yogyakarta, 25-27 April 1995. This proceeding is second part of two books published for the meeting contains papers on nuclear chemistry, process technology, and radioactive waste management and environment. There are 62 papers indexed individually. (ID)

  7. Vulnerability Assessment of the nuclear power plant Vandellos II before a tornado; Evaluacion de vulnerabilidad de C.N. Vandellos II ante tornado

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, A.; Encabo, J.; Vaz-Romero, A.; Moran, M. A.; Roch, M.; Nicolas, P.; Barrera, N.

    2010-07-01

    The purpose of this work was the study of vulnerability to tornado event Vandellos II NPP. To do this, we have evaluated all structures (buildings), security systems and components to the installation of wind stresses, depression and impact of projectiles, generated by a tornado on the site.

  8. In vivo inhibition of nuclear factor of activated T-cells leads to atherosclerotic plaque regression in IGF-II/LDLR-/-ApoB100/100 mice.

    Science.gov (United States)

    Blanco, Fabiana; Heinonen, Suvi E; Gurzeler, Erika; Berglund, Lisa M; Dutius Andersson, Anna-Maria; Kotova, Olga; Jönsson-Rylander, Ann-Cathrine; Ylä-Herttuala, Seppo; Gomez, Maria F

    2018-03-01

    Despite vast clinical experience linking diabetes and atherosclerosis, the molecular mechanisms leading to accelerated vascular damage are still unclear. Here, we investigated the effects of nuclear factor of activated T-cells inhibition on plaque burden in a novel mouse model of type 2 diabetes that better replicates human disease. IGF-II/LDLR -/- ApoB 100/100 mice were generated by crossbreeding low-density lipoprotein receptor-deficient mice that synthesize only apolipoprotein B100 (LDLR -/- ApoB 100/100 ) with transgenic mice overexpressing insulin-like growth factor-II in pancreatic β cells. Mice have mild hyperglycaemia and hyperinsulinaemia and develop complex atherosclerotic lesions. In vivo treatment with the nuclear factor of activated T-cells blocker A-285222 for 4 weeks reduced atherosclerotic plaque area and degree of stenosis in the brachiocephalic artery of IGF-II/LDLR -/- ApoB 100/100 mice, as assessed non-invasively using ultrasound biomicroscopy prior and after treatment, and histologically after termination. Treatment had no impact on plaque composition (i.e. muscle, collagen, macrophages). The reduced plaque area could not be explained by effects of A-285222 on plasma glucose, insulin or lipids. Inhibition of nuclear factor of activated T-cells was associated with increased expression of atheroprotective NOX4 and of the anti-oxidant enzyme catalase in aortic vascular smooth muscle cells. Targeting the nuclear factor of activated T-cells signalling pathway may be an attractive approach for the treatment of diabetic macrovascular complications.

  9. Experimental study of nuclear models. I. Decay schemes and nuclear reactions. II. Muonic x-ray studies. Progress report, October 1, 1974--September 30, 1975

    International Nuclear Information System (INIS)

    Sheline, R.K.

    1975-01-01

    Progress on the research on our AT-(40-1)-2434 Contract is summarized for the twelve month contract year beginning October 1, 1974, and ending September 30, 1975. The main emphasis of our research continues to be an experimental study of nuclear models. Some change of emphasis is occurring. In the past, the emphasis has been overwhelmingly nuclear reaction spectroscopy and comparison with theoretical models. This year an increasing percentage of the emphasis (perhaps 25 percent) is on the study of nuclear structure from the view point of muonic x-ray spectroscopy. A list of publications is included. (U.S.)

  10. Proceedings of INC 02. International Nuclear Conference 2002: Global Trends and Perspectives, Seminar II: Medicine and Health

    International Nuclear Information System (INIS)

    2002-01-01

    The papers discuss the uses of radiations and radioisotopes in Medicine and Health, it included the area of nuclear medicine, biomedical radiography, radiopharmaceuticals; isotope production; cancer treatment, etc

  11. The legal regime governing the peaceful uses of nuclear energy. II. International Regulations. Pt.1. Regulations on peaceful uses

    International Nuclear Information System (INIS)

    1979-12-01

    The first volume on atomic energy law published by CNEN reproduced national laws and regulations in that field. This book constitutes part one of the second volume and deals with international nuclear conventions and cooperation as at 30 June 1978. It reproduces the instruments and conventions which set up the international nuclear agencies, recommendations in the field of radiation protection and nuclear safety, the nuclear third party liability conventions, the international instruments concerning technical and scientific cooperation and finally, the bilateral cooperation agreements between Italy and other nations and its agreements with international organizations (NEA) [fr

  12. Fifty years contribution to research and technological development of Argentina. Part 7

    International Nuclear Information System (INIS)

    Aguirre, Fernando; Boselli, Alfredo; Colangelo, Luis J.; Coll, Jorge A.; Espejo, Hector; Mattei, Clara E.; Ornstein, Roberto M.; Palacios, Tulio A.; Radicella, Renato; Rodrigo, Felix

    2004-01-01

    The paper is the seventh part of a short history of the National Atomic Energy Commission (CNEA). The future of the nuclear activities in the country is outlined, mainly of those related to the energy generation. The completion of the Atucha II nuclear power plant now under construction is supported. (author)

  13. Radiological impact of the management of radioactive waste arising from the Argentine Nuclear Programme

    International Nuclear Information System (INIS)

    Migliori de Beninson, A.; Cancio, D.

    1984-01-01

    The Argentine nuclear programme, as it stands at present, provides for the construction of four nuclear power plants in addition to those of Atucha I and Embalse and for the establishment of such fuel cycle facilities as are required to supply all of these plants. This paper evaluates the radiological impact (collective dose commitment) expected from the management of the radioactive wastes arising in the facilities mentioned above throughout the useful life of the reactors. The maximum individual doses to be expected as a result of the planned high-level-waste repository are also estimated. The evaluations presented are partly specific to the sites under consideration, but they also include estimates of the total collective dose commitments resulting from the management of radioactive waste under the Argentine nuclear programme. (author)

  14. Nuclear

    International Nuclear Information System (INIS)

    Anon.

    2000-01-01

    The first text deals with a new circular concerning the collect of the medicine radioactive wastes, containing radium. This campaign wants to incite people to let go their radioactive wastes (needles, tubes) in order to suppress any danger. The second text presents a decree of the 31 december 1999, relative to the limitations of noise and external risks resulting from the nuclear facilities exploitation: noise, atmospheric pollution, water pollution, wastes management and fire prevention. (A.L.B.)

  15. Conceptual evaluation of type B(U) casks for the nuclear power plants of Argentina

    International Nuclear Information System (INIS)

    Florido, P.C.; Isnardi, E.R.

    1993-01-01

    In Argentina two different nuclear power plants are in operation, Atucha I (PHWR-Siemens) and Embalse (PHWR-CANDU). Thus two very different fuel elements could be potentially transported. In order to optimize the research and development needed for the design and construction of the cask, the cost-benefit and flexibility of the engineering solutions are studied, for the two fuel elements. Different casks, for both types of existing fuel elements (Atucha I and Embalse), for different burnup-levels (regarding the advanced fuel cycle available), decay times, distances, and transported weight were studied. Three materials for shielding were used: uranium lead and steel. Only transport by road was considered, due to the reduced availability of the train. In this stage (conceptual design) small, easy and fast computer programs should be used. The principal issues that have to be fixed are shielding properties, thermal effects and mechanical behavior. As a result of the evaluation, different options for the casks were founded, as well as the importance of different parameters and the effect of two different designs of fuel elements. (author)

  16. Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids - Part 2: Iron(II) reduction/cerium(IV) oxidation titrimetric method

    International Nuclear Information System (INIS)

    2004-01-01

    This first edition of ISO 7097-1 together with ISO 7097-2:2004 cancels and replaces ISO 7097:1983, which has been technically revised, and ISO 9989:1996. ISO 7097 consists of the following parts, under the general title Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids: Part 1: Iron(II) reduction/potassium dichromate oxidation titrimetric method; Part 2: Iron(II) reduction/cerium(IV) oxidation titrimetric method. This part 2. of ISO 7097 describes procedures for determination of uranium in solutions, uranium hexafluoride and solids. The procedures described in the two independent parts of this International Standard are similar: this part uses a titration with cerium(IV) and ISO 7097-1 uses a titration with potassium dichromate

  17. Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids - Part 1: Iron(II) reduction/potassium dichromate oxidation titrimetric method

    International Nuclear Information System (INIS)

    2004-01-01

    This first edition of ISO 7097-1 together with ISO 7097-2:2004 cancels and replaces ISO 7097:1983, which has been technically revised, and ISO 9989:1996. ISO 7097 consists of the following parts, under the general title Nuclear fuel technology - Determination of uranium in solutions, uranium hexafluoride and solids: Part 1: Iron(II) reduction/potassium dichromate oxidation titrimetric method; Part 2: Iron(II) reduction/cerium(IV) oxidation titrimetric method. This part 1. of ISO 7097 describes procedures for the determination of uranium in solutions, uranium hexafluoride and solids. The procedures described in the two independent parts of this International Standard are similar: this part uses a titration with potassium dichromate and ISO 7097-2 uses a titration with cerium(IV)

  18. Communication and surrounding of Asco and Vandellos II nuclear power plants; La comunicacion y el entorno de las centrales de Asco y Vandellos II

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    The Asco and Vandellos-II power plants have always been integrated into the regions where they are located, and they take an active part in the development of surrounding towns through quality employment provided by our facilities, social and cultural support and aid to development promoted by regional councils. Communication to media is a corporate priority defined in our strategic plant, to ensure openness, rigour and punctuality. We also attend to the visitors who want to learn more about our facilities in the visitor center, and we have agreements with agrarian institutions in the area so that students can de practical training in the farms we own for agricultural production. (Author)

  19. Report of the generation of the nuclear bank 'L1PG9121' of the SVEA-96 'collapsed' assemble for the FCS-II program with the FMS codes

    International Nuclear Information System (INIS)

    Alonso V, G.

    1992-01-01

    In this work it is described in a general way the form in that was generated the collapsed bank of the SVEA-96 fuel for Laguna Verde. The formation of the bank it was carried out with the ECLIPSE 86-2D, RECORD 89-1A and POLGEN 88-1B codes of the FMS package installed in the VAX system of the office of the National Commission of Nuclear Safety and Safeguards in Mexico D.F. The formed bank is denominated 'LlPG9121'. All this one carries out following the procedure '6F3/I/CN029/90/P1'. To generate the bank, both RECORD 'cells' that compose the assemble its were 'collapsed' in an alone one, representing this, the complete assemble in what refers to the distribution of fuel bar and enrichment. The collapsed of the assemble was made averaging the content of UO 2 and Gd 2 O 3 in each fuel bar of the one assemble. By this way the x-y array of fuel bars is conserved but a representative fuel cell of all the one assemble is obtained, being this the studied RECORD cell. In accordance with the requirements of nuclear information of FCS-II, the nuclear information generated with RECORD only was of the defined type as series 1 in the generation procedure of nuclear banks '6F3/I/CN029/90/P1'. This only means that only was generated nuclear information as function of the fuel burnt and of the vacuum in the fuel cell. Although the nuclear bank (L1PG9121) it was generated in these circumstances, it was also generates information of the defined type as series 2 with the present control bar for possible reactor analysis under these conditions. (Author)

  20. CAE meteorological database for the PC CREAM program. Atmospheric dilution factor in different points of the CAE (Centro Atomico Ezeiza) and of the argentine nuclear power plants

    International Nuclear Information System (INIS)

    Amado, Valeria A.

    2007-01-01

    In the first part of this work, the EZEIZA.MET file, with the meteorological database of the surroundings of the Ezeiza Atomic Center, is prepared and incorporated into the library of the PC CREAM program. This program was developed by the National Radiological Protection Board and the European Union. Information provided by the National Meteorological Service was used, corresponding to the Ezeiza Meteorological Station during the period 1996-2005. In the second part, a methodology to estimate the atmospheric dilution factor at a point using the PLUME module of the PC CREAM, is presented. The developed methodology was used to estimate the dilution factor at points close to the Ezeiza Atomic Center and nuclear power plants Atucha I and Embalse. The developed methodology was used to estimate the dilution factor at points close to the Ezeiza Atomic Center and nuclear power plants Atucha I and Embalse. In the first case the file with the generated meteorological database is used, whereas for the nuclear power plants the already existing ATUCHALO.MET and EMBALSE.MET files are used. The dilution factors obtained are compared with those obtained in previous work. The proposed methodology is a useful tool to estimate the dilution factors in a simple and systematic way, and simultaneously allows the update of the meteorological information used in the estimations. (author) [es

  1. Development and Validation of Methodology to Model Flow in Ventilation Systems Commonly Found in Nuclear Facilities - Phase II

    Energy Technology Data Exchange (ETDEWEB)

    Strons, Philip [Argonne National Lab. (ANL), Argonne, IL (United States); Bailey, James L. [Argonne National Lab. (ANL), Argonne, IL (United States); Davis, John [Argonne National Lab. (ANL), Argonne, IL (United States); Grudzinski, James [Argonne National Lab. (ANL), Argonne, IL (United States); Hlotke, John [Argonne National Lab. (ANL), Argonne, IL (United States)

    2016-03-01

    In this report we present the results of the Phase II analysis and testing of the flow patterns encountered in the Alpha Gamma Hot Cell Facility (AGHCF), as well as the results from an opportunity to expand upon field test work from Phase I by the use of a Class IIIb laser. The addition to the Phase I work is covered before proceeding to the results of the Phase II work, followed by a summary of findings.

  2. Dynamic Isotope Power System (DIPS) Applications Study. Volume II. Nuclear Integrated Multimission Spacecraft (NIMS) design definition. Final report

    International Nuclear Information System (INIS)

    1979-11-01

    The design requirements for the Nuclear Integrated Multimission Spacecraft. (NIMS) are discussed in detail. The requirements are a function of mission specifications, payload, control system requirements, electric system specifications, and cost limitations

  3. Computational Analysis of Nuclear Safety Parameters of 3 MW TRIGA Mark-II Research Reactor Based on Evaluated Nuclear Data Libraries JENDL-3.3 and ENDF/B-VII.0

    International Nuclear Information System (INIS)

    Khan, Jahirul Haque

    2013-01-01

    The objective of this study is to explain the main nuclear safety parameters of 3 MW TRIGA Mark-II Research Reactor at AERE, Savar, Dhaka, Bangladesh from the viewpoint of reactor safety and also reactor operator. The most important nuclear reactor physics safety parameters are power distribution, power peaking factors, shutdown margin, control rod worth, excess reactivity and fuel temperature reactivity coefficient. These parameters are calculated using the chain of the computer codes the SRAC-PIJ for cell calculation based on neutron transport theory and the SRAC-CITATION for core calculation based on neutron diffusion equation. To achieve this objective the TRIGA model is developed by the 3-D diffusion code SRAC-CITATION based on the group constants that come from the collision probability transport code SRAC-PIJ. In this study the evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 are used. The calculated most important reactor physics parameters are compared to the safety analysis report (SAR) values as well as earlier published MCNP results (numerically benchmark). It was found that the calculated results show a good agreement between the said libraries. Besides, in most cases the calculated results reveal a reasonable agreement with the SAR values (by General Atomic) as well as the MCNP results. In addition, this analysis can be used as the inputs for thermal-hydraulic calculations of the TRIGA fresh core in the steady state and pulse mode operation. Because of power peaking factors, power distributions and temperature reactivity coefficients are the most important reactor safety parameters for normal operation and transient safety analysis in research as well as in power reactors. They form the basis for technical specifications and limitations for reactor operation such as loading pattern limitations for pulse operation (in TRIGA). Therefore, this analysis will be very important to develop the nuclear safety parameters data of 3 MW TRIGA Mark-II

  4. Validation of finite element code DELFIN by means of the zero power experiences at the nuclear power plant of Atucha I

    International Nuclear Information System (INIS)

    Grant, C.R.

    1996-01-01

    Code DELFIN, developed in CNEA, treats the spatial discretization using heterogeneous finite elements, allowing a correct treatment of the continuity of fluxes and currents among elements and a more realistic representation of the hexagonal lattice of the reactor. It can be used for fuel management calculation, Xenon oscillation and spatial kinetics. Using the HUEMUL code for cell calculation (which uses a generalized two dimensional collision probability theory and has the WIMS library incorporated in a data base), the zero power experiences performed in 1974 were calculated. (author). 8 refs., 9 figs., 3 tabs

  5. Nuclear data sensitivity and uncertainty for the Canadian supercritical water-cooled reactor II: Full core analysis

    International Nuclear Information System (INIS)

    Langton, S.E.; Buijs, A.; Pencer, J.

    2015-01-01

    Highlights: • H-2, Pu-239, and Th-232 make large contributions to SCWR modelling sensitivity. • H-2, Pu-239, and Th-232 make large contributions to SCWR modelling uncertainty. • Isotopes of Zr make large contributions to SCWR modelling uncertainty. - Abstract: Uncertainties in nuclear data are a fundamental source of uncertainty in reactor physics calculations. To determine their contribution to uncertainties in calculated reactor physics parameters, a nuclear data sensitivity and uncertainty study is performed on the Canadian supercritical water reactor (SCWR) concept. The nuclear data uncertainty contributions to the neutron multiplication factor k eff are 6.31 mk for the SCWR at the beginning of cycle (BOC) and 6.99 mk at the end of cycle (EOC). Both of these uncertainties have a statistical uncertainty of 0.02 mk. The nuclear data uncertainty contributions to Coolant Void Reactivity (CVR) are 1.0 mk and 0.9 mk for BOC and EOC, respectively, both with statistical uncertainties of 0.1 mk. The nuclear data uncertainty contributions to other reactivity parameters range from as low as 3% of to as high as ten times the values of the reactivity coefficients. The largest contributors to the uncertainties in the reactor physics parameters are Pu-239, Th-232, H-2, and isotopes of zirconium

  6. New Nuclear Materials Including Non Metallic Fuel Elements. Vol. II. Proceedings of the Conference on New Nuclear Materials Technology, Including Non Metallic Fuel Elements

    International Nuclear Information System (INIS)

    1963-01-01

    One of the major aims of the International Atomic Energy Agency in furthering the peaceful uses of atomic energy is to encourage the development of economical nuclear power. Certainly, one of the more obvious methods of producing economical nuclear power is the development of economical fuels that can be used at high temperatures for long periods of time, and which have sufficient strength and integrity to operate under these conditions without permitting the release of fission products. In addition it is desirable that after irradiation these new fuels be economically reprocessed to reduce further the cost of the fuel cycle. As nuclear power becomes more and more competitive with conventional power the interest in new and more efficient higher-temperature fuels naturally increases rapidly. For these reasons, the Agency organized a Conference on New Nuclear Materials Technology, Including Non-Metallic Fuel Elements, which was held from 1 to 5 July 1963 at the International Hotel, Prague, with the assistance and co-operation of the Government of the Czechoslovak Socialist Republic. A total of 151 scientists attended, from 23 countries and 4 international organizations. The participants heard and discussed more than 60 scientific papers. The Agency wishes to thank the scientists who attended this Conference for their papers and for many spirited discussions that truly mark a successful meeting. The Agency wishes also to record its gratitude for the assistance and generous hospitality accorded the Conference, the participants and the Agency's staff by the Government of the Czechoslovak Socialist Republic and by the people of Prague. The scientific information contained in these Proceedings should help to quicken the pace of progress in the fabrication of new and m ore economical fuels, and it is hoped that these proceedings will be found useful to all workers in this and related fields

  7. Emergency planning of the city of Munich with reference to nuclear facilities, especially the nuclear power stations Isar I and II, resp. the reactor in Garching

    International Nuclear Information System (INIS)

    1990-01-01

    During the hearing of Munich's city council of 13.7.1990 thirteen experts were heard on the following subjects: Hazard potential of Isar reactors and FRM reactor and appropriate radioactive waste transports; responsibilities in emergency planning. Some of the experts cannot visualize a major accident and propose not to cater for it. Shelters and evacuation are not planned for Munich, both solutions not being realizable for all inhabitants. Nuclear phaseout is seen by some as a measure of prevention. (HSCH) [de

  8. IAEA’s Perspectives on Global Nuclear Power – Opportunities and Challenges

    International Nuclear Information System (INIS)

    Park, J.K.

    2014-01-01

    Status of global nuclear power: 437 reactors in operation (374.5 GWe); 2 reactors in long-term shutdown; 149 reactors in permanent shutdown; 70 reactors under construction. [As of Sep. 2014] Latest connections to the grid: - Ningde-2, 1000 MW(e), PWR, China; - Atucha-2, 692 MW(e), PHWR, Argentina; - Fuqing-1, 1000 MW(e), PWR, China). [Website: http://www.iaea.org/pris/]. IAEA projections of nuclear power: • Sep. 2014: 374.5 GWe; • 2030 - low 400.6 GWe: 7.0% increase; - high 699.2 GWe: 86.7% increase; • 2050 - low 412.9 GWe: 10.3% increase; - high 1091.7 GWe: 191.5% increase

  9. Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

    Science.gov (United States)

    Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Unruh, Troy C.; McGregor, Douglas S.; Roberts, Jeremy A.

    2017-08-01

    Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Finally, manual reactor power oscillations were observed in real-time.

  10. Scoping of material damage with FISPACT-II and different nuclear data libraries: transmutation, activation, and PKAs

    International Nuclear Information System (INIS)

    Gilbert, M.R.; Sublet, J.-Ch.

