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)

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)

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

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

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)

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)

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.

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

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.

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)

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)

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)

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)

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)

New fuel advanced heavy water reactors

International Nuclear Information System (INIS)

Notari, Carla

1999-01-01

A redesign of the PHWR fuel element (FE) to be used in all Argentine nuclear power plants has been proposed elsewhere. This new FE presents several characteristics aimed to an improved in-core performance and economical benefits derived from the unification of most of the fabrication processes that today constitute two different production lines: one for Embalse nuclear power plant CANDU type fuel and another for Atucha I. Atucha I and Embalse, the two operating nuclear power plants in Argentina, are PHWR of different conception. Atucha I (357 M we) is of pressure vessel type and the fuel elements are full-length assemblies (530 cm of active length) with 36 uranium rods in the cluster and a support one in the outer ring. Embalse (648 M we) is a CANDU pressure tube reactor fuelled with the well known 37 rod / 50 cm length fuel bundles, twelve of which are loaded in each channel. The more relevant changes in the proposed design are an increased subdivision of the fuel material in 52 rods and a 100 cm long bundle. The combined features give the adequate channel pressure drop. The proposed CARA design shows a superior neutronic performance than the standard PHWR fuel elements currently used in Atucha I and Embalse nuclear power plants. A variant of the CARA FE consisting in the elimination of the central four rods, leaving 48 rods and a central free space, is strongly recommended because it saves materials (less uranium, less sheaths) with no loss of burnup. The central D 2 O zone allows a better utilization of the inner rods and compensates the diminished uranium loading. In Embalse no differences in core physics are expected except the beneficial decrease in linear power density. In Atucha I besides the lower power density, a higher exit burnup appears as a consequence of the higher uranium inventory. The exit burnup figures have been calculated with cell and reactor models and the result is that similar fuel management schemes as the proposed for Atucha I for the

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

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

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

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)

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

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

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.)

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)

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

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)

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

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

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.

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

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)

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

Site of the First Spanish Nuclear Power Stations; Los Emplazamientos de las Primeras Centrales Nucleares Espanolas

Energy Technology Data Exchange (ETDEWEB)

Pascual, F.; Alonso, A.; Norena, S.; Sevilla, A. [Junta de Energia Nuclear, Madrid (Spain)

1967-09-15

(e), con reactor de agua en ebullicion y contencion por condensacion de vapor, y la central de Vandellos, de 480 MW(e), con reactor grafito-gas y diseno integrado del tipo EDF-4. Los dos primeros estan situados en el interior del pais, el tercero en la costa mediterranea. Se discuten los criterios convencionales y los relacionados con la seguridad, que se han utilizado en la seleccion. Se pone de relieve la influencia de la orografia espanola en el transporte del equipo pesado y su repercusion en la seleccion de emplazamientos para centrales nucleares. Se hace referencia a la experiencia obtenida en el transporte del equipo de la central de Zorita. Se analizan las dificultades practicas encontradas al aplicar unas determinadas normas de seguridad a emplazamientos de centrales de tipo y caracteristicas de diseno diferentes. Ya que el tipo de reactor y su sistema de contencion influyen en el valor practico de la aplicacion de unas determinadas normas que, ademas, han de adaptarse a las caracteristicas peculiares de cada nacion, se indica la conveniencia de utilizar de forma adecuada y flexible las normas de seguridad para el analisis de los emplazamientos de centrales nucleares. Se describen las caracteristicas geologicas y meteorologicas de los tres emplazamientos y se resumen los trabajos e investigaciones realizados en estos campos. En especial, se destacan los estudios de hidrologia realizados en el caso de Zorita y Santa Maria de Garofia, y se describen las estaciones meteorologicas montadas en ambos emplazamientos, los experimentos de difusion con globos equilibrados realizados en el caso de Santa Maria de Garofia y los resultados de estas investigaciones. (author)

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

Atucha I nuclear power plant moderator system operates with highly radioactive heavy water, a pressure of 115 Bar and temperatures of about 200 C degrees. In March 2000, an increasing leakage of heavy water to the conventional thermal circuit was detected, conducting the plant to a shut down. The development of a number of actions and measures were taken, in order to plug this leakage. The leakage was found in a heat exchanger, which is located in a place of difficult access, with a high radiological yield and which, according to design, it was not considered to be mechanically repaired. It is a U bend tubes heat exchanger, weighting about 20 tons, and with a heavy water flow of 800 tons/h on the primary circuit, and 950 tons/h of ordinary water on the secondary side. Foreseeing this event, it had been designed and constructed special equipment and procedures, by means of a contract, with the Company INVAP SA. Repair works were carried out within a period of eighty-six (86) days, from which, forty five days were used to repair the component itself. A considerable amount of time was used to prepare simulators and the training of personnel. Due to the high radiological yield and the strict care of radiological standards, it was necessary the participation of 300 persons, integrating a collective dose of 4,86 Sv-m. It was necessary the construction of platforms and auxiliary stairs so as to make the work place accessible, as well as lifting and movement devices for heavy components, since this area does not have such kind of facilities. Welding and cutting machines remote controlled as well as manipulators which operated in front of the exchanger tube sheet were used. The aim was the reduction of dose values as much as possible. Special shielding were developed and in some cases it was necessary the adoption of drastic measures such as the cutting of bolts or pipes. The failure was detected and the tube was plugged. Also were plugged those tubes with wall thickness

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

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

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

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.