    2016-01-01

    The uncertainty associated with nuclear data, and the simulated predictions of transmutation, activation, and primary damage events derived from them, is not only that derived based on the quantified errors in a particular nuclear library. Uncertainty also manifests in comparisons between different libraries – if they do not produce the same results, then, since it often impossible to know a priori which library is best, predicted results must be considered to have an uncertainty (at least) as much as the variation between libraries. Of course, this situation is further complicated by the fact that it is not always possible, or practical, to produce results with multi-libraries. There is thus a need, within the nuclear data community, to assess different libraries, and make recommendations about the best choice of library for particular applications, in this case material science

  11. Joint Thesaurus. Part I (A-L) + Part II (M-Z)[International Nuclear Information System. Energy Technology Data Exchange

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-04-01

    This is the 1st revision of the INIS/ETDE Joint Thesaurus. It contains 20 953 valid descriptors and 8 600 forbidden terms. It was last updated in December 2003. The Joint Thesaurus contains the controlled terminology for indexing all information within the subject scope of both INIS (International Nuclear Information System) and ETDE (Energy Technology Data Exchange) information systems. The terminology is intended for use in subject description for input or retrieval of information in those systems. The thesaurus is a terminological control device used in translating from the natural language of documents, indexers or users into a more constrained system language It is also a controlled and dynamic vocabulary of semantically and generically related terms which covers a specific domain of knowledge. The domain of knowledge covered by this Thesaurus includes physics (in particular, plasma physics, atomic and molecular physics, and especially nuclear and high-energy physics), chemistry, materials, earth sciences, radiation biology, radioisotope effects and kinetics, applied life sciences, radiology and nuclear medicine, isotope and radiation source technology, radiation protection, radiation applications, engineering, instrumentation, fossil fuels, synthetic fuels, renewable energy sources, advanced energy systems, fission and fusion reactor technology, safeguards and inspection, waste management, environmental aspects of the production and consumption of energy from nuclear and non-nuclear sources, energy efficiency and energy conservation, economics and sociology of energy production and use, energy policy, and nuclear law. The terms in the Thesaurus are listed alphabetically, and with each alphabetic entry a word block containing the terms associated with the particular entry is displayed. In the word block, terms that have a hierarchical relationship to the entry are identified by the symbols BT and NT, for Broader Term and Narrower Term. Those with an affinitive

  12. Biologically relevant mono- and di-nuclear manganese II/III/IV complexes of mononegative pentadentate ligands

    DEFF Research Database (Denmark)

    Baffert, Carole; Collomb, Marie-Nöelle; Deronzier, Alain

    2003-01-01

    were characterised by UV-visible spectroscopy, ESI mass spectrometry and cyclic voltammetry. In addition, III-IV and II-III species were electrochemically generated. Thus the new mononegative pentadentate ligand systems display significant flexibility in the range of Mn oxidation states and species...

  13. Nuclear Technology. Course 32: Nondestructive Examination (NDE) II. Module 32-3, Fundamentals of Magnetic Particle Testing.

    Science.gov (United States)

    Groseclose, Richard

    This third in a series of six modules for a course titled Nondestructive Examination (NDE) Techniques II explains the principles of magnets and magnetic fields and how they are applied in magnetic particle testing, describes the theory and methods of magnetizing test specimens, describes the test equipment used, discusses the principles and…

  14. Nuclear Technology. Course 32: Nondestructive Examination (NDE) Techniques II. Module 32-4, Operation of Magnetic Particle Test Equipment.

    Science.gov (United States)

    Groseclose, Richard

    This fourth in a series of six modules for a course titled Nondestructive Examination (NDE) Techniques II describes the specific technique variables and options which are available to the test technician, provides instructions for selecting and operating the appropriate test equipment, describes physical criteria for detectable discontinuities,…

  15. Regulatory activities related with the modification of the frequency of the programmed stoppings of the Argentine nuclear centrals; Actividades regulatorias relacionadas con la modficacion de la frecuencia de las paradas programadas de las centrales nucleares argentinas

    Energy Technology Data Exchange (ETDEWEB)

    Marino, E.; Calvo, J.; Waldman, R.; Navarro, R

    2006-07-01

    The mandatory character documentation of the Argentinean nuclear power stations in Embalse and Atucha I, required the realization of a programmed stoppings every twelve months to execute that settled down in the maintenance and surveillance programs for each installation. Nucleoelectrica Argentina S.A., in it character of Responsible Entity of the operation of these power stations, requested to the Argentinean Nuclear Regulatory Authority, in 2003 and 2005 respectively, the authorization to change the period of the repetitive tests and of the preventive maintenance of the systems related with the safety, to extend them from twelve to eighteen months. The mentioned applications were founded in economic aspects and in inclining to a decrease in the doses of the workers that perform in the activities that are carried out in the programmed stops. The adopted position by the Nuclear Regulatory Authority to decide on these applications was based on the result of diverse evaluations that included the use of the Probabilistic Analysis of Safety specific of each power station, the operative experience resultant of the execution of the preventive maintenance program, and of the results of the repetitive tests and of the inspections in service. The regulatory decisions were different in each case. Indeed, the Embalse nuclear power station was authorized by the Regulatory Authority to modify from twelve to eighteen months the period among the realization of the repetitive tests and of the preventive maintenance, conditioned to the execution of some specific regulatory requirements. On the other hand, the Atucha I nuclear power station was not authorized to modify this period. In this presentation that is detailed the acted by the Nuclear Regulatory Authority in both cases, the used analysis tools, and the foundation of the adopted decisions. (Author)

  16. Nuclear model parameter testing for nuclear data evaluation (Reference Input Parameter Library: Phase II). Summary report of the third research co-ordination meeting

    International Nuclear Information System (INIS)

    Herman, M.

    2002-04-01

    This report summarises the results and recommendations of the third Research Co-ordination Meeting on improving and testing the Reference Input Parameter Library: Phase II. A primary aim of the meeting was to review the achievements of the CRP, to assess the testing of the library and to approve the final contents. Actions were approved that will result in completion of the file and a draft report by the end of February 2002. Full release of the library is scheduled for July 2002. (author)

  17. Nuclear model parameter testing for nuclear data evaluation (Reference Input Parameter Library: Phase II). Summary report of the second research co-ordination meeting

    International Nuclear Information System (INIS)

    Herman, M.

    2000-09-01

    This report summarizes the results and recommendations of the Second Research Coordination Meeting on Testing and Improvement of the Reference Input Parameter Library: Phase II. A primary aim of this meeting was to review progress in the CRP work, to review results of testing the library, to establish the RIPL-2 format and to decide on the contents of the library. The actions were agreed with an aim to complete the project by the end of 2001. Separate abstracts were prepared for 10 individual papers

  18. Advanced training course on state systems of accounting for and control of nuclear materials. Volume II. Visual aids

    International Nuclear Information System (INIS)

    Sorenson, R.J.; Schneider, R.A.

    1979-01-01

    Purpose of the course was to train in the accounting and control of nuclear materials in a bulk processing facility, for international safeguards. The Exxon low enriched uranium fabrication plant is used as an example. This volume contains visual aids used for the presentation

  19. Regulatory activities related with the modification of the frequency of the programmed stoppings of the Argentine nuclear centrals

    International Nuclear Information System (INIS)

    Marino, E.; Calvo, J.; Waldman, R.; Navarro, R.

    2006-01-01

    The mandatory character documentation of the Argentinean nuclear power stations in Embalse and Atucha I, required the realization of a programmed stoppings every twelve months to execute that settled down in the maintenance and surveillance programs for each installation. Nucleoelectrica Argentina S.A., in it character of Responsible Entity of the operation of these power stations, requested to the Argentinean Nuclear Regulatory Authority, in 2003 and 2005 respectively, the authorization to change the period of the repetitive tests and of the preventive maintenance of the systems related with the safety, to extend them from twelve to eighteen months. The mentioned applications were founded in economic aspects and in inclining to a decrease in the doses of the workers that perform in the activities that are carried out in the programmed stops. The adopted position by the Nuclear Regulatory Authority to decide on these applications was based on the result of diverse evaluations that included the use of the Probabilistic Analysis of Safety specific of each power station, the operative experience resultant of the execution of the preventive maintenance program, and of the results of the repetitive tests and of the inspections in service. The regulatory decisions were different in each case. Indeed, the Embalse nuclear power station was authorized by the Regulatory Authority to modify from twelve to eighteen months the period among the realization of the repetitive tests and of the preventive maintenance, conditioned to the execution of some specific regulatory requirements. On the other hand, the Atucha I nuclear power station was not authorized to modify this period. In this presentation that is detailed the acted by the Nuclear Regulatory Authority in both cases, the used analysis tools, and the foundation of the adopted decisions. (Author)

  20. Joint thesaurus Part I (A-L) + II (M-Z)[International Nuclear Information System. Energy Technology Data Exchange

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2007-04-15

    This is the second revision of the ETDE/INIS Joint Thesaurus, including all updates up to September 2006. It contains 21 147 valid descriptors and 9 114 forbidden terms. The Joint Thesaurus contains the controlled terminology for indexing all information within the subject scopes of the International Nuclear Information System (INIS) and the Energy Technology Data Exchange (ETDE). The terminology is intended for use in subject descriptions for input or retrieval of information in these systems. The thesaurus is a terminological control device used in translating from the natural language of documents, indexers or users into a more constrained system language It is also a controlled and dynamic vocabulary of semantically and generically related terms which covers a specific domain of knowledge. The basic terminology in this thesaurus goes back to the 1969 edition of the EURATOM Thesaurus. The structure subsequently given to that terminology was the result of a systematic study performed by INIS subject specialists. Further expansion of the thesaurus terminology was done by ETDE to incorporate information on all forms of energy. The ETDE/INIS Joint Thesaurus is the result of continued editing, carried out in parallel to the processing of the INIS and ETDE databases. The domain of knowledge covered by the Joint Thesaurus includes physics (in particular, plasma physics, atomic and molecular physics, and especially nuclear and high-energy physics), chemistry, materials science, earth sciences, radiation biology, radioisotope effects and kinetics, applied life sciences, radiology and nuclear medicine, isotope and radiation source technology, radiation protection, radiation applications, engineering, instrumentation, fossil fuels, synthetic fuels, renewable energy sources, advanced energy systems, fission and fusion reactor technology, safeguards and inspection, waste management, environmental aspects of the production and consumption of energy from nuclear and non-nuclear

  1. Research work with TRIGA Mark II at the Nuclear Chemistry Section of the 'J. Stefan' Institute in Ljubljana

    International Nuclear Information System (INIS)

    Byrne, A.R.; Dermelj, M.; Kosta, L.; Ravkin, V.; Stegnar, P.

    1978-01-01

    The general features of our research programme using TRIGA MK II, as outlined at the last TRIGA Reactor Users Conference in Vienna, September 28-30,1976, remain the same; namely, neutron activation analysis for trace and some minor elements. The four main areas presently investigated are a) environmental studies, b) life sciences research, c) standardization and d) methodology for specific problems arising in the first three topics

  2. Rise, fall and resurrection of chromosome territories: a historical perspective Part II. Fall and resurrection of chromosome territories during the 1950s to 1980s. Part III. Chromosome territories and the functional nuclear architecture: experiments and m

    OpenAIRE

    T Cremer; C Cremer

    2009-01-01

    Part II of this historical review on the progress of nuclear architecture studies points out why the original hypothesis of chromosome territories from Carl Rabl and Theodor Boveri (described in part I) was abandoned during the 1950s and finally proven by compelling evidence forwarded by laser-uvmicrobeam studies and in situ hybridization experiments. Part II also includes a section on the development of advanced light microscopic techniques breaking the classical Abbe limit written for reade...

  3. FISA-2009 Conference on Euratom Research and Training Activities: Nuclear Fission - Past, Present and Future (Generation-II, -III and -IV + Partitioning and Transmutation)

    International Nuclear Information System (INIS)

    Bhatnagar, V.; Deffrennes, M.; Hugon, M.; Manolatos, P.; Ptackova, K.; Van Goethem, G.; Webster, S.

    2011-01-01

    This paper is an introduction to the research and training activities carried out under the Euratom 7th Framework Programme (FP7, 2007-2011) in the field of nuclear fission science and technology, covering in particular nuclear systems and safety, and including innovative reactor systems and partitioning and transmutation. It is based on the more than 40 invited lectures that were delivered by Euratom project coordinators and keynote speakers at the FISA-2009 Conference (), organised by the European Commission DG Research, 22-24 June 2009, Prague, Czech Republic. The Euratom programme must be considered in the context of current and future nuclear technology and the respective research effort: ·Generation-II (i.e. yesterday, NPP construction 1970-2000): safety and reliability of nuclear facilities and energy independence in order to ensure security of supply worldwide; ·Generation-III (i.e. today, construction 2000-2040+): continuous improvement of safety and reliability, and increased industrial competitiveness in a growing energy market; ·Generation-IV (i.e. tomorrow, construction from 2040) for increased sustainability though optimal utilisation of natural resources and waste minimisation, and increased proliferation resistance. Consequently, the focus of the lectures devoted to Generation-II and -III is on the major scientific challenges and technological developments needed to guarantee safety and reliability, in particular issues associated with plant lifetime extension and operation. The focus of the lectures devoted to Generation-IV is on the design objectives and associated research issues that have been agreed upon internationally, in particular the ambitious criteria and technology goals established at the international level by the Generation-IV International Forum (GIF). In the future, electricity must continue to be produced competitively, and in addition high temperature process heat may also be required, while exploiting a maximum of fissile and

  4. Ceramic nuclear waste forms. II. A ceramic-waste composite prepared by hot pressing. Progress report and preprint

    International Nuclear Information System (INIS)

    McCarthy, G.J.

    1975-01-01

    A feasibility study was conducted to determine whether nuclear waste calcine and a crystalline ceramic matrix can be fabricated by hot pressing into a composite waste form with suitable leaching resistance and thermal stability. It was found that a hard, dense composite could be formed using the typical commercial waste formulation PW-4b and a matrix of α-quartz with a small amount of a lead borosilicate glass added as a consolidation aide. Its density, waste loading, and leaching resistance are comparable to the glasses currently being considered for fixation of nuclear wastes. The hot pressed composite offers a closer approach to thermodynamic stability and improved thermal stability (in monolithic form) compared to glass waste forms. Recommendations for further optimization of the hot pressed waste form are given. (U.S.)

  5. Study pertaining to the distribution of iodine pills in the event of a nuclear mishap at the Gentilly II Station

    International Nuclear Information System (INIS)

    Corriveau, R.

    1992-01-01

    This study seeks to understand how volunteers, whose task it is to distribute iodine pills in the event of a nuclear mishap, are likely to react in such a situation. Our postulate is that the uniform application of preventative measures in an emergency situation requires that volunteers adhere to the principles of the ideological apparatus (civil authority). Our findings are that current measures are inadequate for an effective emergency strategy. (author)

  6. Nuclear fuel element design and thermal-hydraulic analysis of Wolsung-1, 600 MWe CANDU-PHWR (Part II)

    International Nuclear Information System (INIS)

    Suk, H.C; Lee, J.C.; Suh, K.S.; Yuk, K.E.; Whang, W.; Park, J.S.; Eim, J.S.; Bang, K.H.; Eim, M.S.; Rim, C.S.

    1982-01-01

    The main objective of the present thermal hydraulic analysis is to determine the thermal hydraulic characteristics of Wolsung-1 600 MWe CANDU-PHW reactor under normal operation. This is to verify and expedite the development of the nuclear fuel design and fabrication as well as the management. The computer program package developed for the stated objective are DOD81, CANREPP, PLOC81 and COBRA-CANDU. (Author)

  7. Belgian class II nuclear facilities such as irradiators and accelerators. Regulatory Body attention points and operating experience feedback

    Energy Technology Data Exchange (ETDEWEB)

    Minne, Etienne; Peters, Christelle; Mommaert, Chantal; Kennes, Christian; Cortenbosch, Geert; Schmitz, Frederic; Haesendonck, Michel van [Bel V, Brussels (Belgium); Carlier, Pascal; Schrayen, Virginie; Wertelaers, An [Federal Agency for Nuclear Control, Brussels (Belgium)

    2016-11-15

    The aim of this paper is to present the Regulatory Body attention points and the operating experience feedback from Belgian ''class IIA'' facilities such as industrial and research irradiators, bulk radionuclides producers and conditioners. Reinforcement of the nuclear safety and radiation protection has been promoted by the Federal Agency for Nuclear Control (FANC) since 2009. This paper is clearly a continuation of the former paper [1] presenting the evolution in the regulatory framework relative to the creation of Bel V, the subsidiary of the FANC, and to the new ''class IIA'' covering heavy installations such as those mentioned above. Some lessons learnt are extracted from the operating experience feedback based on the events declared to the authorities. Even though a real willingness to meet the new safety requirements is observed among the ''class IIA'' licensees, promoting the safety culture, the nuclear safety and radiation protection remains an endless challenge for the Regulatory Body.

  8. Development of Intelligent Database Program for PSI/ISI Data Management of Nuclear Power Plant (Part II)

    International Nuclear Information System (INIS)

    Park, Un Su; Park, Ik Keun; Um, Byong Guk; Lee, Jong Po; Han, Chi Hyun

    2000-01-01

    In a previous paper, we have discussed the intelligent Windows 95-based data management program(IDPIN) which was developed for effective and efficient management of large amounts of pre-/in-service inspection(PSI/ISI) data of Kori nuclear power plants. The IDPIN program enables the prompt extraction of previously conducted PSI/ISI conditions and results so that the time-consuming data management, painstaking data processing and analysis of the past are avoided. In this study, the intelligent Windows based data management program(WS-IDPIN) has been developed as an effective data management of PSI/ISI data for the Wolsong nuclear power plants. The WS-IDPIN program includes the modules of comprehensive management and analysis of PSI/ISI results, statistical reliability assessment program of PSI/ISI results(depth and length sizing performance etc), standardization of UT report form and computerization of UT results. In addition, the program can be further developed as a unique PSI/ISI data management expert system which can be part of the PSI/ISI total support system for Korean nuclear power plants

  9. I. Nuclear and neutron matter calculations with isobars. II. A model calculation of Fermi liquid parameters for liquid 3He

    International Nuclear Information System (INIS)

    Ainsworth, T.L.

    1983-01-01

    The Δ(1232) plays an important role in determining the properties of nuclear and neutron matter. The effects of the Δ resonance are incorporated explicitly by using a coupled channel formalism. A method for constraining a lowest order variational calculation, appropriate when nucleon internal degrees of freedom are made explicity, is presented. Different N-N potentials were calculated and fit to phase shift data and deuteron properties. The potentials were constructed to test the relative importance of the Δ resonance on nuclear properties. The symmetry energy and incompressibility of nuclear matter are generally reproduced by this calculation. Neutron matter results lead to appealing neutron star models. Fermi liquid parameters for 3 He are calculated with a model that includes both direct and induced terms. A convenient form of the direct interaction is obtained in terms of the parameters. The form of the direct interaction ensures that the forward scattering sum rule (Pauli principle) is obeyed. The parameters are adjusted to fit the experimentally determined F 0 /sup s/, F 0 /sup a/, and F 1 /sup s/ Landau parameters. Higher order Landau parameters are calculated by the self-consistent solution of the equations; comparison to experiment is good. The model also leads to a preferred value for the effective mass of 3 He. Of the three parameters only one shows any dependence on pressure. An exact sum rule is derived relating this parameter to a specific summation of Landau parameters

  10. Thermodynamics. Vol. II. Proceedings of the Symposium on Thermodynamics with Emphasis on Nuclear Materials and Atomic Transport in Solids

    International Nuclear Information System (INIS)

    1966-01-01

    Knowledge of the thermodynamics of nuclear materials is vital to the design of reactor fuels and moderating and cooling systems, in fact all facets of nuclear plant operation that involve mixtures of, or contact between, two or more elements in single- or multi-phase systems. The steep thermal gradients and the high temperatures involved in nuclear technology pose special problems for engineers and thermodynamicists, who have found that extrapolation of low-temperature data to high temperatures very often proves invalid. For this reason, standard thermodynamic techniques such as calorimetry and EMF-methods have been extended into high-temperature regions. Since the Agency's last conference on this subject, also held in Vienna (Thermodynamics of Nuclear Materials, 1962), there have been notable advances in calorimetry performed at temperatures greater than 1000°C, and in the use of EMF cells with solid electrolytes operated at similar temperatures. Significant advances have also been made in measuring diffusion parameters at the higher temperatures. An important field covered in this Symposium was the correlation of such atomic transport data with thermodynamic data, a prerequisite if the nuclear engineer is to incorporate diffusion results into his normal process- assessment techniques. Finally the Symposium suggested the requirements for good critical tables. The mere compiling of such data is no longer sufficient; the compiler must have free access to all the data of a particular experiment, he must have an intimate knowledge of experimental work in this field and he must weight every figure quoted in the light of his experience. As a step in this direction, the Agency has called on the services of many well-known experts and is preparing a number of monographs giving critical assessments of thermodynamic data and phase-diagrams for many of the elements of interest in reactor design. Most of the countries engaged in research in thermodynamics were represented at

  11. Thermodynamics. Vol. II. Proceedings of the Symposium on Thermodynamics with Emphasis on Nuclear Materials and Atomic Transport in Solids

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1966-02-15

    Knowledge of the thermodynamics of nuclear materials is vital to the design of reactor fuels and moderating and cooling systems, in fact all facets of nuclear plant operation that involve mixtures of, or contact between, two or more elements in single- or multi-phase systems. The steep thermal gradients and the high temperatures involved in nuclear technology pose special problems for engineers and thermodynamicists, who have found that extrapolation of low-temperature data to high temperatures very often proves invalid. For this reason, standard thermodynamic techniques such as calorimetry and EMF-methods have been extended into high-temperature regions. Since the Agency's last conference on this subject, also held in Vienna (Thermodynamics of Nuclear Materials, 1962), there have been notable advances in calorimetry performed at temperatures greater than 1000 Degree-Sign C, and in the use of EMF cells with solid electrolytes operated at similar temperatures. Significant advances have also been made in measuring diffusion parameters at the higher temperatures. An important field covered in this Symposium was the correlation of such atomic transport data with thermodynamic data, a prerequisite if the nuclear engineer is to incorporate diffusion results into his normal process- assessment techniques. Finally the Symposium suggested the requirements for good critical tables. The mere compiling of such data is no longer sufficient; the compiler must have free access to all the data of a particular experiment, he must have an intimate knowledge of experimental work in this field and he must weight every figure quoted in the light of his experience. As a step in this direction, the Agency has called on the services of many well-known experts and is preparing a number of monographs giving critical assessments of thermodynamic data and phase-diagrams for many of the elements of interest in reactor design. Most of the countries engaged in research in thermodynamics were

  12. ISOLA II: a FORTRAN IV program for the calculation of long-term dose distribution in the vicinity of nuclear installations

    International Nuclear Information System (INIS)

    Huebschmann, W.; Nagel, D.

    The computer code ISOLA serves for the annual calculation of the radiation burden of the environment of the Nuclear Research Center at Karlsruhe resulting from the release of alpha-active and beta-active off-gases. In the improved version ISOLA II the model of a double Gaussian Distribution function is strictly-maintained, so that the influence due to neighboring sectors is included. The emissions are assumed to be constant in time during a given time period. The user may select either the print-out of an isodose map for a desired area (for example a map square 20 km on each edge) or he may obtain a list of doses for up to 2000 filed points (for example in the surrounding communities). The input and output forms will be shown by the use of an example

  13. ISOLA II - A FORTRAN IV code for the calculation of the long-term α- and β-dose distributions in the vicinity of nuclear installations

    International Nuclear Information System (INIS)

    Huebschmann, W.; Nagel, D.

    1975-12-01

    The computer code ISOLA is used to calculate the annual radiation doses caused by α- and β-active off-gases in the environment of the Karlsruhe Nuclear Research Center. In the revised version ISOLA II the double Gaussian distribution model is strictly observed. As a consequence, the contribution of activity from neighbour sectors is taken into account. Up to 15 emitters may be coped with simultaneously. The emission rates are considered to be constant during the given time interval. Optionally either the isodoses chart of a specified area (for instance a square 20 by 20 km) or a list of doses calculated at up to 2,000 locations (for instance the living areas) in the environment may be set up. Input and output are shown for a specific case. (orig.) [de

  14. JPRS Report Nuclear Developments

    Science.gov (United States)

    1988-09-02

    cracks in Atucha I were detected during the administration of Eng Alberto Constantini. Last year Constantini resigned as CNEA president due to...days, Finance Minister Mailson da Nobrega, Mines and Energy Minis- ter Aureliano Chaves, and Planning Minister Joao Batista de Abreu should be

  15. Irradiation positions for fission-track dating in the University of Pavia TRIGA Mark II nuclear reactor

    International Nuclear Information System (INIS)

    Oddone, Massimo; Meloni, Sandro; Balestrieri, Maria Laura; Bigazzi, Giulio

    2002-01-01

    An irradiation position arranged is described in the present paper for fission-track dating in the Triga Mark II reactor of the University of Pavia. Fluence values determined using the NIST glass standard SRM 962a for fission-track dating and the traditional metal foils are compared. Relatively good neutron thermalization (φ th /φ f = 0.956) and lack of significant fluence spatial gradients are good factors for fission-track dating. Finally, international age standards (or putative age standards) irradiated in this new position yielded results consistent with independent reference ages. (author)

  16. Applications of Nuclear Physics

    OpenAIRE

    Hayes, Anna C.

    2017-01-01

    Today the applications of nuclear physics span a very broad range of topics and fields. This review discusses a number of aspects of these applications, including selected topics and concepts in nuclear reactor physics, nuclear fusion, nuclear non-proliferation, nuclear-geophysics, and nuclear medicine. The review begins with a historic summary of the early years in applied nuclear physics, with an emphasis on the huge developments that took place around the time of World War II, and that und...

  17. Plasma Physics and Controlled Nuclear Fusion Research. Vol. II. Proceedings of a Conference on Plasma Physics and Controlled Physics Research

    International Nuclear Information System (INIS)

    1966-01-01

    Research on controlled nuclear fusion was first disclosed at the Second United Nations Conference on the Peaceful Uses of Atomic Energy, held at Geneva in 1958. From the information given, it was evident that a better understanding of the behaviour of hot dense plasmas was needed before the goal of economic energy release from nuclear fusion could be reached. The fact that research since then has been most complex and costly has enhanced the desirability of international co-operation and exchange of information and experience. Having organized its First Conference on Plasma Physics and Controlled Nuclear Fusion Research at Salzburg in 1961, the International Atomic Energy Agency again provided the means for such cooperation in organizing its Second Conference on this subject on 6-10 September, 1965, at Culham, Abingdon, Berks, England. The meeting was arranged with the generous help of the United Kingdom Atomic Energy Authority at their Culham Laboratory, where the facilities and assistance of the staff were greatly appreciated. At the meeting, which was attended by 268 participants from 26 member states and three international organizations, significant results from many experiments, including those from the new and larger machines, became available. It has now become feasible to intercorrelate data obtained from a number of similar machines; this has led to a more complete understanding of plasma behaviour. No breakthrough was reported nor had been expected towards the economical release of the energy from fusion, but there was increased understanding of the problems of production, control and containment of high-density and high-temperature plasmas

  18. Construction of a bibliographic information database and development of retrieval system for research reports in nuclear science and technology (II)

    International Nuclear Information System (INIS)

    Han, Duk Haeng; Kim, Tae Whan; Choi, Kwang; Yoo, An Na; Keum, Jong Yong; Kim, In Kwon

    1996-05-01

    The major goal of this project is to construct a bibliographic information database in nuclear engineering and to develop a prototype retrieval system. To give an easy access to microfiche research report, this project has accomplished the construction of microfiche research reports database and the development of retrieval system. The results of the project are as follows; 1. Microfiche research reports database was constructed by downloading from DOE Energy, NTIS, INIS. 2. The retrieval system was developed in host and web version using access point such as title, abstracts, keyword, report number. 6 tabs., 8 figs., 11 refs. (Author) .new

  19. Construction of a bibliographic information database and development of retrieval system for research reports in nuclear science and technology (II)

    Energy Technology Data Exchange (ETDEWEB)

    Han, Duk Haeng; Kim, Tae Whan; Choi, Kwang; Yoo, An Na; Keum, Jong Yong; Kim, In Kwon [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1996-05-01

    The major goal of this project is to construct a bibliographic information database in nuclear engineering and to develop a prototype retrieval system. To give an easy access to microfiche research report, this project has accomplished the construction of microfiche research reports database and the development of retrieval system. The results of the project are as follows; 1. Microfiche research reports database was constructed by downloading from DOE Energy, NTIS, INIS. 2. The retrieval system was developed in host and web version using access point such as title, abstracts, keyword, report number. 6 tabs., 8 figs., 11 refs. (Author) .new.

  20. A study on nuclear heat load tolerable for NET/TF coils cooled by internal flow of helium II

    International Nuclear Information System (INIS)

    Hofmann, A.

    1988-02-01

    NbTi cables cooled by internal flow of superfluid helium are considered an option for the design of NET/TF coils with about 11 T peak fields. Starting from an available winding cross section of 0.61x0.61 m 2 for a 8 MA turns coil made of a 16 kA conductor it is shown that sufficient hydraulic cross section can be provided within such cables to remove the expected thermal load resulting from nuclear heating with exponential decay from inboard to outboard side of the winding. The concept is a pancake type coil with 1.8 K helium fed-in the high field region of each pancake. The temperature distribution within such coils is calculated, and the local safety margin is determined from temperature and field. The calculation takes account of nuclear and a.c. heating, and of thermal conductance between the individual layers and the coil casing. It is shown that operation with 1.8 K inlet and about 3 K outlet temperature is possible. The electrical insulation with about 0.5 mm thickness proves to provide sufficient thermal insulation. No additional thermal shield is required between the coil casing and the winding package. Two different types of conductors are being considered: a) POLO type cable with quadratic cross section and a central circular coolant duct, and b) an LCT type cable with two conductors wound in hand. Both concepts with about 500 m length of the cooland channels are shown to meet the requirements resulting from a peak nuclear heat load of 0.3 mW/cm 3 in the inboard turns. The hydraulic diameters are sufficient to operate each coils with self-sustained fountain effect pumps. Even appreciably higher heat loads with up to 3 mW/cm 3 of nuclear heating can be tolerated for the POLO type cable when the hydraulic diameter is enlarged to its maximum of 17 mm. (orig.) [de

  1. Optimization in the nuclear fuel cycle II: Concentration of alpha emitters in the air; Otimização no ciclo do combustível nuclear II: concentração de alfa emissores no ar

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, W.S., E-mail: pereiras@gmail.com [Universidade Veiga de Ameida (UVA), Rio de Janeiro, RJ (Brazil); Silva, A.X.; Lopes, J.M.; Carmo, A.S.; Mello, C.R.; Fernandes, T.S., E-mail: lararapls@hotmail.com, E-mail: Ademir@nuclear.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil); Kelecom, A. [Universidade Federal Fluminense (UFF), Niterói, RJ (Brazil)

    2017-07-01

    Optimization is one of the bases of radioprotection and aims to move doses away from the dose limit that is the borderline of acceptable radiological risk. The work aims to use the monitoring of the concentration of alpha emitters in the air as a tool of the optimization process. We analyzed 27 sampling points of airborne alpha concentration in a nuclear fuel cycle facility. The monthly averages were considered statistically different, the highest in the month of February and the lowest in the month of August. All other months were found to have identical mean activity concentration values. Regarding the sampling points, the points with the highest averages were points 12, 15 and 9. These points were indicated for the beginning of the optimization process. Analysis of the production of the facility should be performed to verify possible correlations between production and concentration of alpha emitters in the air.

  2. Structure and conformational dynamics of the domain 5 RNA hairpin of a bacterial group II intron revealed by solution nuclear magnetic resonance and molecular dynamics simulations.

    Science.gov (United States)

    Pechlaner, Maria; Sigel, Roland K O; van Gunsteren, Wilfred F; Dolenc, Jožica

    2013-10-08

    Nuclear magnetic resonance (NMR) nuclear Overhauser enhancement (NOE) data obtained for a 35-nucleotide RNA segment of a bacterial group II intron indicate a helical hairpin structure in which three parts, a terminal pentaloop, a bulge, and a G-A mismatch, display no Watson-Crick base pairing. The 668 NOE upper distance bounds for atom pairs are insufficient to uniquely determine the conformation of these segments. Therefore, molecular dynamics simulations including time-averaged distance restraints have been used to obtain a conformational ensemble compatible with the observed NMR data. The ensemble shows alternating hydrogen bonding patterns for the mentioned segments. In particular, in the pentaloop and in the bulge, the hydrogen bonding networks correspond to distinct conformational clusters that could not be captured by using conventional single-structure refinement techniques. This implies that, to obtain a realistic picture of the conformational ensemble of such flexible biomolecules, it is necessary to properly account for the conformational variability in the structure refinement of RNA fragments.

  3. Comparison between dispersed nuclear power plants and a nuclear energy center at a hypothetical site on Kentucky Lake, Tennessee. Volume II. Transmission of power

    International Nuclear Information System (INIS)

    Reister, D.B.; Zelby, L.W.

    1976-05-01

    A comparison is made among power transmission systems required to serve a single set of load center demands from four modes of siting the generating facilities: a single generation site with an ultimate generation capacity of 48,000 MW; four generation sites each with a generation capacity of 12,000 MW; 10 generation sites each with a generation capacity of 4,800 MW; and a system that resulted when the existing utility plan for future generation was logically expanded. The time period for the study is from the year 1985 to the year 2020, when the full 48,000 MW of new capacity from the single large nuclear energy center is on-line. The load centers served are Huntsville, Alabama; Evansville, Indiana; Paducah, Kentucky; and Chattanooga, Nashville, and Memphis, Tennessee. Generation sites are real locations but are hypothetical in terms of miles of transmission lines, the product of the amount of power transmitted and the distance transmitted (GW-miles), and cost

  4. Physical and welding metallurgy of Gd-enriched austenitic alloys for spent nuclear fuel applications. Part II, nickel base alloys

    International Nuclear Information System (INIS)

    Mizia, Ronald E.; Michael, Joseph Richard; Williams, David Brian; Dupont, John Neuman; Robino, Charles Victor

    2004-01-01

    The physical and welding a metallurgy of gadolinium- (Gd-) enriched Ni-based alloys has been examined using a combination of differential thermal analysis, hot ductility testing. Varestraint testing, and various microstructural characterization techniques. Three different matrix compositions were chosen that were similar to commercial Ni-Cr-Mo base alloys (UNS N06455, N06022, and N06059). A ternary Ni-Cr-Gd alloy was also examined. The Gd level of each alloy was ∼2 wt-%. All the alloys initiated solidification by formation of primary austenite and terminated solidification by a Liquid γ + Ni 5 Gd eutectic-type reaction at ∼1270 C. The solidification temperature ranges of the alloys varied from ∼100 to 130 C (depending on alloy composition). This is a substantial reduction compared to the solidification temperature range to Gd-enriched stainless steels (360 to 400 C) that terminate solidification by a peritectic reaction at ∼1060 C. The higher-temperature eutectic reaction that occurs in the Ni-based alloys is accompanied by significant improvements in hot ductility and solidification cracking resistance. The results of this research demonstrate that Gd-enriched Ni-based alloys are excellent candidate materials for nuclear criticality control in spent nuclear fuel storage applications that require production and fabrication of large amounts of material through conventional ingot metallurgy and fusion welding techniques

  5. Phase II, Title I engineering assessment of inactive uranium mill tailings, Phillips/United Nuclear Site, Ambrosia Lake, New Mexico

    International Nuclear Information System (INIS)

    1977-12-01

    An engineering assessment was performed of the problems resulting from the existence of radioactive uranium mill tailings at the Phillips/United Nuclear site at Ambrosia Lake, New Mexico. Services included the preparation of topographic maps, the performance of core drillings sufficient to determine areas and volumes of tailings, and radiometric measurements to determine radium-contaminated materials, the evaluation of resulting radiation exposures of individuals and nearby populations, the investigation of site hydrology and meteorology, and the evaluation and costing of alternative corrective actions. Radon gas release from the 2.6 million tons of tailings at the Phillips/United Nuclear site constitutes the most significant environmental impact, although windblown tailings and external gamma radiation are also factors. The estimated radiological health effects to the general population are considered to be minimal. The two alternative actions presented are: dike stabilization, fencing, and maintenance; and adding 2 ft of stabilization cover material. Both options include remedial action at off-site structures and on-site decontamination around the tailings pile. Cost estimates for the two options are $920,000 and $2,230,000, respectively

  6. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumentation and measurement techniques in fuel fabrication facilities

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-01-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. A general discussion is given of instrumentation and measurement techniques which are presently used being considered for fuel fabrication facilities. Those aspects which are most significant from the point of view of satisfying regulatory constraints have been emphasized. Sensors and measurement devices have been discussed, together with their interfacing into a computerized system designed to permit real-time data collection and analysis. Estimates of accuracy and precision of measurement techniques have been given, and, where applicable, estimates of associated costs have been presented. A general description of material control and accounting is also included. In this section, the general principles of nuclear material accounting have been reviewed first (closure of material balance). After a discussion of the most current techniques used to calculate the limit of error on inventory difference, a number of advanced statistical techniques are reviewed. The rest of the section deals with some regulatory aspects of data collection and analysis, for accountability purposes, and with the overall effectiveness of accountability in detecting diversion attempts in fuel fabrication facilities. A specific example of application of the accountability methods to a model fuel fabrication facility is given. The effect of random and systematic errors on the total material uncertainty has been discussed, together with the effect on uncertainty of the length of the accounting period

  7. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumentation and measurement techniques in fuel fabrication facilities

    Energy Technology Data Exchange (ETDEWEB)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-01-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. A general discussion is given of instrumentation and measurement techniques which are presently used being considered for fuel fabrication facilities. Those aspects which are most significant from the point of view of satisfying regulatory constraints have been emphasized. Sensors and measurement devices have been discussed, together with their interfacing into a computerized system designed to permit real-time data collection and analysis. Estimates of accuracy and precision of measurement techniques have been given, and, where applicable, estimates of associated costs have been presented. A general description of material control and accounting is also included. In this section, the general principles of nuclear material accounting have been reviewed first (closure of material balance). After a discussion of the most current techniques used to calculate the limit of error on inventory difference, a number of advanced statistical techniques are reviewed. The rest of the section deals with some regulatory aspects of data collection and analysis, for accountability purposes, and with the overall effectiveness of accountability in detecting diversion attempts in fuel fabrication facilities. A specific example of application of the accountability methods to a model fuel fabrication facility is given. The effect of random and systematic errors on the total material uncertainty has been discussed, together with the effect on uncertainty of the length of the accounting period.

  8. Nuclear Electric Propulsion mission engineering study covering the period April 1971 to January 1973. Volume II. Final report

    International Nuclear Information System (INIS)

    1973-03-01

    The results of a mission engineering analysis of nuclear-thermionic electric propulsion spacecraft for unmanned interplanetary and geocentric missions are summarized. Critical technologies assessed are associated with the development of Nuclear Electric Propulsion (NEP), and the impact of its availability on future space programs. Specific areas of investigation include outer planet and comet rendezvous mission analysis, NEP Stage design for geocentric and interplanetary missions NEP system development cost and unit costs, and technology requirements for NEP Stage development. A multi-mission NEP Stage can be developed to perform both multiple geocentric and interplanetary missions. Development program costs for a 1983 launch would be of the order of $275 M, including hardware and reactor development, flight system hardware, and mission support. Recurring unit costs for flight NEP systems would be of the order of $25 M for a 120kWe NEP Stage. Identified pacing NEP technology requirements are the development of 20,000 full power hour ion thrusters and thermionic reactor, and the development of related power conditioning. The resulting NEP Stage design provides both inherent reliability and high payload mass capability. High payload mass capability can be translated into both low payload cost and high payload reliability. NEP Stage and payload integration is compatible with the Space Shuttle

  9. The industrial nuclear fuel cycle in Argentina

    International Nuclear Information System (INIS)

    Koll, J.H.; Kittl, J.E.; Parera, C.A.; Coppa, R.C.; Aguirre, E.J.

    1977-01-01

    The nuclear power program of Argentina for the period 1976-85 is described, as a basis to indicate fuel requirements and the consequent implementation of a national fuel cycle industry. Fuel cycle activities in Argentina were initiated as soon as 1951-2 in the prospection and mining activities through the country. Following this step, yellow-cake production was initiated in plants of limited capacity. National production of uranium concentrate has met requirements up to the present time, and will continue to do so until the Sierra Pintada Industrial Complex starts operation in 1979. Presently, there is a gap in local production of uranium dioxide and fuel elements for the Atucha power station, which are produced abroad using Argentine uranium concentrate. With its background, the argentine program for the installation of nuclear fuel cycle industries is described, and the techno-economical implications considered. Individual projects are reviewed, as well as the present and planned infrastructure needed to support the industrial effort [es

  10. Systematic evaluation program review of NRC Safety Topic VI-7.3 associated with the electrical, instrumentation and control portions of the ECCS actuation system for the Dresden II Nuclear Power Plant

    International Nuclear Information System (INIS)

    St Leger-Barter, G.

    1980-11-01

    This report documents the technical evaluation and review of NRC Safety Topic VI-7.A.3, associated with the electrical, instrumentation, and control portions of the classification of the ECCS actuation system for the Dresden II nuclear power plant, using current licensing criteria

  11. Systematic evaluation program review of NRC safety topic VII-2 associated with the electrical, instrumentation and control portions of the ESF system control logic and design for the Dresden Station, Unit II nuclear power plant

    International Nuclear Information System (INIS)

    St Leger-Barter, G.

    1980-11-01

    This report documents the technical evaluation and review of NRC Safety Topic VII-2, associated with the electrical, instrumentation, and control portions of the ESF system control logic and design for the Dresden Station Unit II nuclear power plant, using current licensing criteria

  12. Comprehensive safety analysis code system for nuclear fusion reactors II: Thermal analysis during plasma disruptions for international thermonuclear experimental reactor

    International Nuclear Information System (INIS)

    Honda, T.; Maki, K.; Okazaki, T.

    1994-01-01

    Thermal characteristics of a fusion reactor [International Thermonuclear Experimental Reactor (ITER) Conceptual Design Activity] during plasma disruptions have been analyzed by using a comprehensive safety analysis code for nuclear fusion reactors. The erosion depth due to disruptions for the armor of the first wall depends on the current quench time of disruptions occurring in normal operation. If it is possible to extend the time up to ∼50 ms, the erosion depth is considerably reduced. On the other hand, the erosion depth of the divertor is ∼570 μm for only one disruption, which is determined only by the thermal flux during the thermal quench. This means that the divertor plate should be exchanged after about nine disruptions. Counter-measures are necessary for the divertor to relieve disruption influences. As other scenarios of disruptions, beta-limit disruptions and vertical displacement events were also investigated quantitatively. 13 refs., 5 figs

  13. Future Industrialization of the World and the Necessity of Nuclear Power, Part II: How Limited are Resources?

    International Nuclear Information System (INIS)

    Jovanovich, Jovan V.