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)

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

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

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

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)

Determination of activity by gamma spectrometry of radionuclides present in drums of residues generated in nuclear centrals; Determinacion de actividad por espectrometria gamma de radionucleidos presentes en tambores de residuos generados en centrales nucleares

Energy Technology Data Exchange (ETDEWEB)

Aguiar, J.C.; Fernandez, J. [Autoridad Regulatoria Nuclear, Av. Del Libertador 8250, Ciudad Autonoma de Buenos Aires (Argentina)]. e-mail: jaguiar@cae.arn.gov.ar

2006-07-01

The generation of radioactive residuals in nuclear centrals as CNA I (Atucha I Nuclear Central) and CNE (Embalse Nuclear Central) makes that the measurement of those radionuclides has been a previous stage to the waste management. A method used in those nuclear centrals it is the gamma spectrometry with HPGe detectors, previous to the immobilization of the residual in a cemented matrix, with this the contact with the external agents and its possible dispersion to the atmosphere in the short term is avoided. The ARN (Nuclear Regulatory Authority) of Argentina it carries out periodically intercomparisons and evaluations of the measurement and procedures systems used in the nuclear power stations for the correct measurement and determination of activity of radioactive residuals by gamma spectrometry. In this work an independent method of measurement is exposed to the nuclear power stations. To determine the activity of the residuals by gamma spectrometry deposited in drums, it is required of the precise knowledge of the efficiency curve for such geometry and matrix. Due to the RNA doesn't have a pattern of these characteristics, a mathematical model has been used to obtain this efficiency curve. For it, it is necessary to determine previously: 1) the geometric efficiency or solid angle sustained by the source-detector system (drum-detector) applying a mathematical model described in this work. 2) To estimate the auto-attenuation factor that present the photons in the cemented matrix, these calculations are carried out with a simple equation and its are verified with the Micro Shield 6.10 program. The container commonly used by these nuclear power stations its are drums for 220 liters constructed with SAE 1010 steel and with a thickness of 0.127 cm, with an approximate weight 7.73 Kg., internal diameter of 57.1 cm, and height: 87 cm. The results obtained until the moment register a discrepancy from 5 to 10% with relationship to the measurements carried out by the

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)

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

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)

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)

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)

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

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)

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)

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

Centralized digital computer control of a research nuclear reactor

International Nuclear Information System (INIS)

Crawford, K.C.

1987-01-01

A hardware and software design for the centralized control of a research nuclear reactor by a digital computer are presented, as well as an investigation of automatic-feedback control. Current reactor-control philosophies including redundancy, inherent safety in failure, and conservative-yet-operational scram initiation were used as the bases of the design. The control philosophies were applied to the power-monitoring system, the fuel-temperature monitoring system, the area-radiation monitoring system, and the overall system interaction. Unlike the single-function analog computers currently used to control research and commercial reactors, this system will be driven by a multifunction digital computer. Specifically, the system will perform control-rod movements to conform with operator requests, automatically log the required physical parameters during reactor operation, perform the required system tests, and monitor facility safety and security. Reactor power control is based on signals received from ion chambers located near the reactor core. Absorber-rod movements are made to control the rate of power increase or decrease during power changes and to control the power level during steady-state operation. Additionally, the system incorporates a rudimentary level of artificial intelligence

Special Nuclear Material Control by the Power Reactor Operator; Controle des Matieres Nucleaires Speciales par l'Exploitant d'une Centrale Nucleaire; Spetsial'nyj kontrol' nalichiya yadernykh materialov operatorom ehnergeticheskogo reaktora; Control de Materiales Nucleares Especiales por Parte de Quienes Operan el Reactor de Potencia

Energy Technology Data Exchange (ETDEWEB)

Cordin, R. A. [Yankee Atomic Electric Company, Boston, MA (United States)

1966-02-15

matieres nucleaires ne se limite pas S de simples travaux d'inventaire mais sert de base a beaucoup d'autres activites qui font partie integrante du programme d'operations de tout reacteur, par exemple les expeditions de combustible irradie, le traitement chimique du combustible epuise et la comptabilite du combustible recupere et des matieres produites au cours du fonctionnement du reacteur, et l'institution et l'application d'un regime d'assurance satisfaisant. (author) [Spanish] Combustible relativamente nuevo y sumamente valioso para la produccion de energia electrica, el uranio requiere un control muy minucioso desde el momento en que la direccion de una central asume la responsabilidad financiera inherente a su posesion hasta que como combustible parcialmente agotado se transfiere a otra instalacion en la que se recupera la parte que no se ha consumido. Antes de que se descubriera la posibilidad de emplear la energia nuclear para producir electricidad, la mayor parte de las empresas que actualmente explotan centrales nucleares explotaban centrales alimentadas con combustibles fosiles y hablan establecido sistemas de control relativamente completos y adecuados para los combustibles de ese tipo. Los responsables de las centrales nucleoelectricas deben disponer de sistemas no menos adecuados para controlar los materiales nucleares especiales que utilizan. La explotacion de los reactores de potencia no es una ciencia antigua, pero durante el tiempo relativamente corto que ha transcurrido desde que se inicio su empleo los ingenieros y hombres de ciencia han mejorado continuamente el diseflo del equipo y los metodos de trabajo con objeto de disminuir los costos de produccion y de lograr que las centrales nucleares puedan competir en el plano economico con las centrales clasicas. La administracion de los materiales nucleares debe efectuarse con metodos modernos y eficientes a fin de que los adelantos tecnologicos que han permitido reducir los costos no resulten inutiles

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)

Management of Spent Nuclear Fuel from Nuclear Power Plant Reactor

International Nuclear Information System (INIS)

Wati, Nurokhim

2008-01-01

Management of spent nuclear fuel from Nuclear Power Plant (NPP) reactor had been studied to anticipate program of NPP operation in Indonesia. In this paper the quantity of generated spent nuclear fuel (SNF) is predicted based on the national electrical demand, power grade and type of reactor. Data was estimated using Pressurized Water Reactor (PWR) NPP type 1.000 MWe and the SNF management overview base on the experiences of some countries that have NPP. There are four strategy nuclear fuel cycle which can be developed i.e: direct disposal, reprocessing, DUPlC (Direct Use of Spent PWR Fuel In Candu) and wait and see. There are four alternative for SNF management i.e : storage at the reactor building (AR), away from reactor (AFR) using wet centralized storage, dry centralized storage AFR and prepare for reprocessing facility. For the Indonesian case, centralized facility of the wet type is recommended for PWR or BWR spent fuel. (author)

Nuclear Energy in Central Europe 98, Proceedings

International Nuclear Information System (INIS)

Ravnik, M.; Jencic, I.; Zagar, T.