    1997-01-01

    Will the future world be forever divided into an industrial, developed and 'rich' on one side, and the primitive, undeveloped, and poor on the other? Is an industrial, affluent and sustainable world of 10-15 billion people owning 5-10 billion cars physically possible to exist? Can the world have enough food, minerals and energy to support such a widespread affluence in a sustainable manner? In previous papers I have argued that even without any major breakthroughs in science and technology, an industrialized, sustainable and affluent world can be created within the next half a century to a century, but only if breeder nuclear power is widely used throughout the world. In this paper I elaborate on the question of future availability of some basic natural resources. (author)

  14. Fundamental Study of Electron Beam Welding of AA6061-T6 Aluminum Alloy for Nuclear Fuel Plate Assembly (II)

    International Nuclear Information System (INIS)

    Kim, Soosung; Lee, Haein; Lee, Donbae; Park, Jongman; Lee, Yoonsang

    2013-01-01

    Certain characteristics, such as solidification cracking, porosity, HAZ (Heat-affected Zone) degradation must be considered during welding. Because of high energy density and low heat input, especially LBW and EBW processes posses the advantage of minimizing the fusing zone and HAZ and producing deeper penetration than arc welding processes. In present study, to apply for the nuclear fuel plate fabrication and assembly, a fundamental EBW experiment using AA6061-T6 aluminum alloy specimens was conducted. Furthermore, to establish the welding process, and satisfy the requirements of the weld quality, EBW apparatus using a electron welding gun and vacuum chamber was developed, and preliminary investigations for optimizing the welding parameters of the specimens using AA6061-T6 aluminum plates were also performed. In this experiment, a feasibility test was carried out by tensile tester, bead-on-plate welding and metallographic examination to comply with the aluminum welding procedure. The EB weld quality of AA6061-T6 aluminum alloy for the fuel plate assembly has been also studied by the mechanical testing and microstructure examinations. This study was carried out to determine the suitable welding process and to investigate tensile strength of AA6061-T6 aluminum alloy. In the present experiment, satisfactory EBW of the square butt weld specimens was developed. In comparison with the rolling directions of test specimens, the tensile strengths were no difference between the longitudinal and transverse welds. Based on this fundamental study, fabrication and assembly of the nuclear fuel plates will be provided for the future Kijang research reactor project

  15. The flow effect in the irradiation embrittlement in pressure vessel steels of nuclear power plants

    International Nuclear Information System (INIS)

    Kempf, Rodolfo A.; Cativa Tolosa, Sebastian; Fortis, Ana M.

    2009-01-01

    This paper deals with the advances in the study of the mechanical behavior of the Reactor Pressure Vessel steels under accelerate irradiations. The objective is to study the effect of lead factors on the interpretation of the mechanisms that induced the embrittlement of the RPV, like those of the reactors Atucha II and CAREM. It is described a device designed to irradiate Charpy specimens with V notch of SA-508 type 3 steel at power reactor temperature, installed in the RA-1 reactor. It is presented also an automatic digital image processing technique for partitioning Charpy fracture surface into regions with a clear physical meaning and appropriate for the work in hot cells. The aim is to obtain the fracture behavior of irradiated specimens with different lead factors in the range of high fluencies and to know the dependence with the composition of the alloy and with the diffusion of other alloy elements. (author)

  16. Environmental monitoring in the surroundings of nuclear power plants of Argentina during 1996 and 1997

    International Nuclear Information System (INIS)

    Canoba, A.C.; Lopez, F.O.; Bruno, H.A.

    1998-01-01

    During 1996 and 1997, the environmental monitoring program in the surroundings of nuclear power plants Atucha 1 (CNA-1) and Embalse (CNE) was performed. For the selection of the sample points, sample type and frequency, recommendations were taken into account, as well as the major pathways of exposure to man. The results of the measurements were, in general, below the lower limit of detection and the doses for each critical group were, in the case of CNA I, below 4.9 E-3 mSv during 1996 and below 4.5 E-3 mSv during 1997, and for CNE, below 7.7 E-3 mSv during 1996 and below 9.4 mSv during 1997. All of these values are well below the dose limits established by todays norms (1mSv/year). (author) [es

  17. Fault diagnosis of generation IV nuclear HTGR components – Part II: The area error enthalpy–entropy graph approach

    International Nuclear Information System (INIS)

    Rand, C.P. du; Schoor, G. van

    2012-01-01

    Highlights: ► Different uncorrelated fault signatures are derived for HTGR component faults. ► A multiple classifier ensemble increases confidence in classification accuracy. ► Detailed simulation model of system is not required for fault diagnosis. - Abstract: The second paper in a two part series presents the area error method for generation of representative enthalpy–entropy (h–s) fault signatures to classify malfunctions in generation IV nuclear high temperature gas-cooled reactor (HTGR) components. The second classifier is devised to ultimately address the fault diagnosis (FD) problem via the proposed methods in a multiple classifier (MC) ensemble. FD is realized by way of different input feature sets to the classification algorithm based on the area and trajectory of the residual shift between the fault-free and the actual operating h–s graph models. The application of the proposed technique is specifically demonstrated for 24 single fault transients considered in the main power system (MPS) of the Pebble Bed Modular Reactor (PBMR). The results show that the area error technique produces different fault signatures with low correlation for all the examined component faults. A brief evaluation of the two fault signature generation techniques is presented and the performance of the area error method is documented using the fault classification index (FCI) presented in Part I of the series. The final part of this work reports the application of the proposed approach for classification of an emulated fault transient in data from the prototype Pebble Bed Micro Model (PBMM) plant. Reference data values are calculated for the plant via a thermo-hydraulic simulation model of the MPS. The results show that the correspondence between the fault signatures, generated via experimental plant data and simulated reference values, are generally good. The work presented in the two part series, related to the classification of component faults in the MPS of different

  18. NKS-B NordRisk II: Nuclear risk from atmospheric dispersion in Northern Europe - Summary report

    International Nuclear Information System (INIS)

    Lauritzen, B.

    2011-05-01

    The objective of the NordRisk II project has been to derive practical means for assessing the risks from long-range atmospheric dispersion of radioactive materials. An atlas over different atmospheric dispersion and deposition scenarios has been developed using historical numerical weather prediction (NWP) model data. The NWP model data covers three years spanning the climate variability associated with the North Atlantic Oscillation, and the atlas considers radioactive releases from 16 release sites in and near the Nordic countries. A statistical analysis of the long-range dispersion and deposition patterns is undertaken to quantify the mean dispersion and deposition as well as the variability. Preliminary analyses show that the large-scale atmospheric dispersion and deposition is near-isotropic, irrespective of the release site and detailed climatology, and allows for a simple parameterization of the global dispersion and deposition patterns. The atlas and the underlying data are made available in a format compatible with the ARGOS decision support system, and have been implemented in ARGOS. (Author)

  19. NKS-B NordRisk II: Nuclear risk from atmospheric dispersion in Northern Europe - Summary report

    Energy Technology Data Exchange (ETDEWEB)

    Lauritzen, B. (Technical Univ. of Denmark, Risoe National Lab. for Sustainable Energy. Radiation Research Div., Roskilde (Denmark))

    2011-05-15

    The objective of the NordRisk II project has been to derive practical means for assessing the risks from long-range atmospheric dispersion of radioactive materials. An atlas over different atmospheric dispersion and deposition scenarios has been developed using historical numerical weather prediction (NWP) model data. The NWP model data covers three years spanning the climate variability associated with the North Atlantic Oscillation, and the atlas considers radioactive releases from 16 release sites in and near the Nordic countries. A statistical analysis of the long-range dispersion and deposition patterns is undertaken to quantify the mean dispersion and deposition as well as the variability. Preliminary analyses show that the large-scale atmospheric dispersion and deposition is near-isotropic, irrespective of the release site and detailed climatology, and allows for a simple parameterization of the global dispersion and deposition patterns. The atlas and the underlying data are made available in a format compatible with the ARGOS decision support system, and have been implemented in ARGOS. (Author)

  20. Fission product release from nuclear fuel II. Validation of ASTEC/ELSA on analytical and large scale experiments

    International Nuclear Information System (INIS)

    Brillant, G.; Marchetto, C.; Plumecocq, W.

    2013-01-01

    Highlights: • A wide range of experiments is presented for the ASTEC/ELSA code validation. • Analytical tests such as AECL, ORNL and VERCORS are considered. • A large-scale experiment, PHEBUS FPT1, is considered. • The good agreement with measurements shows the efficiency of the ASTEC modelling. • Improvements concern the FP release modelling from MOX and high burn-up UO 2 fuels. - Abstract: This article is the second of two articles dedicated to the mechanisms of fission product release from a degraded core. The models of fission product release from nuclear fuel in the ASTEC code have been described in detail in the first part of this work (Brillant et al., this issue). In this contribution, the validation of ELSA, the module of ASTEC that deals with fission product and structural material release from a degraded core, is presented. A large range of experimental tests, with various temperature and conditions for the fuel surrounding atmosphere (oxidising and reducing), is thus simulated with the ASTEC code. The validation database includes several analytical experiments with both bare fuel (e.g. MCE1 experiments) and cladded fuel (e.g. HCE3, VERCORS). Furthermore, the PHEBUS large-scale experiments are used for the validation of ASTEC. The rather satisfactory comparison between ELSA calculations and experimental measurements demonstrates the efficiency of the analytical models to describe fission product release in severe accident conditions

  1. Exxon Nuclear Company WREM-based generic PWR ECCS evaluation model. Appendix C to Volume II. Yankee Rowe example problem

    International Nuclear Information System (INIS)

    1975-01-01

    Core 12, of the Yankee Rowe (YR), plant is to be licensed with the Exxon Nuclear Co. PWR Evaluation Model. This appendix presents methodology and results of an example calculation for the YR plant using the ENC Evaluation Model. This example problem is for the double-ended guillotine cold leg break with a discharge coefficient of 0.6 assuming loss of one emergency diesel. The NSSS supplier has determined this case to give the highest peak cladding temperature (PCT) for Core 11. The YR example problem was performed to determine the maximum acceptable local peak heating rate (Kw/ft). The blowdown was performed with beginning-of-cycle (BOC) ENC fuel at full power with the hot assembly power corresponding to the design peak rod heating rate of 12.9 Kw/ft. The HOT CHANNEL, TOODEE2, and RELAP4-FLOOD runs were made at several reduced hot assembly radial peaks, holding axial peaking constant, until an acceptable PCT was achieved. This procedure results in a PCT of 1834 0 F at a reduced peak linear heating rate of 10.5 Kw/ft for the ENC fuel at BOC. (auth)

  2. CHICSi - a compact ultra-high vacuum compatible detector system for nuclear reaction experiments at storage rings. II. Detectors

    Energy Technology Data Exchange (ETDEWEB)

    Golubev, P.; Avdeichikov, V.; Carlen, L.; Jakobsson, B. E-mail: bo.jakobsson@kosufy.lu.se; Siwek, A.; Veldhuizen, E.J. van; Westerberg, L.; Whitlow, H.J

    2003-03-11

    We describe the detectors for identification of charged particles and fragments in CHICSi, a large solid angle multi-telescope system mounted inside an ultra-high vacuum (UHV), cluster-jet target chamber. CHICSi performs nuclear reaction experiments at storage rings. The telescopes consist of a first very thin, 10-14 {mu}m Si detector, a second 300 {mu}m (or possibly 500 {mu}m) ion implanted Si detector supplemented by a 6 mm GSO(Ce) scintillator read out by a photodiode (PD) or by a third 300 {mu}m Si detector. The telescopes provide full charge separation up to Z=17 and mass resolution up to A=9 in the energy range 0.7-60A MeV. The thin p-i-n diode detector, etched out from a 280 {mu}m Si wafer, and the GSO/PD detector, both exclusively developed for CHICSi, provide an energy resolution {<=}8%, while the standard 300 {mu}m detectors have {<=}2% energy resolution. Radiation stability of the Si detectors is confirmed up to an integrated flux of 10{sup 10} alpha particles. The GSO detector has 70% light collection efficiency with the optical coupling to the PD a simple open, 0.2 mm, gap. A new method, developed to perform absolute energy calibration for the GSO/PD detector is presented.

  3. CHICSi - a compact ultra-high vacuum compatible detector system for nuclear reaction experiments at storage rings. II. Detectors

    International Nuclear Information System (INIS)

    Golubev, P.; Avdeichikov, V.; Carlen, L.; Jakobsson, B.; Siwek, A.; Veldhuizen, E.J. van; Westerberg, L.; Whitlow, H.J.

    2003-01-01

    We describe the detectors for identification of charged particles and fragments in CHICSi, a large solid angle multi-telescope system mounted inside an ultra-high vacuum (UHV), cluster-jet target chamber. CHICSi performs nuclear reaction experiments at storage rings. The telescopes consist of a first very thin, 10-14 μm Si detector, a second 300 μm (or possibly 500 μm) ion implanted Si detector supplemented by a 6 mm GSO(Ce) scintillator read out by a photodiode (PD) or by a third 300 μm Si detector. The telescopes provide full charge separation up to Z=17 and mass resolution up to A=9 in the energy range 0.7-60A MeV. The thin p-i-n diode detector, etched out from a 280 μm Si wafer, and the GSO/PD detector, both exclusively developed for CHICSi, provide an energy resolution ≤8%, while the standard 300 μm detectors have ≤2% energy resolution. Radiation stability of the Si detectors is confirmed up to an integrated flux of 10 10 alpha particles. The GSO detector has 70% light collection efficiency with the optical coupling to the PD a simple open, 0.2 mm, gap. A new method, developed to perform absolute energy calibration for the GSO/PD detector is presented

  4. Proof of fatigue strength of nuclear components part II: Numerical fatigue analysis for transient stratification loading considering environmental effects

    International Nuclear Information System (INIS)

    Krätschmer, D.; Roos, E.; Schuler, X.; Herter, K.-H.

    2012-01-01

    For the construction, design and operation of nuclear components and systems the appropriate technical codes and standards provide detailed analysis procedures which guarantee a reliable behaviour of the structural components throughout the specified lifetime. Especially for cyclic stress evaluation the different codes and standards provide different fatigue analyses procedures to be performed considering the various mechanical and thermal loading histories and geometric complexities of the components. To consider effects of light water reactor coolant environments, new design curves included in report NUREG/CR-6909 for austenitic stainless steels and for low alloy steels have been presented. For the usage of these new design curves an environmental fatigue correction factor for incorporating environmental effects has to be calculated and used. The application of this environmental correction factor to a fatigue analysis of a nozzle with transient stratification loads, derived by in-service monitoring, has been performed. The results are used to compare with calculated usage factors, based on design curves without taking environmental effects particularly into account. - Highlights: ► We model an nozzle for fatigue analysis und mechanical and thermal loading conditions. ► A simplified as well as a general elastic–plastic fatigue analysis considering environmental effects is performed. ► The influence of different factors calculating the environmental factor F en are shown. ► The presented numerical evaluation methodology allows the consideration of all relevant parameters to assess lifetime.

  5. U.S. Department Of Energy's nuclear engineering education research: highlights of recent and current research-II. 5. Automation of Nuclear Fuel Pellet Quality Control

    International Nuclear Information System (INIS)

    Keyvan, Shahla; Song, Xiaolong

    2001-01-01

    At the present time, nuclear fuel pellet inspection is performed by humans using the naked eye for judgment and decision making as to whether to accept or reject the pellet. Unnecessary re-fabrication of pellets will be costly, and having too many low-quality pellets in a fuel assembly is unacceptable. The current practice of pellet inspection by humans is tedious and subject to inconsistencies and error. In addition, manual inspection is cumbersome since the inspector must keep the pellet at arm's length and must wear glasses to protect the lenses of his or her eyes. The pellets are taken from a pellet sizing machine, dumped onto a rack, and shaken into rows; they are then viewed as a group. The entire group is rotated 90 deg four times to provide the inspector with a 360-deg view of each pellet. The pellets are examined for certain types of cracks, chips, and unusual markings, i.e., water stains and machine banding. These defects appear at any location on the pellet surface image with different intensity, size, shape, and background noise. Figure 1 shows typical defective fuel pellets with chip, banded, and end defects. The goal of this work is to automate the pellet inspection process. A prototype of such an inspection system is developed. The system examines photographic images of pellets using various artificial intelligence techniques for image analysis and defect classification. Figure 2 shows the user interface of this inspection system, which is built using Java programming language. A total of 252 pellets with various defects was available for this research. Each pellet was photographed four times at rotations of 90 deg. The resultant black-and-white negatives were scanned into the computer in 256 gray scale mode. The inspection of a fuel pellet by image analysis involves several steps, as described in Fig. 3 and as follows: Step 1-On-line image conversion: This process involves on-line digitization of the input image. Step 2-Reference model: The second

  6. DJ-1 Modulates Nuclear Erythroid 2–Related Factor-2–Mediated Protection in Human Primary Alveolar Type II Cells in Smokers

    Science.gov (United States)

    Bahmed, Karim; Messier, Elise M.; Zhou, Wenbo; Tuder, Rubin M.; Freed, Curt R.; Chu, Hong Wei; Kelsen, Steven G.; Bowler, Russell P.; Mason, Robert J.

    2016-01-01

    Cigarette smoke (CS) is a main source of oxidative stress and a key risk factor for emphysema, which consists of alveolar wall destruction. Alveolar type (AT) II cells are in the gas exchange regions of the lung. We isolated primary ATII cells from deidentified organ donors whose lungs were not suitable for transplantation. We analyzed the cell injury obtained from nonsmokers, moderate smokers, and heavy smokers. DJ-1 protects cells from oxidative stress and induces nuclear erythroid 2–related factor-2 (Nrf2) expression, which activates the antioxidant defense system. In ATII cells isolated from moderate smokers, we found DJ-1 expression by RT-PCR, and Nrf2 and heme oxygenase (HO)-1 translocation by Western blotting and immunocytofluorescence. In ATII cells isolated from heavy smokers, we detected Nrf2 and HO-1 cytoplasmic localization. Moreover, we found high oxidative stress, as detected by 4-hydroxynonenal (4-HNE) (immunoblotting), inflammation by IL-8 and IL-6 levels by ELISA, and apoptosis by terminal deoxynucleotidyl transferase dUTP nick end labeling (TUNEL) assay in ATII cells obtained from heavy smokers. Furthermore, we detected early DJ-1 and late Nrf2 expression after ATII cell treatment with CS extract. We also overexpressed DJ-1 by adenovirus construct and found that this restored Nrf2 and HO-1 expression and induced nuclear translocation in heavy smokers. Moreover, DJ-1 overexpression also decreased ATII cell apoptosis caused by CS extract in vitro. Our results indicate that DJ-1 activates the Nrf2-mediated antioxidant defense system. Furthermore, DJ-1 overexpression can restore the impaired Nrf2 pathway, leading to ATII cell protection in heavy smokers. This suggests a potential therapeutic strategy for targeting DJ-1 in CS-related lung diseases. PMID:27093578

  7. A localized navigation algorithm for Radiation Evasion for nuclear facilities. Part II: Optimizing the “Nearest Exit” Criterion

    Energy Technology Data Exchange (ETDEWEB)

    Khasawneh, Mohammed A., E-mail: mkha@ieee.org [Department of Electrical Engineering, Jordan University of Science and Technology (Jordan); Al-Shboul, Zeina Aman M., E-mail: xeinaaman@gmail.com [Department of Electrical Engineering, Jordan University of Science and Technology (Jordan); Jaradat, Mohammad A., E-mail: majaradat@just.edu.jo [Department of Mechanical Engineering, Jordan University of Science and Technology (Jordan); Malkawi, Mohammad I., E-mail: mmalkawi@aimws.com [College of Engineering, Jadara University, Irbid 221 10 (Jordan)

    2013-06-15

    Highlights: ► A new navigation algorithm for Radiation Evasion around nuclear facilities. ► An optimization criteria minimized under algorithm operation. ► A man-borne device guiding the occupational worker towards paths that warrant least radiation × time products. ► Benefits of using localized navigation as opposed to global navigation schemas. ► A path discrimination function for finding the navigational paths exhibiting the least amounts of radiation. -- Abstract: In this extension from part I (Khasawneh et al., in press), we modify the navigation algorithm which was presented with the objective of optimizing the “Radiation Evasion” Criterion so that navigation would optimize the criterion of “Nearest Exit”. Under this modification, algorithm would yield navigation paths that would guide occupational workers towards Nearest Exit points. Again, under this optimization criterion, algorithm leverages the use of localized information acquired through a well designed and distributed wireless sensor network, as it averts the need for any long-haul communication links or centralized decision and monitoring facility thereby achieving a more reliable performance under dynamic environments. As was done in part I, the proposed algorithm under the “Nearest Exit” Criterion is designed to leverage nearest neighbor information coming in through the sensory network overhead, in computing successful navigational paths from one point to another. For comparison purposes, the proposed algorithm is tested under the two optimization criteria: “Radiation Evasion” and “Nearest Exit”, for different numbers of step look-ahead. We verify the performance of the algorithm by means of simulations, whereby navigational paths are calculated for different radiation fields. We, via simulations, also, verify the performance of the algorithm in comparison with a well-known global navigation algorithm upon which we draw our conclusions.

  8. A localized navigation algorithm for Radiation Evasion for nuclear facilities. Part II: Optimizing the “Nearest Exit” Criterion

    International Nuclear Information System (INIS)

    Khasawneh, Mohammed A.; Al-Shboul, Zeina Aman M.; Jaradat, Mohammad A.; Malkawi, Mohammad I.

    2013-01-01

    Highlights: ► A new navigation algorithm for Radiation Evasion around nuclear facilities. ► An optimization criteria minimized under algorithm operation. ► A man-borne device guiding the occupational worker towards paths that warrant least radiation × time products. ► Benefits of using localized navigation as opposed to global navigation schemas. ► A path discrimination function for finding the navigational paths exhibiting the least amounts of radiation. -- Abstract: In this extension from part I (Khasawneh et al., in press), we modify the navigation algorithm which was presented with the objective of optimizing the “Radiation Evasion” Criterion so that navigation would optimize the criterion of “Nearest Exit”. Under this modification, algorithm would yield navigation paths that would guide occupational workers towards Nearest Exit points. Again, under this optimization criterion, algorithm leverages the use of localized information acquired through a well designed and distributed wireless sensor network, as it averts the need for any long-haul communication links or centralized decision and monitoring facility thereby achieving a more reliable performance under dynamic environments. As was done in part I, the proposed algorithm under the “Nearest Exit” Criterion is designed to leverage nearest neighbor information coming in through the sensory network overhead, in computing successful navigational paths from one point to another. For comparison purposes, the proposed algorithm is tested under the two optimization criteria: “Radiation Evasion” and “Nearest Exit”, for different numbers of step look-ahead. We verify the performance of the algorithm by means of simulations, whereby navigational paths are calculated for different radiation fields. We, via simulations, also, verify the performance of the algorithm in comparison with a well-known global navigation algorithm upon which we draw our conclusions

  9. Human factors analysis and design methods for nuclear waste retrieval systems. Volume II. A compendium of human factors design data

    International Nuclear Information System (INIS)

    Casey, S.M.

    1980-04-01

    This document is a compilation of human factors engineering design recommendations and data, selected and organized to assist in the design of a nuclear waste retrieval system. Design guidelines from a variety of sources have been evaluated, edited, and expanded for inclusion in this document, and, where appropriate, portions of text from selected sources have been included in their entirety. A number of human factors engineering guidelines for equipment designers have been written over the past three decades, each tailored to the needs of the specific system being designed. In the case of this particular document, a review of the preliminary human operator functions involved in each phase of the retrieval process was performed, resulting in the identification of areas of design emphasis upon which this document should be based. Documents containing information and design data on each of these areas were acquired, and data and design guidelines related to the previously identified areas of emphasis were extracted and reorganized. For each system function, actions were first assigned to operator and/or machine, and the operator functions were then described. Separate lists of operator functions were developed for each of the areas of retrieval activities - survey and mapping, remining, floor flange emplacement, plug and canister overcoring, plug and canister removal and transport, and CWSRS activity. These functions and the associated man-machine interface were grouped into categories based on task similarity, and the principal topics of human factors design emphasis were extracted. These topic areas are reflected in the contents of the 12 sections of this document

  10. Deep repository for spent nuclear fuel. SR-97-Post-closure safety. Main Report. Volume I and II

    International Nuclear Information System (INIS)

    Hedin, A.

    1999-11-01

    In preparation for coming site investigations for siting of a deep repository for spent nuclear fuel, the Swedish Government and nuclear regulatory authorities have requested an assessment of the repository's long-term safety. The purpose is to demonstrate whether the risk of harmful effects in individuals in the vicinity of the repository complies with the acceptance criterion formulated by the Swedish regulatory authorities, i.e. that the risk may not exceed 10 -6 per year. Geological data are taken from three sites in Sweden to shed light on different conditions in Swedish granitic bedrock. The future evolution of the repository system is analyzed in the form of five scenarios. The first is a base scenario where the repository is postulated to be built entirely according to specifications and where present-day conditions in the surroundings are postulated to persist. The four other scenarios show how the evolution of the repository differs from that in the base scenario if the repository contains a few initially defective canisters, in the event of climate change, earthquakes, and future inadvertent human intrusion. The time horizon for the analyses is at most one million years, in accordance with preliminary regulations. By means of model studies and calculations, the base scenario analyzes how the radioactivity of the fuel declines with time, the repository's thermal evolution as a result of the decay heat in the fuel, the hydraulic evolution in buffer and backfill when they become saturated with water, and the long-term groundwater flow in the geosphere on the three sites. The overall conclusion of the analyses in the base scenario is that the copper canisters isolating capacity is not threatened by either the mechanical or chemical stresses to which it is subjected. The safety margins are great even in a million-year perspective. The internal evolution in initially defective canisters and the possible resultant migration of radionuclides in buffer, geosphere

  11. Deep repository for spent nuclear fuel. SR-97-Post-closure safety. Main Report. Volume I and II

    Energy Technology Data Exchange (ETDEWEB)

    Hedin, A [ed.

    1999-11-01

    In preparation for coming site investigations for siting of a deep repository for spent nuclear fuel, the Swedish Government and nuclear regulatory authorities have requested an assessment of the repository's long-term safety. The purpose is to demonstrate whether the risk of harmful effects in individuals in the vicinity of the repository complies with the acceptance criterion formulated by the Swedish regulatory authorities, i.e. that the risk may not exceed 10{sup -6} per year. Geological data are taken from three sites in Sweden to shed light on different conditions in Swedish granitic bedrock. The future evolution of the repository system is analyzed in the form of five scenarios. The first is a base scenario where the repository is postulated to be built entirely according to specifications and where present-day conditions in the surroundings are postulated to persist. The four other scenarios show how the evolution of the repository differs from that in the base scenario if the repository contains a few initially defective canisters, in the event of climate change, earthquakes, and future inadvertent human intrusion. The time horizon for the analyses is at most one million years, in accordance with preliminary regulations. By means of model studies and calculations, the base scenario analyzes how the radioactivity of the fuel declines with time, the repository's thermal evolution as a result of the decay heat in the fuel, the hydraulic evolution in buffer and backfill when they become saturated with water, and the long-term groundwater flow in the geosphere on the three sites. The overall conclusion of the analyses in the base scenario is that the copper canisters isolating capacity is not threatened by either the mechanical or chemical stresses to which it is subjected. The safety margins are great even in a million-year perspective. The internal evolution in initially defective canisters and the possible resultant migration of radionuclides in buffer

  12. Deep repository for spent nuclear fuel. SR-97-Post-closure safety. Main Report. Volume I and II

    Energy Technology Data Exchange (ETDEWEB)

    Hedin, A. [ed.

    1999-11-01

    In preparation for coming site investigations for siting of a deep repository for spent nuclear fuel, the Swedish Government and nuclear regulatory authorities have requested an assessment of the repository's long-term safety. The purpose is to demonstrate whether the risk of harmful effects in individuals in the vicinity of the repository complies with the acceptance criterion formulated by the Swedish regulatory authorities, i.e. that the risk may not exceed 10{sup -6} per year. Geological data are taken from three sites in Sweden to shed light on different conditions in Swedish granitic bedrock. The future evolution of the repository system is analyzed in the form of five scenarios. The first is a base scenario where the repository is postulated to be built entirely according to specifications and where present-day conditions in the surroundings are postulated to persist. The four other scenarios show how the evolution of the repository differs from that in the base scenario if the repository contains a few initially defective canisters, in the event of climate change, earthquakes, and future inadvertent human intrusion. The time horizon for the analyses is at most one million years, in accordance with preliminary regulations. By means of model studies and calculations, the base scenario analyzes how the radioactivity of the fuel declines with time, the repository's thermal evolution as a result of the decay heat in the fuel, the hydraulic evolution in buffer and backfill when they become saturated with water, and the long-term groundwater flow in the geosphere on the three sites. The overall conclusion of the analyses in the base scenario is that the copper canisters isolating capacity is not threatened by either the mechanical or chemical stresses to which it is subjected. The safety margins are great even in a million-year perspective. The internal evolution in initially defective canisters and the possible resultant migration of radionuclides in

  13. Development of a standard data base for FBR core nuclear design. 10. Reevaluation of atomic number density of JOYO Mk-II core

    Energy Technology Data Exchange (ETDEWEB)

    Numata, Kazuyuki; Sato, Wakaei [Japan Nuclear Cycle Development Inst., Oarai, Ibaraki (Japan). Oarai Engineering Center; Ishikawa, Makoto; Arii, Yoshio [Nuclear Energy System Incorporation, Tokyo (Japan)

    1999-07-01

    The material composition of JOYO Mk-II core components in its initial core was reevaluated as a part of the effort for developing a standard data base for FBR core nuclear design. The special feature of the reevaluation is to treat the decay of Pu-241 isotope, so that the atomic number densities of Pu-241 and Am-241 in fuel assemblies can be exactly evaluated on the initial critical date, Nov. 22nd, 1982. Further, the atomic number densities of other core components were also evaluated to improve the analytical accuracy. Those include the control rods which were not so strictly evaluated in the past, and the dummy fuels and the neutron sources which were not treated in the analytical model so far. The results of the present reevaluation were as follows: (1) The changes of atomic number densities of the major nuclides such as Pu-239, U-235 and U-238 were about {+-}0.2 to 0.3%. On the other hand, the number density of Pu-241, which was the motivation of the present work, was reduced by 12%. From the fact, the number densities in the past analysis might be based on the isotope measurement of the manufacturing point of time without considering the decay of Pu-241. (2) As the other core components, the number densities of control rods and outer reflector-type A were largely improved. (author)

  14. cobalt (ii), nickel (ii)

    African Journals Online (AJOL)

    DR. AMINU

    Department of Chemistry Bayero University, P. M. B. 3011, Kano, Nigeria. E-mail: hnuhu2000@yahoo.com. ABSTRACT. The manganese (II), cobalt (II), nickel (II) and .... water and common organic solvents, but are readily soluble in acetone. The molar conductance measurement [Table 3] of the complex compounds in.

  15. Thermal phenomenae in nuclear fuel rods

    International Nuclear Information System (INIS)

    Baigorria, Carlos.

    1983-12-01

    Thermal phenomenae occurring in a nuclear fuel rod under irradiation are studied. The most important parameters of either steady or transient thermal states are determined. The validity of applying the Fourier's approximation equations to these problems is also studied. A computer program TRANS is developed in order to study the transient cases. This program solves a system of coupled, non-linear partial differential equations, of parabolic type, in cylindrical coordinates with various boundary conditions. The benchmarking of the TRANS program is done by comparing its predictions with the analytical solution of some simplified transient cases. Complex transient cases such as those corresponding to characteristic reactor accidents are studied, in particular for typical pressurized heavy water reactor (PHWR) fuel rods, such as those of Atucha I. The Stefan problem emerging in the case of melting of the fuel element is solved. Qualitative differences between the classical Stefan problem, without inner sources, and that one, which includes sources are discussed. The MSA program, for solving the Stefan problem with inner sources is presented; and furthermore, it serves to predict thermal evolution, when the fuel element melts. Finally a model for fuel phase change under irradiation is developed. The model is based on the dimensional invariants of the percolation theory when applied to the connectivity of liquid spires nucleated around each fission fragment track. Suggestions for future research into the subject are also presented. (autor) [es

  16. Decommissioning and deactivation of nuclear facilities

    International Nuclear Information System (INIS)

    Anasco, Roberto; Harriague, Santiago; Hey, Alfredo M.; Fabbri, Silvio; Garonis, Omar H.

    2003-01-01

    The National Atomic Energy Commission (CNEA) is responsible for the decommissioning and deactivation of all relevant nuclear facilities in Argentina. A D and D Subprogram was created in 2000, within Technology Branch of the CNEA, in order to fulfill this responsibility. The D and D Subprogram has organized its activities in four fields: Planning; Technology development; Human resources development and training; International cooperation. The paper describes the work already done in those 4 areas, as well as the nuclear facilities existing in the country. Planning is being developed for the decommissioning of research reactors, beginning with RA-1, as well as for the Atucha I nuclear power station. An integral Management System has been developed, compatibilizing requirements from ISO 9001, ISO 14001, the national norm for Safety and Occupational Health (equivalent to BS 8800), and IAEA 50-SG Q series. Technology development is for the time being concentrated on mechanical decontamination and concrete demolition. A review has been made of technologies already developed both by CNEA and Nucleoelectrica Argentina S.A. (the nuclear power utility) in areas of chemical and electrochemical decontamination, cutting techniques and robotics. Human resources development has been based on training abroad in the areas of decontamination, cutting techniques, quality assurance and planning, as well as on specific courses, seminars and workshops. An IAEA regional training course on D and D has been given on April 2002 at CNEA's Constituyentes Atomic Center, with the assistance of 22 university graduates from 13 countries in the Latin American and Caribbean Region, and 11 from Argentina. CNEA has also given fellowships for PhD and Master thesis on the subject. International cooperation has been intense, and based on: - IAEA Technical Cooperation Project and experts missions; - Cooperation agreement with the US Department of Energy; - Cooperation agreement with Germany

  17. Ligand design for alkali-metal-templated self-assembly of unique high-nuclearity CuII aggregates with diverse coordination cage units: crystal structures and properties.

    Science.gov (United States)

    Du, Miao; Bu, Xian-He; Guo, Ya-Mei; Ribas, Joan

    2004-03-19

    The construction of two unique, high-nuclearity Cu(II) supramolecular aggregates with tetrahedral or octahedral cage units, [(mu(3)-Cl)[Li subset Cu(4)(mu-L(1))(3)](3)](ClO(4))(8)(H(2)O)(4.5) (1) and [[Na(2) subset Cu(12)(mu-L(2))(8)(mu-Cl)(4)](ClO(4))(8)(H(2)O)(10)(H(3)O(+))(2)](infinity) (2) by alkali-metal-templated (Li(+) or Na(+)) self-assembly, was achieved by the use of two newly designed carboxylic-functionalized diazamesocyclic ligands, N,N'-bis(3-propionyloxy)-1,4-diazacycloheptane (H(2)L(1)) or 1,5-diazacyclooctane-N,N'-diacetate acid (H(2)L(2)). Complex 1 crystallizes in the trigonal R3c space group (a = b = 20.866(3), c = 126.26(4) A and Z = 12), and 2 in the triclinic P1 space group (a = 13.632(4), b = 14.754(4), c = 19.517(6) A, alpha = 99.836(6), beta = 95.793(5), gamma = 116.124(5) degrees and Z = 1). By subtle variation of the ligand structures and the alkali-metal templates, different polymeric motifs were obtained: a dodecanuclear architecture 1 consisting of three Cu(4) tetrahedral cage units with a Li(+) template, and a supramolecular chain 2 consisting of two crystallographically nonequivalent octahedral Cu(6) polyhedra with a Na(+) template. The effects of ligand functionality and alkali metal template ions on the self-assembly processes of both coordination supramolecular aggregates, and their magnetic behaviors are discussed in detail.

  18. Applications of nuclear physics

    Science.gov (United States)

    Hayes, A. C.

    2017-02-01

    Today the applications of nuclear physics span a very broad range of topics and fields. This review discusses a number of aspects of these applications, including selected topics and concepts in nuclear reactor physics, nuclear fusion, nuclear non-proliferation, nuclear-geophysics, and nuclear medicine. The review begins with a historic summary of the early years in applied nuclear physics, with an emphasis on the huge developments that took place around the time of World War II, and that underlie the physics involved in designs of nuclear explosions, controlled nuclear energy, and nuclear fusion. The review then moves to focus on modern applications of these concepts, including the basic concepts and diagnostics developed for the forensics of nuclear explosions, the nuclear diagnostics at the National Ignition Facility, nuclear reactor safeguards, and the detection of nuclear material production and trafficking. The review also summarizes recent developments in nuclear geophysics and nuclear medicine. The nuclear geophysics areas discussed include geo-chronology, nuclear logging for industry, the Oklo reactor, and geo-neutrinos. The section on nuclear medicine summarizes the critical advances in nuclear imaging, including PET and SPECT imaging, targeted radionuclide therapy, and the nuclear physics of medical isotope production. Each subfield discussed requires a review article unto itself, which is not the intention of the current review; rather, the current review is intended for readers who wish to get a broad understanding of applied nuclear physics.

  19. Applications of nuclear physics

    International Nuclear Information System (INIS)

    Hayes-Sterbenz, Anna Catherine

    2017-01-01

    Today the applications of nuclear physics span a very broad range of topics and fields. This review discusses a number of aspects of these applications, including selected topics and concepts in nuclear reactor physics, nuclear fusion, nuclear non-proliferation, nuclear-geophysics, and nuclear medicine. The review begins with a historic summary of the early years in applied nuclear physics, with an emphasis on the huge developments that took place around the time of World War II, and that underlie the physics involved in designs of nuclear explosions, controlled nuclear energy, and nuclear fusion. The review then moves to focus on modern applications of these concepts, including the basic concepts and diagnostics developed for the forensics of nuclear explosions, the nuclear diagnostics at the National Ignition Facility, nuclear reactor safeguards, and the detection of nuclear material production and trafficking. The review also summarizes recent developments in nuclear geophysics and nuclear medicine. The nuclear geophysics areas discussed include geo-chronology, nuclear logging for industry, the Oklo reactor, and geo-neutrinos. The section on nuclear medicine summarizes the critical advances in nuclear imaging, including PET and SPECT imaging, targeted radionuclide therapy, and the nuclear physics of medical isotope production. Lastly, each subfield discussed requires a review article unto itself, which is not the intention of the current review; rather, the current review is intended for readers who wish to get a broad understanding of applied nuclear physics.

  20. 4. International Conference on Current Problems in Nuclear Physics and Atomic Energy (NPAE-Kyiv2012). Proceedings. Part I and Part II

    International Nuclear Information System (INIS)

    Vyshnevskyi, Ivan M.

    2012-01-01

    Such wide area of topics, discussed during the Conference, is closely connected with the interests of our country to develop the fundamental research in the field of nuclear physics, which is the base of nuclear energy. The purpose of the Conference was to bring together scientists to share their knowledge in the current problems in nuclear physics and atomic energy. consideration of the spherical ground-state proton emitters, while nuclear deformations are supposed to be further included by standard way

  1. Structure for Transparency in Nuclear Waste Management. Comparative Review of the Structures for Nuclear Waste Management in France, Sweden and the UK. A Report from the RISCOM II Project

    Energy Technology Data Exchange (ETDEWEB)

    Espejo, Raul [Syncho Ltd., Lincoln (United Kingdom)

    2002-11-01

    This report presents a comparison of the structures for nuclear waste management in France, Sweden and the UK. The source materials for this comparison are studies carried out in each of these countries by Syncho Ltd. over the past 5 years. The Swedish structural review was sponsored by SKI and SSI, and carried out as a pilot study during the years 1996 and 1997 as part of the RISCOM Pilot Project. The structural reviews of the British and French nuclear waste management systems have been in progress for the past two years (2001-2002) within the framework of RISCOM II, sponsored by the European Union. This report offers preliminary comparative views of the three systems. As with each of the individual studies more work and information are necessary to confirm and strengthen the findings. To set the context for this report it is important to remind the reader that the study in Sweden was undertaken 5 years ago, that the French case took place at the same time of significant structural changes in the country's nuclear waste management system and that the British case was undertaken at the same time of a far-reaching Government consultation process. In all cases the number of people interviewed was small. In summary, comparing the structures for transparency suggests that once existing channels for transparency are diagnosed, it should be possible to use benchmarks of good practice in one country to design methods to improve participation and communications in others. The framework used in this report allows making comparisons beyond factual reports of similarities or differences. An important conclusion of this report is that the democratic deficits that we experience today as citizens in all societies can be ameliorated if sufficient attention is paid to producing requisite organisations, with adequate communications, capable of bridging the gaps between the silent majorities and those experts and politicians responsible for policy decisions. It is the wisdom

  2. Structure for Transparency in Nuclear Waste Management. Comparative Review of the Structures for Nuclear Waste Management in France, Sweden and the UK. A Report from the RISCOM II Project

    International Nuclear Information System (INIS)

    Espejo, Raul

    2002-11-01

    This report presents a comparison of the structures for nuclear waste management in France, Sweden and the UK. The source materials for this comparison are studies carried out in each of these countries by Syncho Ltd. over the past 5 years. The Swedish structural review was sponsored by SKI and SSI, and carried out as a pilot study during the years 1996 and 1997 as part of the RISCOM Pilot Project. The structural reviews of the British and French nuclear waste management systems have been in progress for the past two years (2001-2002) within the framework of RISCOM II, sponsored by the European Union. This report offers preliminary comparative views of the three systems. As with each of the individual studies more work and information are necessary to confirm and strengthen the findings. To set the context for this report it is important to remind the reader that the study in Sweden was undertaken 5 years ago, that the French case took place at the same time of significant structural changes in the country's nuclear waste management system and that the British case was undertaken at the same time of a far-reaching Government consultation process. In all cases the number of people interviewed was small. In summary, comparing the structures for transparency suggests that once existing channels for transparency are diagnosed, it should be possible to use benchmarks of good practice in one country to design methods to improve participation and communications in others. The framework used in this report allows making comparisons beyond factual reports of similarities or differences. An important conclusion of this report is that the democratic deficits that we experience today as citizens in all societies can be ameliorated if sufficient attention is paid to producing requisite organisations, with adequate communications, capable of bridging the gaps between the silent majorities and those experts and politicians responsible for policy decisions. It is the wisdom of the

  3. Application of the neutron noise analysis technique in nuclear power plants

    International Nuclear Information System (INIS)

    Lescano, Victor H.; Wentzeis, Luis M.