1998-01-01

Regional Meeting for Nuclear Energy in Central Europe is an annual meeting of the Nuclear Society of Slovenia. The proceedings contain 63 articles from Slovenia, sorounding countries and countries of the Central and Eastern European Region. Topics are: Research Reactors, Nuclear Methods, Reactor Physics, Thermal Hydraulics, Structural Analysis, Probabilistic Safety Assessment, Severe Accidents, NPP Operation and Nuclear Waste disposal

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.

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

Preparation of mandatory documentation before the start up of the RA-0 `zero power` nuclear reactor at Cordoba National University; Preparacion de la documentacion mandatoria para la puesta en marcha del reactor nuclear RA-0 en la Universidad Nacional de Cordoba

Energy Technology Data Exchange (ETDEWEB)

Martin, H R; Keil, W M; Pezzi, N

1992-12-31

Before the start up of the RA-0 `zero power` nuclear reactor installed at Cordoba National University, it was necessary to send to the Regulatory Authority the mandatory documentation which is required in the licensing process. With the previous papers existing for the operation in the first years of the `70, a work program for the future operational training personnel was elaborated. Based on the Authority`s applicable rules and the recommendations and with particular criteria originated in the working university conditions, the SAFETY report of RA-0 nuclear reactor was prepared. This paper describes the principal contents, items and documents involved in the safety report. (Author). [Espanol] Con motivo de la nueva puesta en servicio del REACTOR NUCLEAR RA-0 fue necesario elaborar la documentacion mandatoria requerida por la Autoridad Regulatoria Nacional. Siguiendo los lineamientos de las normas y recomendaciones vigentes e incluyendo criterios propios en lo que debia ser el contenido final de dicha documentacion, fue preparado lo que se ha denominado el INFORME DE SEGURIDAD DEL REACTOR NUCLEAR RA-0. Este documento que se describe en este trabajo, si bien contiene las habituales descripciones de todos los Informes de Seguridad, incluye otros aspectos que no siendo requeridos expresamente en el mismo, han dado una mayor coherencia a la conformacion de todos los aspectos que interrelacionan las areas de seguridad fisica, radiologica, nuclear y de control de materiales nucleares bajo salvaguardias. (Autor).

Nuclear reactor instrumentation at research reactor renewal

International Nuclear Information System (INIS)

Baers, B.; Pellionisz, P.

1981-10-01

The paper overviews the state-of-the-art of research reactor renewals. As a case study the instrumentation reconstruction of the Finnish 250 kW TRIGA reactor is described, with particular emphasis on the nuclear control instrumentation and equipment which has been developed and manufactured by the Central Research Institute for Physics, Budapest. Beside the presentation of the nuclear instrument family developed primarily for research reactor reconstructions, the quality assurance policy conducted during the manufacturing process is also discussed. (author)

The program of reactors and nuclear power plants; Programa de reactores y centrales nucleares

Energy Technology Data Exchange (ETDEWEB)

Calabrese, Carlos R [Comision Nacional de Energia Atomica, General San Martin (Argentina). Centro Atomico Constituyentes

2001-07-01

Into de framework of the program of research reactors and nuclear power plants, the operating Argentine reactors are described. The uses of the research reactors in Argentina are summarized. The reactors installed by Argentina in other countries (Peru, Algeria, Egypt) are briefly described. The CAREM project for the design and construction of an innovator small power reactor (27 MWe) is also described in some detail. The next biennial research and development program for reactor is briefly outlined.

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)

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

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)

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

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)

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)

THERMOS, district central heating nuclear reactors

International Nuclear Information System (INIS)

Patarin, L.

1981-02-01

In order to expand the penetration of uranium in the national energy balance sheet, the C.E.A. has been studying nuclear reactors for several years now, that are capable of providing heat at favourable economic conditions. In this paper the THERMOS model is introduced. After showing the attraction of direct town heating by nuclear energy, the author describes the THERMOS project, defines the potential market, notably in France, and applies the lay-out study to the Grenoble Nuclear Study Centre site with district communal heating in mind. The economic aspects of the scheme are briefly mentioned [fr

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)

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

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

Uso de detectores de neutrinos para el monitoreo de reactores nucleares Uso de detectores de neutrinos para el monitoreo de reactores nucleares

Directory of Open Access Journals (Sweden)

Gerardo Moreno

2012-02-01

Full Text Available Se estudia la factibilidad del uso de los detectores de antineutrinos para el monitoreo de reactores nucleares. Usando un modelo sencillo de cascada de fisión a dos componentes, se ilustra la dependencia del número de antineutrinos detectados a una distancia L del reactor según la composición nuclear del combustible. Se explica el principio de detección de neutrinos de reactores en base al decaimiento beta inverso y se describe como los detectores de neutrinos pueden emplearse para el monitoreo de la producción de materiales fisibles en el reactor. Se comenta como generalizar este análisis al caso real de un reactor nuclear in situ y uno de los principales experimentos internacionales dedicados a este propósito. We study the feasibility to use antineutrinos detectors for monitoring of nuclear reactors. Using a simple model of fission shower with two components, we illustrate how the numbers of antineutrinos detected at a distance L from the reactor depend on the composition of the nuclear combustible. We explain the principles of reactor neutrino detection using inverse beta decays and we describe how neutrinos detectors can be used for monitoring the production of fissile materials within the reactors. We comment how to generalize this analysis to the realistic case of a nuclear reactor in situ and one of the main international experiments dedicated to study the use of neutrinos detectors as nuclear safeguards.