    1999-01-01

    Using the neutron noise analysis in nuclear power plants, and without producing any perturbation in the normal operation of the plant, information of the vibration state of the reactor internals and the behavior of the operating conditions of the reactor primary circuit can be obtained. In Argentina, the neutron noise analysis technique is applied in customary way in the nuclear power plants Atucha I and Embalse. A database was constructed and vibration frequencies corresponding to different reactor internals were characterized. Reactor internals with particular mechanical vibrations have been detected and localized. In the framing of a cooperation project between Argentina and Germany, we participated in the measurements, analysis and modelisation, using the neutron noise technique, in the Obrigheim and Gundremmingen nuclear power plants. In the nuclear power plant Obrigheim (PWR, 350 M We), correlations between the signals measured from self-power neutron detectors and accelerometers located inside the reactor core, were made. In the nuclear power plant Gundremmingen (BWR, 1200 M We) we participated in the study of a particular mechanical vibration detected in one of the instrumentation tube. (author)

  4. Decommissioning and deactivation of nuclear facilities; Desmantelamiento y clausura de instalaciones nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Anasco, Roberto; Harriague, Santiago; Hey, Alfredo M [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Gerencia de Tecnologia y Medio Ambiente; Fabbri, Silvio [Comision Nacional de Energia Atomica, General San Martin (Argentina). Centro Atomico Constituyentes; Garonis, Omar H [Comision Nacional de Energia Atomica, General San Martin (Argentina). Dept. de Gestion de Calidad

    2003-07-01

    The National Atomic Energy Commission (CNEA) is responsible for the decommissioning and deactivation of all relevant nuclear facilities in Argentina. A D and D Subprogram was created in 2000, within Technology Branch of the CNEA, in order to fulfill this responsibility. The D and D Subprogram has organized its activities in four fields: Planning; Technology development; Human resources development and training; International cooperation. The paper describes the work already done in those 4 areas, as well as the nuclear facilities existing in the country. Planning is being developed for the decommissioning of research reactors, beginning with RA-1, as well as for the Atucha I nuclear power station. An integral Management System has been developed, compatibilizing requirements from ISO 9001, ISO 14001, the national norm for Safety and Occupational Health (equivalent to BS 8800), and IAEA 50-SG Q series. Technology development is for the time being concentrated on mechanical decontamination and concrete demolition. A review has been made of technologies already developed both by CNEA and Nucleoelectrica Argentina S.A. (the nuclear power utility) in areas of chemical and electrochemical decontamination, cutting techniques and robotics. Human resources development has been based on training abroad in the areas of decontamination, cutting techniques, quality assurance and planning, as well as on specific courses, seminars and workshops. An IAEA regional training course on D and D has been given on April 2002 at CNEA's Constituyentes Atomic Center, with the assistance of 22 university graduates from 13 countries in the Latin American and Caribbean Region, and 11 from Argentina. CNEA has also given fellowships for PhD and Master thesis on the subject. International cooperation has been intense, and based on: - IAEA Technical Cooperation Project and experts missions; - Cooperation agreement with the US Department of Energy; - Cooperation agreement with Germany

  5. Experience in construction and operation of HWR plants in Argentina

    International Nuclear Information System (INIS)

    Madero, C.C.; Cosentino, J.O.

    1982-01-01

    ''ATUCHA I'', the first nuclear power plant in Argentina, is in commerical operation since 1974 with a high capacity factor. The reactor is based on the MZFR prototype designed by SIEMENS with natural uranium and heavy water and PWR technic. The plant was built by SIEMENS on a turnkey contract and was rated 340 MWe. The offer presented in that opportunity by KWU was based on two reactors (ATUCHA I type) inside one single containment, due to the limitation in power of the reactor. Subsequent changes in the nature of the contract resulted in an active participation of CNEA engineering groups in the erection and commissioning of the reactor. In 1978 the national government approved a nuclear power plan to install four 600 MWe HWR plants until 1995. To start implementing this program, CNEA called for tenders for the supply of components and services for the ATUCHA II plant, in connection with the establishment of a local engineering company and the supply and construction of a heavy water production plant. In 1980 a contract was signed with KWU and the local company ENACE was formed to act as architect engineer and site coordinator. The plant will be located in the ATUCHA I site and the reactor will be similar but double in power to that one. Following the schedule of the nuclear plan, CNEA has just started preliminary studies for the next nuclear plant. ENACE will be responsible for the preparation of an offer for an ATUCHA reactor type. Local engineering and manufacturing firms, upon request and coordination from CNEA, and evaluating the local capacity to participate in the design and construction of a CANDU type nuclear plant. Final decision on this fourth nuclear plant in Argentina will be taken middle 1983. (J.P.N.)

  6. Assessment and management of ageing of major nuclear power plant components important to safety: In-containment instrumentation and control cables. Volume II

    International Nuclear Information System (INIS)

    2000-12-01

    At present, there are over four hundred operational nuclear power plants (NPPs) in IAEA Member States. Operating experience has shown that ineffective control of the ageing degradation of major NPP components (e.g. caused by unanticipated phenomena and by operating, maintenance, design or manufacturing error) can jeopardize plant safety and also plant life. Ageing in these NPPs must be therefore effectively managed to ensure the availability of design functions throughout the plant service life. From the safety perspective, this means controlling within acceptable limits the ageing degradation and wear-out of plant components important to safety so that adequate safety margins remain, i.e. integrity and functional capability in excess of normal operating requirements. This publication is one in a series of guidance reports on the assessment and management of ageing of the major NPP components important to safety. The reports are based on experience and practices of NPP operators, regulators, designers, manufacturers, and technical support organizations and a widely accepted Methodology for the Management of Ageing of NPP Components Important to Safety, which was issued by the IAEA in 1992. The current practices for the assessment of safety margins (fitness-for-service) and the inspection, monitoring and mitigation of ageing degradation of selected components of Canadian deuterium-uranium (CANDU) reactors, boiling water reactors (BWRs), pressurized water reactors (PWRs), including the Soviet designed 'water moderated and water cooled energy reactors' (WWERs), are documented in the reports. These practices are intended to help all involved directly and indirectly in ensuring the safe operation of NPPs, and to provide a common technical basis for dialogue between plant operators and regulators when dealing with age related licensing issues. The guidance reports are directed at technical experts and managers from NPPs and from regulatory, plant design, manufacturing

  7. Computational analysis of neutronic parameters for TRIGA Mark-II research reactor using evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3

    International Nuclear Information System (INIS)

    Altaf, M.H.; Badrun, N.H.; Chowdhury, M.T.

    2015-01-01

    Highlights: • SRAC-PIJ code and SRAC-CITATION have been utilized to model the core. • Most of the simulated results show no significant differences with references. • Thermal peak flux varies a bit due to up condition of TRIGA. • ENDF/B-VII.0 and JENDL-3.3 libraries perform well for neutronics analysis of TRIGA. - Abstract: Important kinetic parameters such as effective multiplication factor, k eff , excess reactivity, neutron flux and power distribution, and power peaking factors of TRIGA Mark II research reactor in Bangladesh have been calculated using the comprehensive neutronics calculation code system SRAC 2006 with the evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3. In the code system, PIJ code was employed to obtain cross section of the core cells, followed by the integral calculation of neutronic parameters of the reactor conducted by CITATION code. All the analyses were performed using the 7-group macroscopic cross section library. Results were compared to the experimental data, the safety analysis report (SAR) of the reactor provided by General Atomic as well as to the simulated values by numerically benchmarked MCNP4C, WIMS-CITATION and SRAC-CITATION codes. The maximum power densities at the hot spot were found to be 169.7 W/cc and 170.1 W/cc for data libraries ENDF/B-VII.0 and JENDL-3.3, respectively. Similarly, the total peaking factors based on ENDF/B-VII.0 and JENDL-3.3 were calculated as 5.68 and 5.70, respectively, which were compared to the original SAR value of 5.63, as well as to MCNP4C, WIMS-CITATION and SRAC-CITATION results. It was found in most cases that the calculated results demonstrate a good agreement with our experiments and published works. Therefore, this analysis benchmarks the code system and will be helpful to enhance further neutronics and thermal hydraulics study of the reactor

  8. Proceedings of the Sixth Arab Conference on the Peaceful Uses of Atomic Energy, Vol.II. Scientific Presentation (Reactors, Materials, Fuel Cycles and Nuclear Safety)

    International Nuclear Information System (INIS)

    2003-10-01

    The publication has been set up as a textbook for researching dealing with health protection during work with Human needs of Nuclear Science and applications. The book consists of the following chapters: Personnel and working environment monitoring; analytical techniques; radiation protection harmonized and integrated policy for the arab country; Nuclear safety; fuel cycles; nuclear medicine; accelerators; medical applications; radiation chemistry; hydrology; environmental studies; biological effects of ionizing radiation on agriculture; radiation accidents

  9. Proceedings of the Sixth Arab Conference on the Peaceful Uses of Atomic Energy, Vol.II. Scientific Presentation (Reactors, Materials, Fuel Cycles and Nuclear Safety)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-10-01

    The publication has been set up as a textbook for researching dealing with health protection during work with Human needs of Nuclear Science and applications. The book consists of the following chapters: Personnel and working environment monitoring; analytical techniques; radiation protection harmonized and integrated policy for the arab country; Nuclear safety; fuel cycles; nuclear medicine; accelerators; medical applications; radiation chemistry; hydrology; environmental studies; biological effects of ionizing radiation on agriculture; radiation accidents.

  10. cobalt (ii), nickel (ii)

    African Journals Online (AJOL)

    DR. AMINU

    ... and H. J. Abdullahi. Department of Chemistry Bayero University, P. M. B. 3011, Kano, Nigeria ... The condensation reaction of primary amines and the compounds containing a ..... Inorganic, Nuclear Chemistry, 40: 1179-. 1180. Nakomato, K.

  11. Nsls-II Boster

    Science.gov (United States)

    Gurov, S. M.; Akimov, A. V.; Akimov, V. E.; Anashin, V. V.; Anchugov, O. V.; Baranov, G. N.; Batrakov, A. M.; Belikov, O. V.; Bekhtenev, E. A.; Blum, E.; Bulatov, A. V.; Burenkov, D. B.; Cheblakov, P. B.; Chernyakin, A. D.; Cheskidov, V. G.; Churkin, I. N.; Davidsavier, M.; Derbenev, A. A.; Erokhin, A. I.; Fliller, R. P.; Fulkerson, M.; Gorchakov, K. M.; Ganetis, G.; Gao, F.; Gurov, D. S.; Hseuh, H.; Hu, Y.; Johanson, M.; Kadyrov, R. A.; Karnaev, S. E.; Karpov, G. V.; Kiselev, V. A.; Kobets, V. V.; Konstantinov, V. M.; Kolmogorov, V. V.; Korepanov, A. A.; Kramer, S.; Krasnov, A. A.; Kremnev, A. A.; Kuper, E. A.; Kuzminykh, V. S.; Levichev, E. B.; Li, Y.; Long, J. De; Makeev, A. V.; Mamkin, V. R.; Medvedko, A. S.; Meshkov, O. I.; Nefedov, N. B.; Neyfeld, V. V.; Okunev, I. N.; Ozaki, S.; Padrazo, D.; Petrov, V. V.; Petrichenkov, M. V.; Philipchenko, A. V.; Polyansky, A. V.; Pureskin, D. N.; Rakhimov, A. R.; Rose, J.; Ruvinskiy, S. I.; Rybitskaya, T. V.; Sazonov, N. V.; Schegolev, L. M.; Semenov, A. M.; Semenov, E. P.; Senkov, D. V.; Serdakov, L. E.; Serednyakov, S. S.; Shaftan, T. V.; Sharma, S.; Shichkov, D. S.; Shiyankov, S. V.; Shvedov, D. A.; Simonov, E. A.; Singh, O.; Sinyatkin, S. V.; Smaluk, V. V.; Sukhanov, A. V.; Tian, Y.; Tsukanova, L. A.; Vakhrushev, R. V.; Vobly, P. D.; Utkin, A. V.; Wang, G.; Wahl, W.; Willeke, F.; Yaminov, K. R.; Yong, H.; Zhuravlev, A.; Zuhoski, P.

    The National Synchrotron Light Source II is a third generation light source, which was constructed at Brookhaven National Laboratory. This project includes a highly-optimized 3 GeV electron storage ring, linac preinjector, and full-energy synchrotron injector. Budker Institute of Nuclear Physics built and delivered the booster for NSLS-II. The commissioning of the booster was successfully completed. This paper reviews fulfilled work by participants.

  12. Ferromagnetic dinuclear mixed-valence Mn(II)/Mn(III) complexes: building blocks for the higher nuclearity complexes. structure, magnetic properties, and density functional theory calculations.

    Science.gov (United States)

    Hänninen, Mikko M; Välivaara, Juha; Mota, Antonio J; Colacio, Enrique; Lloret, Francesc; Sillanpää, Reijo

    2013-02-18

    A series of six mixed-valence Mn(II)/Mn(III) dinuclear complexes were synthesized and characterized by X-ray diffraction. The reactivity of the complexes was surveyed, and structures of three additional trinuclear mixed-valence Mn(III)/Mn(II)/Mn(III) species were resolved. The magnetic properties of the complexes were studied in detail both experimentally and theoretically. All dinuclear complexes show ferromagnetic intramolecular interactions, which were justified on the basis of the electronic structures of the Mn(II) and Mn(III) ions. The large Mn(II)-O-Mn(III) bond angle and small distortion of the Mn(II) cation from the ideal square pyramidal geometry were shown to enhance the ferromagnetic interactions since these geometrical conditions seem to favor the orthogonal arrangement of the magnetic orbitals.

  13. Federal Republic of Germany: Prospects for nuclear energy from 1972-1992

    Energy Technology Data Exchange (ETDEWEB)

    Hilger-Haunschild, H [Federal Ministry for Education and Science, Bonn (Germany)

    1972-07-01

    The number of nuclear power stations, both built and planned, in the Federal Republic of Germany, bears witness to what has been achieved so far. At present, nuclear power stations in operation generate a total of about 2000 MWe, while power stations with a total capacity for a further 10 000 MWe are under construction. The first export orders demonstrate the competitiveness of the German nuclear power industry -power stations are now being built by West German firms at Atucha in Argentina, Borselle in the Netherlands, and Zwentendorf in Austria. Because of parallel technological advances reached by the world's major industrialized nations, and the large funds necessary for further nuclear development, international cooperation is increasingly important. The federal Government therefore follows a policy of joint-development projects, particularly within a European framework. The SNR 300 fast breeder reactor, which is to be constructed with Belgium and the Netherlands, and the development of the gas centrifuge technique being carried out with the Netherlands and the United Kingdom, are excellent examples of this policy.

  14. Relationship between IEEE Std. 7-4.3.2-1993 and ASME NQA-1, parts I and II revisions and the impact on nuclear power generating stations

    International Nuclear Information System (INIS)

    Blauw, R.J.

    1996-01-01

    Clear understanding of software related design control requirements is key to growth in the use of computers in nuclear power generating stations. Inconsistent terminology within the nuclear and software standards arena has impacted the ability of both nuclear station system engineers (i.e., the domain expert) to clearly communicate with the software/computer hardware experts. In order for computer development to occur both groups need to have a common terminology basis. Without this commonality, inappropriate application of requirements could result. This paper will present a overview of ongoing efforts within the Institute of Electrical and Electronics Engineers Nuclear Power Engineering Committee (IEEE NPEC) and the American Society of Mechanical Engineers Nuclear Quality Assurance (ASME NQA) Committee to develop this commonality

  15. A world's dilemma 'upon which the sun never sets'. The nuclear waste management strategy. Western European nation states and the United States of America. Pt. II

    International Nuclear Information System (INIS)

    Sanders, Mark Callis; Sanders, Charlotta E.

    2016-01-01

    The management of spent nuclear fuel (SNF) and nuclear wastes demands a strategy to provide for the safe, secure, and permanent disposal of radioactive material from power generation, defense uses, and other activities. Nation states have taken different paths to nuclear waste management and are at various stages of the development of a nuclear waste management strategy. A strategy may include developing a geological repository, nuclear fuel reprocessing, interim storage, as well as discussions of the creation of a multinational storage facility. The paper provides an overview of the strategy used (or being developed) and its place within the legal framework. The paper concludes that though each nation state must look outward to its shared international obligations, there must also be an inward reflection of a nation state to its own traditions, customs, and legal/law making regimes.

  16. Revision of documents guide to obtain the renovation of licence of the nuclear power plant of Laguna Verde, Units I and II

    International Nuclear Information System (INIS)

    Jarvio C, G.; Fernandez S, G.

    2008-01-01

    Unquestionably the renovation of license of the nuclear power plants , it this converting in a promising option to be able to gather in sure, reliable form and economic those requests about future energy in the countries with facilities of this type. This work it analyzes four documents guide and their application for the determination of the renovation of it licenses in nuclear plants that will serve of base for their aplication in the Nuclear Power Plant of Laguna Verde Units I and 2. The four documents in question are: the one Inform Generic of Learned Lessons on the Aging (NUREG - 1801), the Standard Revision Plan for the Renovation of License (NUREG - 1800), the Regulatory Guide for Renovation of License (GR-1.188), and the NEI 95-10, developed by the Institute of Nuclear Energy that is an Industrial Guide to Implement the Requirements of the 1OCFR Part 54-the Rule of Renovation of It licenses. (Author)

  17. Proceeding on the scientific meeting and presentation on basic research of nuclear science and technology (book II): chemical, waste processing technology and environment

    International Nuclear Information System (INIS)

    Prayitno; Syarip; Samin; Darsono; Agus Taftazani; Sudjatmoko; Tri Mardji Atmono; Dwi Biyantoro; Gede Sutresna W; Tjipto Sujitno; Slamet Santosa; Herry Poernomo; Bambang Siswanto; Eko Edy Karmanto; Endro Kismolo; Budi Setiawan; Prajitno; Jumari; Wahini Nurhayati

    2015-06-01

    Scientific Meeting and Presentation on Basic Research in Nuclear Science and Technology is an annual activity held by Centre for Accelerator Science and Technology, National Nuclear Energy Agency, in Yogyakarta, for monitoring research activities achieved by the Agency. The papers presented in the meeting were collected into proceedings which were divided into two groups that are chemistry, environmental and waste treatment technology process . The proceedings consists of three articles from keynote speakers and 24 articles from BATAN and others participants.(PPIKSN)

  18. Experiment CATETO II

    International Nuclear Information System (INIS)

    Hendriks, J.A.; Freudenreich, W.E.

    1994-03-01

    In the irradiation experiment CATETO II different reduced activation (RA) steels will be irradiated up to 2.5 dpa at a temperature of 300 C. The results of the calculation of the nuclear constants, the reactivity effect, and the activity of the steel samples are presented. (orig.)

  19. Comparison between a finite difference model (PUMA) and a finite element model (DELFIN) for simulation of the reactor of the atomic power plant of Atucha I

    International Nuclear Information System (INIS)

    Grant, C.R.

    1996-01-01

    The reactor code PUMA, developed in CNEA, simulates nuclear reactors discretizing space in finite difference elements. Core representation is performed by means a cylindrical mesh, but the reactor channels are arranged in an hexagonal lattice. That is why a mapping using volume intersections must be used. This spatial treatment is the reason of an overestimation of the control rod reactivity values, which must be adjusted modifying the incremental cross sections. Also, a not very good treatment of the continuity conditions between core and reflector leads to an overestimation of channel power of the peripherical fuel elements between 5 to 8 per cent. Another code, DELFIN, developed also in CNEA, treats the spatial discretization using heterogeneous finite elements, allowing a correct treatment of the continuity of fluxes and current among elements and a more realistic representation of the hexagonal lattice of the reactor. A comparison between results obtained using both methods in done in this paper. (author). 4 refs., 3 figs

  20. Mission Plan for the Civilian Radioactive Waste Management Program. Volume I. Part I. Overview and current program plans; Part II. Information required by the Nuclear Waste Policy Act of 1982

    International Nuclear Information System (INIS)

    1985-06-01

    The Misson Plan is divided into two parts. Part I describes the overall goals, objectives, and strategy for the disposal of spent nuclear fuel and high-level waste. It explains that, to meet the directives of the Nuclear Waste Policy Act, the DOE intends to site, design, construct, and start operating a mined geologic repository by January 31, 1998. The Act specifies that the costs of these activities will be borne by the owners and generators of the waste received at the repository. Part I further describes the other components of the waste-management program - monitored retrievable storage, Federal interim storage, and transportation - as well as systems integration activities. Also discussed are institutional plans and activities as well as the program-management system being implemented by the Office of Civilian Radioactive Waste Management. Part II of the Mission Plan presents the detailed information required by Section 301(a) of the Act - key issues and information needs; plans for obtaining the necessary information; potential financial, institutional, and legal issues; plans for the test and evaluation facility; the principal results obtained to date from site investigations; information on the site-characterization programs; information on the waste package; schedules; costs; and socioeconomic impacts. In accordance with Section 301(a) of the Act, Part II is concerned primarily with the repository program

  1. Presentation of safety after closure of the repository for spent nuclear fuel. Main report of the project SR-Site. Part II; Redovisning av saekerhet efter foerslutning av slutfoervaret foer anvaent kaernbraensle. Huvudrapport fraan projekt SR-Site. Del II

    Energy Technology Data Exchange (ETDEWEB)

    2011-07-01

    The purpose of the safety assessment SR-Site is to investigate whether a safe repository for spent nuclear fuel by KBS-3 type can be constructed at Forsmark in Oesthammar in Sweden. The location of the Forsmark has been selected based on results of several surveys from surface conditions at depth in Forsmark and in Laxemar in Oskarshamn. The choice of location is not justified in SR-Site Report, but in other attachments to SKB's permit applications. SR-Site Report is an important part of SKB's permit applications to construct and operate a repository for spent nuclear fuel at Forsmark in Oesthammar. The purpose of the report in the applications is to show that a repository at Forsmark is safe after closure

  2. Copper (II)

    African Journals Online (AJOL)

    CLEMENT O BEWAJI

    Valine (2 - amino - 3 – methylbutanoic acid), is a chemical compound containing .... Stability constant (Kf). Gibb's free energy. ) (. 1. −. ∆. Mol. JG. [CuL2(H2O)2] ... synthesis and characterization of Co(ii), Ni(ii), Cu (II), and Zn(ii) complexes with ...

  3. DECOVALEX - Mathematical models of coupled T-H-M processes for nuclear waste repositories. Executive summary for Phases I,II and III

    International Nuclear Information System (INIS)

    Jing, L.; Stephansson, O.; Tsang, C.F.; Kautsky, F.