Dispositivo de posicionamiento de muestras biológicas para su irradiación en un canal radial de un reactor nuclear // Biological samples positioning device for irradiations on a radial channel at the nuclear research reactor

Directory of Open Access Journals (Sweden)

Maritza Rodríguez - Gual

2010-05-01

Full Text Available ResumenPor la demanda de un dispositivo experimental para el posicionamiento de las muestras biológicaspara su irradiación en un canal radial de un reactor nuclear de investigaciones en funcionamiento, seconstruyó y se puso en marcha un dispositivo para la colocación y retirada de las muestras en laposición de irradiación de dicho canal. Se efectuaron las valoraciones económicas comparando conotro tipo de dispositivo con las mismas funciones. Este trabajo formó parte de un proyectointernacional entre Cuba y Brasil que abarcó el estudio de los daños inducidos por diferentes tipos deradiación ionizante en moléculas de ADN. La solución propuesta es comprobada experimentalmente,lo que demuestra la validez práctica del dispositivo. Como resultado del trabajo, el dispositivoexperimental para la irradiación de las muestras biológicas se encuentra instalado y funcionando yapor 5 años en el canal radial # 3(BH#3 Palabras claves: reactor nuclear de investigaciones, dispositivo para posicionamiento de muestras,___________________________________________________________________________AbstractFor the demand of an experimental device for biological samples positioning system for irradiationson a radial channel at the nuclear research reactor in operation was constructed and started up adevice for the place and remove of the biological samples from the irradiation channels withoutinterrupting the operation of the reactor. The economical valuations are effected comparing withanother type of device with the same functions. This work formed part of an international projectbetween Cuba and Brazil that undertook the study of the induced damages by various types ofionizing radiation in DNA molecules. Was experimentally tested the proposed solution, whichdemonstrates the practical validity of the device. As a result of the work, the experimental device forbiological samples irradiations are installed and operating in the radial beam hole #3(BH#3

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

Flow of ideal fluid through a central region of a nuclear reactor wire-spaced fuel subassembly

International Nuclear Information System (INIS)

Schmid, J.

1991-04-01

The results are given of calculations of the flow of an ideal fluid through the central region of a nuclear reactor wire-spaced fuel subassembly. The computer code used is briefly described. (author). 10 figs., 4 refs

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

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

Argentine activities within the IWGATWR framework

International Nuclear Information System (INIS)

Antunez, H.M.

1991-01-01

The status of power reactors and nuclear activity in Argentina for the period 1988-1990 are described. The reactor shutdown at the Central Nuclear Atucha I and repair works are discussed. Joint venture of the Comision Nacional de Energia Atomica and KWU developments are reported, including the conceptual design of an additional heat sink for pressurized water reactors, a feasibility analysis of the Co-60 production in pressure vessel PHWRs and a study of severe postulated accidents for pressure vessel PHWR. 11 refs, 4 figs

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

Fourth Regional Meeting: Nuclear Energy in Central Europe, Proceedings

Energy Technology Data Exchange (ETDEWEB)

Mavko, B; Cizelj, L [eds.; Nuclear Society of Slovenia (Slovenia)

1997-07-01

Fourth Regional Meeting for Nuclear Energy in Central Europe is an annual meeting of the Nuclear Society of Slovenia. The proceedings contain 89 articles from Slovenia, surrounding countries and countries of the Central and Eastern European Region. Topics are: Research Reactors, Reactor Physics, Probabilistic Safety Assessment, Severe Accidents, Ageing and Integrity, Thermal Hydraulics, NPP Operation Experiance, Radioactive Waste Management, Environment and Other Aspects, Public and Nuclear Energy, SG Replacement and Plant Uprating.

Nuclear policies in Central Europe. Environmental policy and enlargement of the European Union: Austria's policies towards Nuclear Reactors in neighboring countries

International Nuclear Information System (INIS)

Getzner, M.

2003-01-01

Austria's anti-nuclear policies are rooted in the successful anti-nuclear referendum on the Zwentendorf nuclear power plant (Lower Austria) in 1978 and the great impact of the Chernobyl catastrophe on Austria in 1986. Since about 1990, official Austria has pursued anti-nuclear policies not only at home but also abroad. In particular, reactors in Central and Eastern European Countries (CEEC) are the focal points of Austria's foreign anti-nuclear policies. Strategies include increasing nuclear safety, promoting energy efficiency and sustainable energy sources (such as renewable resources), and extending international legal frameworks to account for nuclear safety. Involvement in domestic energy issues in other countries is not an easy task, and while Austrian policy makers have had some success in increasing awareness of nuclear safety in Europe, they have also made a number of strategic mistakes. Notwithstanding real and substantiated concerns regarding nuclear safety, Austrian policies have lost credibility during recent years. This book explores the history and the development of Austrian anti-nuclear policies, and discusses the political economy of such policies. Particular emphasis is laid on the 2002 referendum against the Temelin reactor in the neighboring Czech Republic. (orig.)

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

Nuclear reactors

International Nuclear Information System (INIS)

Yoshioka, Michiko.

1985-01-01

Purpose: To obtain an optimum structural arrangement of IRM having a satisfactory responsibility to the inoperable state of a nuclear reactor and capable of detecting the reactor power in an averaged manner. Constitution: As the structural arrangement of IRM, from 6 to 16 even number of IRM are bisected into equial number so as to belong two trip systems respectively, in which all of the detectors are arranged at an equal pitch along a circumference of a circle with a radius rl having the center at the position of the central control rod in one trip system, while one detector is disposed near the central control rod and other detectors are arranged substantially at an equal pitch along the circumference of a circle with a radius r2 having the center at the position for the central control rod in another trip system. Furthermore, the radius r1 and r2 are set such that r1 = 0.3 R, r2 = 0.5 R in the case where there are 6 IRM and r1 = 0.4 R and R2 = 0.8 R where there are eight IRM where R represents the radius of the reactor core. (Kawakami, Y.)

The Centralized Reliability Data Organization (CREDO); an advanced nuclear reactor reliability, availability, and maintainability data bank and data analysis center

International Nuclear Information System (INIS)

Knee, H.E.

1991-01-01

The Centralized Reliability Data Organization (CREDO) is a data bank and data analysis center, which since 1985 has been jointly sponsored by the US Department of Energy's (US DOE's) Office of Technology Support Programs and Japan's Power Reactor and Nuclear Fuel Development Corporation (PNC). It focuses on reliability, availability and maintainability (RAM) data for components (e.g. valves, pumps, etc.) operating in advanced nuclear reactor facilities. As originally intended, the purpose of the CREDO system was to provide a centralized source of accurate, up-to-date data and information for use in RAM analyses necessary for meeting DOE's data needs in the areas of advanced reactor safety assessments, design and licensing. In particular, creation of the CREDO system was considered an essential element needed to fulfill the DOE Breeder Reactor Safety Program's commitment of 'identifying and exploiting areas in which probabilistic methods can be developed and used in making reactor safety Research and Development choices and optimizing designs of safety systems'. CREDO and its operation are explained. (author)

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).

Third Regional Meeting: Nuclear Energy in Central Europe, Proceedings

International Nuclear Information System (INIS)

Stritar, A.; Jencic, I.

1996-01-01

Third Regional Meeting for Nuclear Energy in Central Europe is an annual meeting of the Nuclear Society of Slovenia. The proceedings contain 71 articles from Slovenia, surrounding countries and countries of the Central and Eastern European Region. Topics are: Research Reactors, Reactor Physics, Probabilistic Safety Assessment, Severe Accident management, Thermal Hydraulics, NPP Operation, Radioactive Waste Management, Main Components Integrity, Environment and Other Aspects and Public Information

Third Regional Meeting: Nuclear Energy in Central Europe, Proceedings

Energy Technology Data Exchange (ETDEWEB)

Stritar, A; Jencic, I [Nuclear Society of Slovenia (Slovenia); eds.

1996-07-01

Third Regional Meeting for Nuclear Energy in Central Europe is an annual meeting of the Nuclear Society of Slovenia. The proceedings contain 71 articles from Slovenia, surrounding countries and countries of the Central and Eastern European Region. Topics are: Research Reactors, Reactor Physics, Probabilistic Safety Assessment, Severe Accident management, Thermal Hydraulics, NPP Operation, Radioactive Waste Management, Main Components Integrity, Environment and Other Aspects and Public Information.

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)

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

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.)

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)

International Conference Nuclear Energy in Central Europe 99, V. 1. Proceedings

Energy Technology Data Exchange (ETDEWEB)

Gortnar, O; Stritar, A [Nuclear Society of Slovenia (Slovenia)

1999-07-01

International Conference Nuclear Energy in Central Europe is an annual meeting of the Nuclear Society of Slovenia. The proceedings contain 101 articles from Slovenia, surrounding countries and countries of the Central and Eastern European Region. Topics are: Reactor Physics, Research Reactors, Thermal Hydraulics, Structural Analysis, Probabilistic Safety Assessment, Severe Accidents, NPP Operation, Nuclear Energy and Public, Radioactive Waste, Radiological Protection and Environmental Issues, Nuclear Methods and Monte Carlo and Deterministic Transport Calculations.

International Conference Nuclear Energy in Central Europe 99, V. 1. Proceedings

International Nuclear Information System (INIS)

Gortnar, O.; Stritar, A.

1999-01-01

International Conference Nuclear Energy in Central Europe is an annual meeting of the Nuclear Society of Slovenia. The proceedings contain 101 articles from Slovenia, surrounding countries and countries of the Central and Eastern European Region. Topics are: Reactor Physics, Research Reactors, Thermal Hydraulics, Structural Analysis, Probabilistic Safety Assessment, Severe Accidents, NPP Operation, Nuclear Energy and Public, Radioactive Waste, Radiological Protection and Environmental Issues, Nuclear Methods and Monte Carlo and Deterministic Transport Calculations

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

Technology, safety and costs of decommissioning nuclear reactors at multiple-reactor stations

International Nuclear Information System (INIS)

Wittenbrock, N.G.

1982-01-01

Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWR) and large (1155-MWe) boiling water reactors (BWR) at multiple-reactor stations. Three decommissioning alternatives are studied: DECON (immediate decontamination), SAFSTOR (safe storage followed by deferred decontamination), and ENTOMB (entombment). Safety and costs of decommissioning are estimated by determining the impact of probable features of multiple-reactor-station operation that are considered to be unavailable at a single-reactor station, and applying these estimated impacts to the decommissioning costs and radiation doses estimated in previous PWR and BWR decommissioning studies. The multiple-reactor-station features analyzed are: the use of interim onsite nuclear waste storage with later removal to an offsite waste disposal facility, the use of permanent onsite nuclear waste disposal, the dedication of the site to nuclear power generation, and the provision of centralized services

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

Atucha I nuclear power plant moderator system operates with highly radioactive heavy water, a pressure of 115 Bar and temperatures of about 200 C degrees. In March 2000, an increasing leakage of heavy water to the conventional thermal circuit was detected, conducting the plant to a shut down. The development of a number of actions and measures were taken, in order to plug this leakage. The leakage was found in a heat exchanger, which is located in a place of difficult access, with a high radiological yield and which, according to design, it was not considered to be mechanically repaired. It is a U bend tubes heat exchanger, weighting about 20 tons, and with a heavy water flow of 800 tons/h on the primary circuit, and 950 tons/h of ordinary water on the secondary side. Foreseeing this event, it had been designed and constructed special equipment and procedures, by means of a contract, with the Company INVAP SA. Repair works were carried out within a period of eighty-six (86) days, from which, forty five days were used to repair the component itself. A considerable amount of time was used to prepare simulators and the training of personnel. Due to the high radiological yield and the strict care of radiological standards, it was necessary the participation of 300 persons, integrating a collective dose of 4,86 Sv-m. It was necessary the construction of platforms and auxiliary stairs so as to make the work place accessible, as well as lifting and movement devices for heavy components, since this area does not have such kind of facilities. Welding and cutting machines remote controlled as well as manipulators which operated in front of the exchanger tube sheet were used. The aim was the reduction of dose values as much as possible. Special shielding were developed and in some cases it was necessary the adoption of drastic measures such as the cutting of bolts or pipes. The failure was detected and the tube was plugged. Also were plugged those tubes with wall thickness

Proceedings of the International Conference Nuclear Energy in Central Europe 2001

International Nuclear Information System (INIS)

Jencic, I.; Glumac, B.