    1996-06-01

    This executive summary presents the motivation, structure, objectives, methodologies and results of the first stage of the international DECOVALEX project - DECOVALEX I (1992-1995). The acronym stands for Development of Coupled Models and their Validation against Experiment in Nuclear Waste Isolation, and the project is an international effort to develop mathematical models, numerical methods and computer codes for coupled thermo-hydro-mechanical processes in fractured rocks and buffer materials for geological isolation of spent nuclear fuel and other radioactive wastes, and validate them against laboratory and field experiments. 24 refs

  4. Cyclam Derivatives with a Bis(phosphinate) or a Phosphinato-Phosphonate Pendant Arm: Ligands for Fast and Efficient Copper(II) Complexation for Nuclear Medical Applications

    Czech Academy of Sciences Publication Activity Database

    David, T.; Kubíček, V.; Gutten, Ondrej; Lubal, P.; Kotek, J.; Pietzsch, H.-J.; Rulíšek, Lubomír; Hermann, P.

    2015-01-01

    Roč. 54, č. 24 (2015), s. 11751-11766 ISSN 0020-1669 R&D Projects: GA ČR(CZ) GA14-31419S Grant - others:COST(XE) TD1004 Institutional support: RVO:61388963 Keywords : cyclam derivatives * radiolabelling * quantum chemical calculations * copper(II) chelation Subject RIV: CA - Inorganic Chemistry Impact factor: 4.820, year: 2015

  5. Environmental monitoring and ecological studies program. 1974 annual report for the Prairie Island Nuclear Generating Plant near Red Wing, Minnesota. Volume II

    International Nuclear Information System (INIS)

    1975-06-01

    Data are presented from studies on the effects of thermal effluents from the Prairie Island nuclear power plant on fish and invertebrate populations in the Mississippi River in the vicinity of the plant. Populations of aquatic and terrestrial plants and birds in the immediate vicinity of the plant were also characterized. (U.S.)

  6. Nuclear disarmament or survival of nuclear arms?

    International Nuclear Information System (INIS)

    Stroot, J.P.

    1997-01-01

    START II has not yet been ratified by the US or Russian parliaments. Doubts may be raised over whether it will ever be. In the best case there will be more than 20,000 nuclear warheads in the arsenals of these two countries by the year 2003. All five nuclear states consider that nuclear weapons are an essential component of their national defense. It might sound childish but, the whole story is is so often childish: the five powers refuse to break their nuclear toys. They take even all possible measures to maintain and improve them and to ensure the survivability of their arsenals. To prepare for the next arms race..

  7. Exxon Nuclear Company ECCS evaluation of a 2-loop Westinghouse PWR with dry containment using the ENC WREM-II ECCS model. Large break example problem

    International Nuclear Information System (INIS)

    Krajicek, J.E.

    1977-01-01

    This document is presented as a demonstration of the ENC WREM-II ECCS model calculational procedure applied to a Westinghouse 2-loop PWR with a dry containment (R. E. Ginna plant, for example). The hypothesized Loss-of-Coolant Accident (LOCA) investigated was a split break with an area equal to twice the pipe cross-sectional area. The break was assumed to occur in one pump discharge pipe (DECLS break). The analyses involved calculations using the ENC WREM-II model. The following codes were used: RELAP4-EM/ENC26A for blowdown and hot channel analyses, RELAP4-EM FLOOD/ENC26A for core reflood analysis, CONTEMPT LT/22 modified for containment backpressure analysis, and TOODEE2/APR77 for heatup analysis

  8. Detection of local sodium boiling in the nuclear boiling generator in KNK II and in the cores of SNR 300 and SNR 2

    International Nuclear Information System (INIS)

    Erhardt, J.; Hoppe, P.

    1977-03-01

    As a basis of a global detection system, the detection of local boiling in sodium cooled reactors via surveillance of the neutron flux background noise is of special importance. With the help of parameter studies it is investigated in the present report, which parts of the core of SNR 300 and SNR 2 could be monitored with such a detection system. As a comparison the detection sensibility of the planned boiling generator in KNK II is determined

  9. Collaborative Russian-US work in nuclear material protection, control and accounting at the Institute of Physics and Power Engineering. II. extension to additional facilities

    International Nuclear Information System (INIS)

    Kuzin, V.V.; Pshakin, G.M.; Belov, A.P.

    1996-01-01

    During 1995, collaborative Russian-US nuclear material protection, control and accounting (MPC ampersand A) tasks at the Institute of Physics and Power Engineering (IPPE) in Obninsk, Russia focused on improving the protection of nuclear materials at the BFS Fast Critical Facility. BFS has thousands of fuel disks containing highly enriched uranium and weapons-grade plutonium that are used to simulate the core configurations of experimental reactors in two critical assemblies. Completed tasks culminated in demonstrations of newly implemented equipment and methods that enhanced the MPC ampersand A at BFS through computerized accounting, nondestructive inventory verification measurements, personnel identification and assess control, physical inventory taking, physical protection, and video surveillance. The collaborative work is now being extended. The additional tasks encompass communications and tamper-indicating devices; new storage alternatives; and systemization of the MPC ampersand A elements that are being implemented

  10. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumntation, and measurement techniques in fuel fabrication facilities, P.O.1236909. Final report

    International Nuclear Information System (INIS)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-12-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. Some of the material included has appeared elswhere and it has been summarized. An extensive bibliography is included. A spcific example of application of the accountability methods to a model fuel fabrication facility which is based on the Westinghouse Anderson design

  11. Material control in nuclear fuel fabrication facilities. Part II. Accountability, instrumntation, and measurement techniques in fuel fabrication facilities, P. O. 1236909. Final report

    Energy Technology Data Exchange (ETDEWEB)

    Borgonovi, G.M.; McCartin, T.J.; McDaniel, T.; Miller, C.L.; Nguyen, T.

    1978-12-01

    This report describes the measurement techniques, the instrumentation, and the procedures used in accountability and control of nuclear materials, as they apply to fuel fabrication facilities. Some of the material included has appeared elswhere and it has been summarized. An extensive bibliography is included. A spcific example of application of the accountability methods to a model fuel fabrication facility which is based on the Westinghouse Anderson design.

  12. Health effects model for nuclear power plant accident consequence analysis. Part I. Introduction, integration, and summary. Part II. Scientific basis for health effects models

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.S.; Moeller, D.W.; Cooper, D.W.

    1985-07-01

    Analysis of the radiological health effects of nuclear power plant accidents requires models for predicting early health effects, cancers and benign thyroid nodules, and genetic effects. Since the publication of the Reactor Safety Study, additional information on radiological health effects has become available. This report summarizes the efforts of a program designed to provide revised health effects models for nuclear power plant accident consequence modeling. The new models for early effects address four causes of mortality and nine categories of morbidity. The models for early effects are based upon two parameter Weibull functions. They permit evaluation of the influence of dose protraction and address the issue of variation in radiosensitivity among the population. The piecewise-linear dose-response models used in the Reactor Safety Study to predict cancers and thyroid nodules have been replaced by linear and linear-quadratic models. The new models reflect the most recently reported results of the follow-up of the survivors of the bombings of Hiroshima and Nagasaki and permit analysis of both morbidity and mortality. The new models for genetic effects allow prediction of genetic risks in each of the first five generations after an accident and include information on the relative severity of various classes of genetic effects. The uncertainty in modeloling radiological health risks is addressed by providing central, upper, and lower estimates of risks. An approach is outlined for summarizing the health consequences of nuclear power plant accidents. 298 refs., 9 figs., 49 tabs.

  13. Health effects model for nuclear power plant accident consequence analysis. Part I. Introduction, integration, and summary. Part II. Scientific basis for health effects models

    International Nuclear Information System (INIS)

    Evans, J.S.; Moeller, D.W.; Cooper, D.W.

    1985-07-01

    Analysis of the radiological health effects of nuclear power plant accidents requires models for predicting early health effects, cancers and benign thyroid nodules, and genetic effects. Since the publication of the Reactor Safety Study, additional information on radiological health effects has become available. This report summarizes the efforts of a program designed to provide revised health effects models for nuclear power plant accident consequence modeling. The new models for early effects address four causes of mortality and nine categories of morbidity. The models for early effects are based upon two parameter Weibull functions. They permit evaluation of the influence of dose protraction and address the issue of variation in radiosensitivity among the population. The piecewise-linear dose-response models used in the Reactor Safety Study to predict cancers and thyroid nodules have been replaced by linear and linear-quadratic models. The new models reflect the most recently reported results of the follow-up of the survivors of the bombings of Hiroshima and Nagasaki and permit analysis of both morbidity and mortality. The new models for genetic effects allow prediction of genetic risks in each of the first five generations after an accident and include information on the relative severity of various classes of genetic effects. The uncertainty in modeloling radiological health risks is addressed by providing central, upper, and lower estimates of risks. An approach is outlined for summarizing the health consequences of nuclear power plant accidents. 298 refs., 9 figs., 49 tabs

  14. Proceeding of the Scientific Meeting and Presentation on Basic Research of Nuclear Science and Technology: Book II. Nuclear Chemistry, Process Technology, and Radioactive Waste Processing and Environment; Pertemuan dan Presentasi Ilmiah Penelitian Dasar Ilmu Pengetahuan dan Teknologi Nuklir. Buku II. Kimia Nuklir, Teknologi Proses, dan Pengolahan Limbah Radioaktif dan Lingkungan

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-06-01

    The proceeding contains papers presented on Scientific Meeting and Presentation on on Basic Research of Nuclear Science and Technology, held in Yogyakarta, 25-27 April 1995. This proceeding is second part of two books published for the meeting contains papers on nuclear chemistry, process technology, and radioactive waste management and environment. There are 62 papers indexed individually. (ID)

  15. Activities for the life management of nuclear power plant in Argentina

    International Nuclear Information System (INIS)

    Chocron, M.; Fabbri, S.; Versaci, R.A.

    2003-01-01

    Since September 1994 the organisation of Argentine nuclear power activities includes the following entities: Argentine Nucleo Electric S.A (NASA) with operates two NPPs (Atucha-1 and Embalse), National Regulatory Board of Nuclear Activities (ARN), National Atomic Energy Commission Commission (CNEA) Within this scheme, one of the main activities undertaken by CNEA is to provide technological assistance to NASA for NPP operation. Works on Life Management of NPPs are included in these activities. In the past, NPP Life Management in Argentina was mainly based on the corrective maintenance concept based on the replacement of damaged part detected during the periodical outages. In recent times, as a consequential of an increasing concern about ageing a Subprogram was created to take care of these subjects. The subprogram, now in the first steps, comprise several tasks: The first task involves the identification of components ageing mechanism and materials likely to age. In the area of mechanical components. a starting point is matrix involving major components such as: Reactor Pressure Vessel, Reactor Internals, Steam Generators, Pressure Tubes, Piping, etc. In this work the apply methodology for the selection of the components will be presented. (Full text)

  16. (II) complexes

    African Journals Online (AJOL)

    activities of Schiff base tin (II) complexes. Neelofar1 ... Conclusion: All synthesized Schiff bases and their Tin (II) complexes showed high antimicrobial and ...... Singh HL. Synthesis and characterization of tin (II) complexes of fluorinated Schiff bases derived from amino acids. Spectrochim Acta Part A: Molec Biomolec.

  17. Nuclear law - Nuclear safety

    International Nuclear Information System (INIS)

    Pontier, Jean-Marie; Roux, Emmanuel; Leger, Marc; Deguergue, Maryse; Vallar, Christian; Pissaloux, Jean-Luc; Bernie-Boissard, Catherine; Thireau, Veronique; Takahashi, Nobuyuki; Spencer, Mary; Zhang, Li; Park, Kyun Sung; Artus, J.C.

    2012-01-01

    This book contains the contributions presented during a one-day seminar. The authors propose a framework for a legal approach to nuclear safety, a discussion of the 2009/71/EURATOM directive which establishes a European framework for nuclear safety in nuclear installations, a comment on nuclear safety and environmental governance, a discussion of the relationship between citizenship and nuclear, some thoughts about the Nuclear Safety Authority, an overview of the situation regarding the safety in nuclear waste burying, a comment on the Nome law with respect to electricity price and nuclear safety, a comment on the legal consequences of the Fukushima accident on nuclear safety in the Japanese law, a presentation of the USA nuclear regulation, an overview of nuclear safety in China, and a discussion of nuclear safety in the medical sector

  18. Coordination Nature of 4-Mercaptoaniline to Sn(II Ion: Formation of a One Dimensional Coordination Polymer and Its Decomposition to a Mono Nuclear Sn(IV Complex

    Directory of Open Access Journals (Sweden)

    Eon S. Burkett

    2014-12-01

    Full Text Available The coordination of the bifunctional ligand 4-mercaptoaniline with aqueo us tin(II metal ion was studied. A coordination polymer was synthesized when an aqueous solution of SnCl2 was treated with 4-MA. The crystalline material is stable under atmospheric conditions retaining its oxidation state. However, when submerged in a solution saturated with oxygen, the compound oxidizes to a mononuclear tin(IV complex. Both the compounds were characterized by single crystal X-ray diffraction studies. Although the structure of the tin(IV complex was previously reported, crystal structure of this compound was redetermined.

  19. Evaluation of the ICRU operational magnitudes implantation for the photon radiation at the Angra I and Angra II nuclear power plants

    International Nuclear Information System (INIS)

    Viana, Ronaldo do Nascimento

    2006-01-01

    The measurements of photon radiation field intensity are usually performed by a radiation protection technician trained and having skill in using radiation rate meters. Nowadays, these measurements are reported on exposure quantity and used to protect exposed individuals against the radiation risks while executing their activities. The International Commission on Radiation Units and Measurements - ICRU - defined the operational quantity ambient equivalent dose H * (10). This quantity is accepted by the scientific community as the best estimative of the protection quantity effective dose, which can not be directly measured. The operational quantity H * (10) was introduced in Brazilian rules by the National Commission of Nuclear Energy - CNEN (2005a), although its adoption was conditioned to studies of convenience and applicability of implementation. The present work may contribute to these studies, as it presents the evaluation of H * (10)'s implementation at the Nuclear Central Almirante Alvaro Alberto - CNAAA. The evaluation involved radiological tests - the energy dependence and angular dependence - applied to six types of photon radiation rate meters utilized at the CNAAA, with represent around 83% of the total number of rate meters in use by CNAAA. The result of this evaluation is favorable to the quantity H * (10)'s implementation. Suggestions are presented in order to update de rate meters and the technical and administrative procedures related to the Laboratory of Calibration of Rate Meters - LCMR, belonging to CNAAA. Thus, it could be possible to perform the calibration of the rate meters at the nuclear installation. The results obtained allows to carry out new evaluations of H * (10)'s implementation on installations that perform measurements with radiation rate meters on the practice of radiation protection, in order to adopt the H * (10) quantity in our country. (author)

  20. Three Mile Island nuclear reactor accident of March 1979. Environmental radiation data: Volume II. A report to the President's Commission on the Accident at Three Mile Island

    International Nuclear Information System (INIS)

    Bretthauer, E.W.; Grossman, R.F.; Thome, D.J.; Smith, A.E.

    1981-03-01

    This report contains a listing of environmental radiation monitoring data collected in the vicinity of Three Mile Island (TMI) following the March 28, 1979 accident. These data were collected by the EPA, NRC, DOE, HHS, the Commonwealth of Pennsylvania, or the Bethlehem Steel Corporation. The original report was printed in September 1979 and the update was released in December 1979. Table 6-Summary of Department of Health, Education, and Welfare (HEW) sampling and analytical procedures; Table 7-Computer printout of environmental data collected by HEW; Table 8-Summary of US Nuclear Regulatory Commission (NRC) sampling and analytical procedures

  1. RELAP4/MOD5: a computer program for transient thermal-hydraulic analysis of nuclear reactors and related systems. User's manual. Volume II. Program implementation

    International Nuclear Information System (INIS)

    1976-06-01

    A discussion is presented of the use of the RELAP4/MOD5 computer program in simulating the thermal-hydraulic behavior of light-water reactor systems when subjected to postulated transients such as a LOCA, pump failure, or nuclear excursion. The volume is divided into main sections which cover: (1) program description, (2) input data, (3) problem initialization, (4) user guidelines, (5) output discussion, (6) source program description, (7) implementation requirements, (8) data files, (9) description of PLOTR4M, (10) description of STH20, (11) summary flowchart, (12) sample problems, (13) problem definition, and (14) problem input

  2. Summary of the Phase II, Title I engineering assessment of inactive uranium mill tailings, Phillips/United Nuclear Site, Ambrosia Lake, New Mexico

    International Nuclear Information System (INIS)

    1977-12-01

    An engineering assessment was performed of the problems resulting from the existence of radioactive uranium mill tailings at the Phillips/United Nuclear site at Ambrosia Lake, New Mexico. Services included the preparation of topographic maps, the performance of core drillings sufficient to determine areas and volumes of tailings and radiometric measurements to determine radium-contaminated materials, the evaluation of resulting radiation exposures of individuals and nearby populations, the investigation of site hydrology and meteorology, and the evaluation and costing of alternative corrective actions. Radon gas release from the 2.6 million tons of tailings at the Phillips/United Nuclear site constitutes the most significant environmental impact, although windblown tailings and external gamma radiation are also factors. The estimated radiological health effects to the general population are considered to be minimal. The two alternative actions presented are: dike stabilization, fencing, and maintenance; and adding 2 ft of stabilization cover material. Both options include remedial action at off-site structures and on-site decontamination around the tailings pile. Cost estimates for the two options are $920,000 and $2,230,000, respectively

  3. Studies of the hydrodynamic evolution of matter produced in fluctuations in p-barp collisions and in ultrarelativistic nuclear collisions. II. Transverse-momentum distributions

    International Nuclear Information System (INIS)

    Kataja, M.; Ruuskanen, P.V.; McLerran, L.D.; von Gersdorff, H.

    1986-01-01

    We study solutions to the hydrodynamic equations appropriate for ultrarelativistic nuclear collisions. We find that the matter produced in such collisions spends time t>30 fm/c at temperatures larger than 150 MeV. The transverse momentum of protons, kaons, and pions is computed in the central region of ultrarelativistic nuclear collisions. Assuming Bjorken's initial conditions for the hydrodynamic equations, and a bag-model equation of state, we show that the transverse-momentum distribution as a function of dN/dy does reflect properties of the equation of state. We demonstrate that such a distribution approximately scales as a function of (1/A)dN/dy. The relation between p/sub t/ and dN/dy is shown to be significantly altered under different assumptions about the equation of state. The transverse-momentum distribution of heavy hadrons is shown to be much enhanced relative to that of light pions. These distributions are little changed by differences in the assumptions about the initial transverse density and velocity profile. We are unable to fit the observed correlation between p/sub t/ and dE/dy observed in the Japanese-American Cooperative Emulsion Experiment

  4. Cations-clays interactions: the Fe(II) case; application to the problematic of the French deep nuclear repository field concept

    International Nuclear Information System (INIS)

    Tournassat, Ch.

    2003-07-01

    Solute Fe(II) - montmorillonite interactions are studied in anoxic conditions and at room temperature for reaction times from hour to week. Fe 2+ is shown to be sorbed on cation exchange site with the same affinity than Ca 2+ . In chloride anionic medium, Fe(II) form ionic pairs - FeCl + - which is sorbed with almost the same affinity than CaCl + and MgCl + are. The exchange thermodynamics constants derived from this study are used to simulate the change in the exchanger composition as clay river particles enter seawater. In high concentration chloride medium, as seawater, monovalent ions (Na + and CaCl + , MgCl + ionic pairs) are shown to be the major species of the exchanger. Fe 2+ is sorbed specifically on the montmorillonite edge surfaces with a very high affinity. Simple complexation model are able to model the sorption data and show that the Fe 2+ affinity for clay edge surfaces is ∼ 1000 times higher than the Zn 2+ one. Moessbauer experiments combined to sorption, titration and dissolution experiments show that the Fe 2+ sorption is due to several different reactions: - effective competitive sorption with replacement of previously sorbed or structural cations (Zn 2+ , Mg 2+ ); - cooperative sorption together with H 4 SiO 4 , in agreement with a possible surface precipitation of a Fe - Si phase; - a sorption mechanism followed by an oxidation reaction, with a release of two H + in solution per Fe(II) sorbed, and a product (Fe(Ill)) fitting better octahedral surface 'sites'. All these phenomena can not be taken into account in a classical surface complexation model. Hence, an innovative model is developed to model clay - solute interactions, based on a morphological and structural approach. Montmorillonite edge surface area was determined using two independent methods, AFM measurement and low-pressure gas adsorption, that give the same value for this area, i.e. 8.5 m 2 g -1 . The clay - solute interface was found to be constituted by a mix of, at least, 27

  5. Optimization in the nuclear fuel cycle II: Surface contamination; Otimização no ciclo do combustível nuclear III: contaminação de superfície

    Energy Technology Data Exchange (ETDEWEB)

    Pereira, W.S., E-mail: pereiras@gmail.com [Universidade Veiga de Ameida (UVA), Rio de Janeiro, RJ (Brazil); Silva, A.X.; Lopes, J.M.; Carmo, A.S.; Fernandes, T.S.; Mello, C.R., E-mail: lararapls@hotmail.com, E-mail: Ademir@nuclear.ufrj.br [Universidade Federal do Rio de Janeiro (UFRJ), RJ (Brazil); Kelecom, A. [Universidade Federal Fluminense (UFF), Niterói, RJ (Brazil)

    2017-07-01

    Optimization is one of the bases of radioprotection and aims to move doses away from the dose limit that is the borderline of acceptable radiological risk. This work aims to use the monitoring of surface contamination as a tool of the optimization process. 53 surface contamination points were analyzed at a nuclear fuel cycle facility. Three sampling points were identified with monthly mean values of contamination higher than 1 Bq ∙ cm{sup -2}, points 28, 42 and 47. These points were indicated for the beginning of the optimization process.