2001-01-01

International Conference Nuclear Energy in Central Europe is an annual meeting of the Nuclear Society of Slovenia. The proceedings contain 98 articles from Slovenia, surrounding countries and countries of the Central and Eastern European Region. Topics are: reactor physics, thermal hydraulics, probabilistic safety assessment (PSA) and severe accidents, nuclear materials, NPP and research reactor operation, environmental issues and radiation measurement, fusion, radioactive waste and regulatory issues and public relations

Nuclear Reactor Physics

Science.gov (United States)

Stacey, Weston M.

2001-02-01

An authoritative textbook and up-to-date professional's guide to basic and advanced principles and practices Nuclear reactors now account for a significant portion of the electrical power generated worldwide. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. Nuclear reactor physics is the core discipline of nuclear engineering, and as the first comprehensive textbook and reference on basic and advanced nuclear reactor physics to appear in a quarter century, this book fills a large gap in the professional literature. Nuclear Reactor Physics is a textbook for students new to the subject, for others who need a basic understanding of how nuclear reactors work, as well as for those who are, or wish to become, specialists in nuclear reactor physics and reactor physics computations. It is also a valuable resource for engineers responsible for the operation of nuclear reactors. Dr. Weston Stacey begins with clear presentations of the basic physical principles, nuclear data, and computational methodology needed to understand both the static and dynamic behaviors of nuclear reactors. This is followed by in-depth discussions of advanced concepts, including extensive treatment of neutron transport computational methods. As an aid to comprehension and quick mastery of computational skills, he provides numerous examples illustrating step-by-step procedures for performing the calculations described and chapter-end problems. Nuclear Reactor Physics is a useful textbook and working reference. It is an excellent self-teaching guide for research scientists, engineers, and technicians involved in industrial, research, and military applications of nuclear reactors, as well as government regulators who wish to increase their understanding of nuclear reactors.

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)

Operating Experience in Nuclear Power Plants with Boiling-Water Reactors; Experience acquise dans l'exploitation des reacteurs a eau bouillante; Opyt ehkspluatatsii kipyashchago reaktora; Experiencia adquirida con la explotacion de reactores de agua hirviente

Energy Technology Data Exchange (ETDEWEB)

Ascherl, R. J. [General Electric Company, San Jose, CA (United States)

1963-10-15

nucleoelectrica de Humboldt Bay (Pacific Gas and Electric Company, Eureka, California), la central nucleoelectrica de Garigliano (Societa Elettronucleare Nazionale, Scauri, Italia), y el reactor de potencia japones para fines de demostracion (Instituto de Investigaciones Nucleares, Tokai Mura, Japon). La puesta en marcha y el funcionamiento inicial de estas centrales confirma las esperanzas despertadas por las centrales de Dresden, Kahl y Vallecitos. Los registros de rendimiento de las centrales de Dresden, Kahl y Vallecitos han puesto claramente de manifiesto la estabilidad y seguridad de los reactores de agua hirviente. Ademas, los niveles de radiacion dentro de la central y en sus alrededores se han mantenido muy por debajo de los limites fijados en los permisos de explotacion. Ha quedado confirmada la sencillez y facilidad de explotacion de los reactores de agua hirviente. La rapidez de la respuesta del reactor con doble ciclo de Dresden a las fluctuaciones de la demanda es excelente. Ninguno de los trabajos de conservacion y reparacion, sea cual fuere su importancia, exige calificaciones especiales ni entrana riesgos de irradiacion excesivos. La inspeccion general y la revision de la turbina de la central de Dresden puso de manifiesto la inexistencia de problemas de conservacion despues de mas de 12 000 h de funcionamiento con el vapor en ciclo directo, a pesar de que la explotacion prosiguio despues de haberse comDtobado algunas fallas en los elementos combustibles. (author) [Russian] K nastoyashchemu vremeni priobreten znachitel'nyj opyt v ehkspluatatsii ehlektrostantsij s kipyashchim reaktorom. K kontsu 1962 goda proizvedeno svyshe 2200 mln.kvt.ch ehlektroehnergii na trekh ehlektrostantsiyakh ehnergosistemy kommunal'nogo obsluzhivaniyagna Drezdenskoj atomnoj ehlektrostantsii, ''Kommonvels Ehdison kompani'', Morris, Illinojs, na atomnoj ehlektrostantsii Valleeitos,''Pasifik gaz ehnd ehlektrik kompani'', Plezanton, Kaliforniya, i na Kal'skoj atomnoj ehlektrostantsii

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

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

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)

Nuclear safety in light water reactors severe accident phenomenology

CERN Document Server

Sehgal, Bal Raj

2011-01-01

This vital reference is the only one-stop resource on how to assess, prevent, and manage severe nuclear accidents in the light water reactors (LWRs) that pose the most risk to the public. LWRs are the predominant nuclear reactor in use around the world today, and they will continue to be the most frequently utilized in the near future. Therefore, accurate determination of the safety issues associated with such reactors is central to a consideration of the risks and benefits of nuclear power. This book emphasizes the prevention and management of severe accidents to teach nuclear professionals

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)

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

Physical Characteristics of the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Ngo Quang Huy

1994-10-01

The operation of the TRIGA MARK II reactor of nominal power 250 KW has been stopped as all the fuel elements have been dismounted and taken away in 1968. The reconstruction of the reactor was accomplished with Russian technological assistance after 1975. The nominal power of the reconstructed reactor is of 500 KW. The recent Dalat reactor is unique of its kind in the world: Russian-designed core combined with left-over infrastructure of the American-made TRIGA II. The reactor was loaded in November 1983. It has reached physical criticality on 1/11/1983 (without central neutron trap) and on 18/12/1983 (with central neutron trap). The power start up occurred in February 1984 and from 20/3/1984 the reactor began to be operated at the nominal power 500 KW. The selected reports included in the proceedings reflect the start up procedures and numerous results obtained in the Dalat Nuclear Research Institute and the Centre of Nuclear Techniques on the determination of different physical characteristics of the reactor. These characteristics are of the first importance for the safe operation of the Dalat reactor

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

Limits to the Recognizability of Flaws in Non-Destructive Testing Steam-Generator Tubes for Nuclear-Power Plants; Limitations de l'Efficacite de la Detection des Defauts par Essais Non Destructifs de Tubes de Bouilleurs pour Centrales Nucleaires; Predely ehffektivnosti nedestruktivnoj defektoskopii trub parogeneratorov yadernykh ehlektrostantsij; Limites para Distinguir los Defectos en el Ensayo No Destructivo de Tubo para Generadores de Vapor Destinados a Centrales Nucleares

Energy Technology Data Exchange (ETDEWEB)

Kuhlmann, A.; Adamsky, F. -J. [Technischer Ueberwachungs-Verein Reinland e.V., Cologne (Germany)

1965-10-15

In the Federal Republic of Germany there are nuclear reactors under construction with steam generators inside the reactor pressure-vessel. As a result design repairs of steam- generator tubes are very difficult and cause large shut-down times of the nuclear-power plant. It is known that numerous troubles in operating conventional power plants are results of steam-generator tube damages. Because of the high total costs of these reactors it. is necessary to construct the steam generators especially in such a manner that the load factor of the power plant is as high as possible. The Technischer Ueberwachungs-Verein Rheinland was charged to supervise and to test fabrication and construction of the steam generators to see that this part of the plant was as free of defects as possible. The experience gained during this work is of interest for manufacture and construction of steam generators for nuclear-power plants in general. This paper deals with the efficiency limits of non-destructive testing steam-generator tubes. The following tests performed will be discussed in detail: (a) Automatic ultrasonic testing of the straight tubes in the production facility; (b) Combined ultrasonic and radiographic testing of the bent tubes and tube weldings; (c) Other non-destructive tests. (author) [French] En Allemagne, on construit actuellement des generatrices nucleaires dans lesquelles les bouilleurs sont installes a l'interieur du caisson etanche du reacteur. Il s'ensuit que les reparations des tubes des bouilleurs sont tres difficiles et provoquent de longs arrets des centrales nucleaires. On sait que de nombreuses perturbations du fonctionnement des centrales classiques sont dues a des dommages dans les tubes des bouilleurs. C'est en grande partie en raison du prix de revient eleve de ces reacteurs qu'il est necessaire de construire les bouilleurs notamment de sorte que le facteur d'utilisation de la centrale soit aussi eleve que possible. C'est le ' Technischer Ueberwachungs

Nuclear reactor physics course for reactor operators

International Nuclear Information System (INIS)

Baeten, P.

2006-01-01

The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

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)

Nuclear reactors

International Nuclear Information System (INIS)

Barre, Bertrand

2015-10-01

After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

Nuclear reactor physics

CERN Document Server

Stacey, Weston M

2010-01-01

Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-01-01

The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors. Refs, figs, tabs.

Water channel reactor fuels and fuel channels: Design, performance, research and development. Proceedings of a technical committee meeting

International Nuclear Information System (INIS)

1998-01-01

The International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT) recommended holding a Technical Committee Meeting on Water Channel Reactor Fuel including into this category fuels and pressure tubes/fuel channels for Atucha-I and II, BWR, CANDU, FUGEN and RBMK reactors. The IWGFPT considered that even if the characteristics of Atucha, CANDUs, BWRs, FUGEN and RBMKs differ considerably, there are also common features. These features include materials aspects, as well as core, fuel assembly and fuel rod design, and some safety issues. There is also some similarity in fuel power history and operating conditions (Atucha-I and II, FUGEN and RBMK). Experts from 11 countries participated at the meeting and presented papers on technology, performance, safety and design, and materials aspects of fuels and pressure tubes/fuel channels for the above types of water channel reactors

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

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)

On possibility of vibrational stabilization in nuclear instable stationary regimes in nuclear reactors

International Nuclear Information System (INIS)

Trakhtenberg, A.M.

1987-01-01

A principle possibility of applying the vibrational stabilization method to nuclear reactors is studied. The problem of securing the stability of nuclear reactor operation steady-state regimes is one of the central ones in dynamics theory and nuclear reaction operation experience. In particular, the problem of xenon oscillation suppressing in a reactor, occuring as a result of steady-state regime instability is urgent. Investigation is conducted using the simpliest reactor model, repesenting it as a non-linear object with concentrated parameters. It is proved that vibrational stabilization is achieved by periodic fluctuations of the control rod positions in the reactor core and boric acid concentration in the coolant with period 1s 4 s. In practice stabilization is effective, when the steady-state regime is located near the stability boundary, which appears to be dangerous, i.e. self-oscillations with inadmissibly high amplitude occure in the reactor

Description of the PIE facility for research reactors irradiated fuels in CNEA

International Nuclear Information System (INIS)

Bisca, A.; Coronel, R.; Homberger, V.; Quinteros, A.; Ratner, M.