  6. Systematic evaluation program review of NRC Safety Topic VI-10.A associated with the electrical, instrumentation and control portions of the testing of reactor trip system and engineered safety features, including response time for the Dresden station, Unit II nuclear power plant

    International Nuclear Information System (INIS)

    St Leger-Barter, G.

    1980-11-01

    This report documents the technical evaluation and review of NRC Safety Topic VI-10.A, associated with the electrical, instrumentation, and control portions of the testing of reactor trip systems and engineered safety features including response time for the Dresden II nuclear power plant, using current licensing criteria

  7. Transfer of Tritium in the Environment after Accidental Releases from Nuclear Facilities. Report of Working Group 7 Tritium Accidents of EMRAS II Topical Heading Approaches for Assessing Emergency Situations. Environmental Modelling for Radiation Safety (Emras II) Programme

    International Nuclear Information System (INIS)

    2014-07-01

    Environmental assessment models are used for evaluating the radiological impact of actual and potential releases of radionuclides to the environment. They are essential tools for use in the regulatory control of routine discharges to the environment and also in planning measures to be taken in the event of accidental releases. They are also used for predicting the impact of releases which may occur far into the future, for example, from underground radioactive waste repositories. It is important to verify, to the extent possible, the reliability of the predictions of such models by a comparison with measured values in the environment or with predictions of other models. The IAEA has been organizing programmes of international model testing since the 1980s. These programmes have contributed to a general improvement in models, in the transfer of data and in the capabilities of modellers in Member States. IAEA publications on this subject over the past three decades demonstrate the comprehensive nature of the programmes and record the associated advances which have been made. From 2009 to 2011, the IAEA organized a programme entitled Environmental Modelling for RAdiation Safety (EMRAS II), which concentrated on the improvement of environmental transfer models and the development of reference approaches to estimate the radiological impacts on humans, as well as on flora and fauna, arising from radionuclides in the environment. Different aspects were addressed by nine working groups covering three themes: reference approaches for human dose assessment, reference approaches for biota dose assessment and approaches for assessing emergency situations. This publication describes the work of the Tritium Accidents Working Group

  8. Interactions of trans-acting factor(s) with the estradiol response element and nuclear factor 1 of the vitellogenin II gene of Japanese quail.

    Science.gov (United States)

    Gupta, S; Upadhayay, R; Kanungo, M S

    1996-08-01

    This study was directed at achieving an understanding of the mechanisms by which steroid hormones control the synthesis of vitellogenin (VTG) protein in the liver of the Japanese quail. Northern hybridization shows that administration of estradiol alone or with progesterone stimulates the synthesis of VTG mRNA. Gel mobility shift assay of DNA fragments containing the ERE and NF 1 shows that estradiol alone or with progesterone increases the levels of nuclear proteins that bind to these cis-acting elements of the promoter of the VTG gene. The cooperative effect of the two hormones seen at the level of expression of the VTG gene may be due to protein-protein interactions of trans-acting factors that bind to ERE and NF 1.

  9. Optimization of automation: II. Estimation method of ostracism rate based on the loss of situation awareness of human operators in nuclear power plants

    International Nuclear Information System (INIS)

    Lee, Seung Min; Kim, Man Cheol; Kim, Jong Hyun; Seong, Poong Hyun

    2015-01-01

    Highlights: • We analyze the relationship between Out-of-the-Loop and the loss of human operators’ situation awareness. • We propose an ostracism rate estimation method by only considering the negative effects of automation. • The ostracism rate reflects how much automation interrupts human operators to receive information. • The higher the ostracism rate is, the lower the accuracy of human operators’ SA will be. - Abstract: With the introduction of automation in various industries including the nuclear field, its side effect, referred to as the Out-of-the-Loop (OOTL) problem, has emerged as a critical issue that needs to be addressed. Many studies have been attempted to analyze and solve the OOTL problem, but this issue still needs a clear solution to provide criteria for introducing automation. Therefore, a quantitative estimation method for identifying negative effects of automation is proposed in this paper. The representative aspect of the OOTL problem in nuclear power plants (NPPs) is that human operators in automated operations are given less information than human operators in manual operations. In other words, human operators have less opportunity to obtain needed information as automation is introduced. From this point of view, the degree of difficulty in obtaining information from automated systems is defined as the Level of Ostracism (LOO). Using the LOO and information theory, we propose the ostracism rate, which is a new estimation method that expresses how much automation interrupts human operators’ situation awareness. We applied production rules to describe the human operators’ thinking processes, Bayesian inference to describe the production rules mathematically, and information theory to calculate the amount of information that human operators receive through observations. The validity of the suggested method was proven by conducting an experiment. The results show that the ostracism rate was significantly related to the accuracy

  10. Environmental Sensitivity in Nuclear Emergencies in Rural and Semi-natural Environments. Report of Working Group 8, Environmental Sensitivity of EMRAS II Topical Heading Approaches for Assessing Emergency Situations. Environmental Modelling for RAdiation Safety (EMRAS II) Programme

    International Nuclear Information System (INIS)

    2013-11-01

    Environmental assessment models are used for evaluating the radiological impact of actual and potential releases of radionuclides to the environment. They are essential tools for use in the regulatory control of routine discharges to the environment and also in planning measures to be taken in the event of accidental releases. They are also used for predicting the impact of releases which may occur far into the future, for example, from underground radioactive waste repositories. It is important to verify, to the extent possible, the reliability of the predictions of such models by comparison with measured values in the environment or by comparing them with the predictions of other models. The IAEA has been organizing programmes of international model testing since the 1980s. The programmes have contributed to a general improvement in models, in transfer data and in the capabilities of modellers in Member States. IAEA publications on this subject over the past three decades demonstrate the comprehensive nature of the programmes and record the associated advances which have been made. From 2009 to 2011, the IAEA organized a programme entitled Environmental Modelling for RAdiation Safety (EMRAS II), which concentrated on the improvement of environmental transfer models and the development of reference approaches to estimate the radiological impacts on humans, as well as on flora and fauna, arising from radionuclides in the environment. The following topics were addressed in nine working groups: Reference Approaches for Human Dose Assessment - Working Group 1: Reference Methodologies for Controlling Discharges of Routine Releases; - Working Group 2: Reference Approaches to Modelling for Management and Remediation at NORM and Legacy Sites; - Working Group 3: Reference Models for Waste Disposal Reference Approaches for Biota Dose Assessment; - Working Group 4: Biota Modelling; - Working Group 5: Wildlife Transfer Coefficient Handbook; - Working Group 6: Biota Dose

  11. Process variables consistency at Atucha I NPP

    International Nuclear Information System (INIS)

    Arostegui, E.; Aparicio, M.; Herzovich, P.; Wenzel, J.; Urrutia, G.

    1996-01-01

    A method to evaluate the different systems performance has been developed and is still under assessment. In order to perform this job a process computer upgraded in 1992 was used. In this sense and taking into account that the resolution and stability of instrumentation is higher than its accuracy process data were corrected by software. In this was, much time spent in recalibration, and also human errors were avoided. Besides, this method allowed a better record of instrumentation performance and also an early detection of instruments failure. On the other hand, the process modelization, mainly heat and material balances has also been used to check that sensors, transducers, analog to digital converters and computer software are working properly. Some of these process equations have been introduced into the computer codes, so in some cases, it is possible to have an ''on line'' analysis of process variables and process instrumentation behaviour. Examples of process analysis are: Heat exchangers, i.e. the power calculated using shell side temperatures is compared with the tube side values; turbine performance is compared with condenser water temperature; power measured on the secondary side (one minute average measurements optimized in order to eliminate process noise are compared with power obtained from primary side data); the calibration of temperatures have been made by direct measurement of redundant sensors and have shown to be the best method; in the case of pressure and differential pressure transducers are cross checked in service when it is possible. In the present paper, details of the examples mentioned above and of other ones are given and discussed. (author). 2 refs, 1 fig., 1 tab

  12. Process variables consistency at Atucha I NPP

    Energy Technology Data Exchange (ETDEWEB)

    Arostegui, E; Aparicio, M; Herzovich, P; Wenzel, J [Central Nuclear Atucha I, Nucleoelectrica S.A., Lima, Buenos Aires (Argentina); Urrutia, G [Comision Nacional de Energia Atomica, Buenos Aires (Argentina)

    1997-12-31

    A method to evaluate the different systems performance has been developed and is still under assessment. In order to perform this job a process computer upgraded in 1992 was used. In this sense and taking into account that the resolution and stability of instrumentation is higher than its accuracy process data were corrected by software. In this was, much time spent in recalibration, and also human errors were avoided. Besides, this method allowed a better record of instrumentation performance and also an early detection of instruments failure. On the other hand, the process modelization, mainly heat and material balances has also been used to check that sensors, transducers, analog to digital converters and computer software are working properly. Some of these process equations have been introduced into the computer codes, so in some cases, it is possible to have an ``on line`` analysis of process variables and process instrumentation behaviour. Examples of process analysis are: Heat exchangers, i.e. the power calculated using shell side temperatures is compared with the tube side values; turbine performance is compared with condenser water temperature; power measured on the secondary side (one minute average measurements optimized in order to eliminate process noise are compared with power obtained from primary side data); the calibration of temperatures have been made by direct measurement of redundant sensors and have shown to be the best method; in the case of pressure and differential pressure transducers are cross checked in service when it is possible. In the present paper, details of the examples mentioned above and of other ones are given and discussed. (author). 2 refs, 1 fig., 1 tab.

  13. Options Study - Phase II

    Energy Technology Data Exchange (ETDEWEB)

    R. Wigeland; T. Taiwo; M. Todosow; W. Halsey; J. Gehin

    2010-09-01

    The Options Study has been conducted for the purpose of evaluating the potential of alternative integrated nuclear fuel cycle options to favorably address the issues associated with a continuing or expanding use of nuclear power in the United States. The study produced information that can be used to inform decisions identifying potential directions for research and development on such fuel cycle options. An integrated nuclear fuel cycle option is defined in this study as including all aspects of the entire nuclear fuel cycle, from obtaining natural resources for fuel to the ultimate disposal of used nuclear fuel (UNF) or radioactive wastes. Issues such as nuclear waste management, especially the increasing inventory of used nuclear fuel, the current uncertainty about used fuel disposal, and the risk of nuclear weapons proliferation have contributed to the reluctance to expand the use of nuclear power, even though it is recognized that nuclear power is a safe and reliable method of producing electricity. In this Options Study, current, evolutionary, and revolutionary nuclear energy options were all considered, including the use of uranium and thorium, and both once-through and recycle approaches. Available information has been collected and reviewed in order to evaluate the ability of an option to clearly address the challenges associated with the current implementation and potential expansion of commercial nuclear power in the United States. This Options Study is a comprehensive consideration and review of fuel cycle and technology options, including those for disposal, and is not constrained by any limitations that may be imposed by economics, technical maturity, past policy, or speculated future conditions. This Phase II report is intended to be used in conjunction with the Phase I report, and much information in that report is not repeated here, although some information has been updated to reflect recent developments. The focus in this Options Study was to

  14. Rise, fall and resurrection of chromosome territories: a historical perspective. Part II. Fall and resurrection of chromosome territories during the 1950s to 1980s. Part III. Chromosome territories and the functional nuclear architecture: experiments and models from the 1990s to the present.

    Science.gov (United States)

    Cremer, T; Cremer, C

    2006-01-01

    Part II of this historical review on the progress of nuclear architecture studies points out why the original hypothesis of chromosome territories from Carl Rabl and Theodor Boveri (described in part I) was abandoned during the 1950s and finally proven by compelling evidence forwarded by laser-uv-microbeam studies and in situ hybridization experiments. Part II also includes a section on the development of advanced light microscopic techniques breaking the classical Abbe limit written for readers with little knowledge about the present state of the theory of light microscopic resolution. These developments have made it possible to perform 3D distance measurements between genes or other specifically stained, nuclear structures with high precision at the nanometer scale. Moreover, it has become possible to record full images from fluorescent structures and perform quantitative measurements of their shapes and volumes at a level of resolution that until recently could only be achieved by electron microscopy. In part III we review the development of experiments and models of nuclear architecture since the 1990s. Emphasis is laid on the still strongly conflicting views about the basic principles of higher order chromatin organization. A concluding section explains what needs to be done to resolve these conflicts and to come closer to the final goal of all studies of the nuclear architecture, namely to understand the implications of nuclear architecture for nuclear functions.

  15. DETAILS OF OPERATIONS PERFORMED BY THE REMOTE CONTROL ROBOT (CONCEPT TO THE HORIZONTAL FUEL CHANNEL DURING DECOMMISSIONING PHASE OF NUCLEAR REACTOR CALANDRIA STRUCTURE. PART II: INSIDE OPERATIONS

    Directory of Open Access Journals (Sweden)

    Constantin POPESCU

    2017-05-01

    Full Text Available The authors contribution to this paper is to present a concept solution of a remote control robot (RCR used for decommissioning of the horizontal fuel channels pressure tube in the CANDU nuclear reactor. In this paper the authors highlight few details of geometry, operations, constraints by kinematics and dynamics of the robot movement inside of the reactor fuel channel. Inside operations performed has as the main steps of dismantling process the followings: unblock and extract the channel closure plug (from End Fitting - EF, unblock and extract the channel shield plug (from Lattice Tube - LT, cut the ends of the pressure tube, extract the pressure tube and cut it in small parts, sorting and storage extracted items in the safe robot container. All steps are performed in automatic mode. The remote control robot (RCR represents a safety system controlled by sensors and has the capability to analyze any error registered and decide next activities or abort the inside decommissioning procedure in case of any risk rise in order to ensure the environmental and workers protection.

  16. Inertial electrostatic confinement and nuclear fusion in the interelectrode plasma of a nanosecond vacuum discharge. II: Particle-in-cell simulations

    International Nuclear Information System (INIS)

    Kurilenkov, Yu. K.; Tarakanov, V. P.; Gus'kov, S. Yu.

    2010-01-01

    Results of particle-in-sell simulations of ion acceleration by using the KARAT code in a cylindrical geometry in the problem formulation corresponding to an actual experiment with a low-energy vacuum discharge with a hollow cathode are presented. The fundamental role of the formed virtual cathode is analyzed. The space-time dynamics of potential wells related to the formation of the virtual cathode is discussed. Quasi-steady potential wells (with a depth of ∼80% of the applied voltage) cause acceleration of deuterium ions to energies about the electron beam energy (∼50 keV). In the well, a quasi-isotropic velocity distribution function of fast ions forms. The results obtained are compared with available data on inertial electrostatic confinement fusion (IECF). In particular, similar correlations between the structure of potential wells and the neutron yield, as well as the scaling of the fusion power density, which increases with decreasing virtual cathode radius and increasing potential well depth, are considered. The chosen electrode configuration and potential well parameters provide power densities of nuclear DD fusion in a nanosecond vacuum discharge noticeably higher than those achieved in other similar IECF systems.

  17. Mechanical behaviors of the dispersion nuclear fuel plates induced by fuel particle swelling and thermal effect II: Effects of variations of the fuel particle diameters

    International Nuclear Information System (INIS)

    Ding Shurong; Wang Qiming; Huo Yongzhong

    2010-01-01

    In order to predict the irradiation mechanical behaviors of plate-type dispersion nuclear fuel elements, the total burnup is divided into two stages: the initial stage and the increasing stage. At the initial stage, the thermal effects induced by the high temperature differences between the operation temperatures and the room temperature are mainly considered; and at the increasing stage, the intense mechanical interactions between the fuel particles and the matrix due to the irradiation swelling of fuel particles are focused on. The large-deformation thermo-elasto-plasticity finite element analysis is performed to evaluate the effects of particle diameters on the in-pile mechanical behaviors of fuel elements. The research results indicate that: (1) the maximum Mises stresses and equivalent plastic strains at the matrix increase with the fuel particle diameters; the effects of particle diameters on the maximum first principal stresses vary with burnup, and the considered case with the largest particle diameter holds the maximum values all along; (2) at the cladding near the interface between the fuel meat and the cladding, the Mises stresses and the first principal stresses undergo major changes with increasing burnup, and different variations exist for different particle diameter cases; (3) the maximum Mises stresses at the fuel particles rise with the particle diameters.

  18. Feasibility study on commercialization of fast breeder reactor cycle systems interim report of phase II. Technical study report for nuclear fuel cycle systems

    International Nuclear Information System (INIS)

    Sato, Koji; Amamoto, Ippei; Inoue, Akira

    2004-06-01

    As a part of the feasibility study on commercialization of fast breeder reactor cycle systems, the plant concept concerning the fuel cycle systems (combination of the reprocessing and the fuel fabrication) has been constructed to reduce their total cost by the introduction of various innovative techniques and to apply their utmost superior efficiency from such standpoints of a decrease in the environmental burden, better resource utilization and proliferation resistance improvement by the low decontamination transuranium element (TRU) recycle. This interim report of Phase II describes the results of an on-going study which will cover a five-year period. For oxide fuels, the system which combines the use of the advanced aqueous reprocessing using three main methods such as the crystallization method, the simplified solvent extraction method, and the extraction chromatography method for minor actinide (MA) recovery, as well as the simplified pelletizing fuel fabrication which rationalized a powder mixing process etc., has abundant current results and a high technical feasibility for the basic process. Though this system faces difficulties in the technical development of control technology of the extraction chromatography and the fabrication technology of low decontamination TRU fuel etc., its expected practical use is possible at an early stage. As for the super-critical direct extraction reprocessing, it is necessary to fulfill more basic data although further economical improvement of an advanced aqueous reprocessing is expected. The system which combines the advanced aqueous reprocessing and the gelation sphere packing fuel fabrication has the advantage of lesser dispersion of the fine powder due to the use of solution and granule in the fuel fabrication process. However, this system will shoulder additional cost for the reagent recovery process and the waste liquid treatment process due to need to dispose of a large bulk of process waste liquid. The system which

  19. The Spitzer-IRAC/MIPS Extragalactic Survey (SIMES). II. Enhanced Nuclear Accretion Rate in Galaxy Groups at z ∼ 0.2

    Science.gov (United States)

    Baronchelli, I.; Rodighiero, G.; Teplitz, H. I.; Scarlata, C. M.; Franceschini, A.; Berta, S.; Barrufet, L.; Vaccari, M.; Bonato, M.; Ciesla, L.; Zanella, A.; Carraro, R.; Mancini, C.; Puglisi, A.; Malkan, M.; Mei, S.; Marchetti, L.; Colbert, J.; Sedgwick, C.; Serjeant, S.; Pearson, C.; Radovich, M.; Grado, A.; Limatola, L.; Covone, G.

    2018-04-01

    For a sample of star-forming galaxies in the redshift interval 0.15 < z < 0.3, we study how both the relative strength of the active galactic nucleus (AGN) infrared emission, compared to that due to the star formation (SF), and the numerical fraction of AGNs change as a function of the total stellar mass of the hosting galaxy group ({M}group}* ) between 1010.25 and 1011.9 M ⊙. Using a multicomponent spectral energy distribution SED fitting analysis, we separate the contribution of stars, AGN torus, and star formation to the total emission at different wavelengths. This technique is applied to a new multiwavelength data set in the SIMES field (23 not-redundant photometric bands), spanning the wavelength range from the UV (GALEX) to the far-IR (Herschel) and including crucial AKARI and WISE mid-IR observations (4.5 μm < λ < 24 μm), where the black hole thermal emission is stronger. This new photometric catalog, which includes our best photo-z estimates, is released through the NASA/IPAC Infrared Science Archive (IRSA). Groups are identified through a friends-of-friends algorithm (∼62% purity, ∼51% completeness). We identified a total of 45 galaxies requiring an AGN emission component, 35 of which are in groups and 10 in the field. We find the black hole accretion rate (BHAR) ∝ ({M}group}* {)}1.21+/- 0.27 and (BHAR/SFR) ∝ ({M}group}* {)}1.04+/- 0.24, while, in the same range of {M}group}* , we do not observe any sensible change in the numerical fraction of AGNs. Our results indicate that the nuclear activity (i.e., the BHAR and the BHAR/SFR ratio) is enhanced when galaxies are located in more massive and richer groups.

  20. Nuclear magnetic resonance studies of ancient buried wood-II. Observations on the origin of coal from lignite to bituminous coal

    Science.gov (United States)

    Hatcher, P.G.; Breger, I.A.; Szeverenyi, N.; Maciel, G.E.

    1982-01-01

    Coalified logs ranging in age from Late Pennsylvania to Miocene and in rank from lignite B to bituminous coal were analyzed by 13C nuclear magnetic resonance (NMR) utilizing the cross-polarization, magic-angle spinning technique, as well as by infrared spectroscopy. The results of this study indicate that at least three major stages of coalification can be observed as wood gradually undergoes transformation to bituminous coal. The first stage involves hydrolysis and loss of cellulose from wood with retention and differential concentration of the resistant lignin. The second stage involves conversion of the lignin residues directly to coalified wood of lignitic rank, during which the oxygen content of intermediate diagenetic products remains constant as the hydrogen content and the carbon content increases. These changes are thought to involve loss of methoxyl groups, water, and C3 side chains from the lignin. In the third major stage of coalification, the coalified wood increases in rank to subbituminous and bituminous coal; during this stage the oxygen content decreases, hydrogen remains constant, and the carbon content increases. These changes are thought to result from loss of soluble humic acids that are rich in oxygen and that are mobilized during compaction and dewatering. Relatively resistant resinous substances are differentially concentrated in the coal during this stage. The hypothesis that humic acids are formed as mobile by-products of the coalification of lignin and function only as vehicles for removal of oxygen represents a dramatic departure from commonly accepted views that they are relatively low-molecular-weight intermediates formed during the degradation of lignin that then condense to form high-molecular-weight coal structures. ?? 1982.