2002-01-01

The PIE Facility (LAPEP), located at the Ezeiza Atomic Center (CAE), was designed to carry out destructive and non-destructive post-irradiation examinations (PIE) on research and power reactor spent fuels, reactor internals and other irradiated materials, and to perform studies related with: Station lifetime extension; Fuel performance; Development of new fuels; and Failures and determination of their causes. LAPEP is a relevant facility where research and development can be carried out. It is worth mentioning that in this facility the PIE corresponding to the Surveillance Program for the Atucha I Nuclear Power Plant (CNA-1) were successfully performed. Materials testing during the CNA-1 repair and the study of failures in fuel element plugs of the Embalse Nuclear Power Plant (CNE) were also performed. (author)

Virtual nuclear reactor for education of nuclear reactor physics

International Nuclear Information System (INIS)

Tsuji, Masashi; Narabayashi, Takashi; Shimazu, Youichiro

2008-01-01

As one of projects that were programmed in the cultivation program for human resources in nuclear engineering sponsored by the Ministry of Economy, Trade and Industry, the development of a virtual reactor for education of nuclear reactor physics started in 2007. The purpose of the virtual nuclear reactor is to make nuclear reactor physics easily understood with aid of visualization. In the first year of this project, the neutron slowing down process was visualized. The data needed for visualization are provided by Monte Carlo calculations; The flights of the respective neutrons generated by nuclear fissions are traced through a reactor core until they disappear by neutron absorption or slow down to a thermal energy. With this visualization and an attached supplement textbook, it is expected that the learners can learn more clearly the physical implication of neutron slowing process that is mathematically described by the Boltzmann neutron transport equation. (author)

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

Final results of the cadmium and spectral ratios obtained inside of the fuel rod positioned in the central position of the IPEN/MB-01 nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Bitelli, Ulysses d' Utra; Mura, Luiz Ernesto C.; Santos, Diogo Feliciano dos, E-mail: ubitelli@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil); Lambiasi, Beatriz G.N. [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), SP (Brazil)

2015-07-01

The spectral ratios are very important to determine some nuclear reactors parameters such as reaction rates, fuel lifetime, etc and some safety operational conditions. This study aims to determine the spectral ratios in 2 (two) spatial positions located inside the core of the Nuclear Reactor IPEN/MB-01. These places are at the central position of the nuclear reactor core in an asymptotic neutron flux region. The experiment consists in inserting different activation foil detectors inside an experimental fuel rod. The experimental rod is assembled at the central position of the reactor core. Activation neutron foil detectors of different elements such as {sup 197}Au, {sup 238}U, {sup 45}Sc, {sup 58}Ni, {sup 24}Mg, {sup 47}Ti and {sup 115m}In were used to cover a large range of neutron spectrum. Saturation activity per target nucleus was obtained by gamma spectrometry using a HPGe system. The experimental cadmium ratios compared with values computed by MCNP-4C code show good agreement. (author)

Advanced nuclear reactor and nuclear fusion power generation

International Nuclear Information System (INIS)

2000-04-01

This book comprised of two issues. The first one is a advanced nuclear reactor which describes nuclear fuel cycle and advanced nuclear reactor like liquid-metal reactor, advanced converter, HTR and extra advanced nuclear reactors. The second one is nuclear fusion for generation energy, which explains practical conditions for nuclear fusion, principle of multiple magnetic field, current situation of research on nuclear fusion, conception for nuclear fusion reactor and economics on nuclear fusion reactor.

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)

Heavy water moderated tubular type nuclear reactor

International Nuclear Information System (INIS)

Oohashi, Masahisa.

1986-01-01

Purpose: To enable to effectively change the volume of heavy water per unit fuel lattice in heavy water moderated pressure tube type nuclear reactors. Constitution: In a nuclear reactor in which fuels are charged within pressure tubes and coolants are caused to flow between the pressure tubes and the fuels, heavy water tubes for recycling heavy water are disposed to a gas region formed to the outside of the pressure tubes. Then, the pressure tube diameter at the central portion of the reactor core is made smaller than that at the periphery of the reactor core. Further, injection means for gas such as helium is disposed to the upper portion for each of the heavy water tubes so that the level of the heavy water can easily be adjusted by the control for the gas pressure. Furthermore, heavy water reflection tubes are disposed around the reactor core. In this constitution, since the pitch for the pressure tubes can be increased, the construction and the maintenance for the nuclear reactor can be facilitated. Also, since the liquid surface of the heavy water in the heavy water tubes can be varied, nuclear properties is improved and the conversion ratio is improved. (Ikeda, J.)

Status report about the works for the start up of the RA-0 `zero power` nuclear reactor at the Cordoba National University; Estado actual de avance de las tareas para la nueva puesta en marcha del reactor nuclear RA-0 en la Universidad Nacional de Cordoba

Energy Technology Data Exchange (ETDEWEB)

Martin, H R; Carballido, C; Oliveras, T

1992-12-31

After two years of works at the Cordoba National University for the new start-up of the RA-0 `zero power` nuclear reactor, the results obtained are herein presented. Starting with practically null infrastructure at the beginning, specially in human resources and instrumentation of the reactor, the objectives can be considered satisfactory. The training in work of the future operational staff, the design and the construction of the instrumentation and the fitting of the installations are the principal items described in this paper. An special attention is devoted to the insertion of this type of installation in the university organization, usually not prepared for the quality and control activities, which is necessarily considered in these type of works. (Author). [Espanol] Luego de aproximadamente dos anos de trabajo para la nueva puesta en marcha del REACTOR NUCLEAR RA-0, se han alcanzado los resultados presentados en este trabajo. Partiendo de una infraestructura practicamente inexistente en cuanto a recursos humanos y estado de las instalaciones, los avances logrados son significativos. Comenzando por la capacitacion y el entrenamiento del futuro personal de operacion y pasando por la adecuacion de los equipos y componentes, hasta la confeccion de la documentacion mandatoria, se muestran los aspectos mas destacables de los trabajos realizados. Una atencion especial se dedica a la insercion de una instalacion de este tipo en el ambito universitario, el cual por sus particulares caracteristicas, ha debido ser tenido en cuenta permanentemente para la futura operacion de las instalaciones. (Autor).

Use of hafnium in control bars of nuclear reactors; Uso de hafnio en barras de control de reactores nucleares

Energy Technology Data Exchange (ETDEWEB)

Ramirez S, J.R.; Alonso V, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: jrrs@nuclear.inin-mx

2003-07-01

Recently the use of hafnium as neutron absorber material in nuclear reactors has been reason of investigation by virtue of that this material has nuclear properties as to the neutrons absorption and structural that can prolong the useful life of the control mechanisms of the nuclear reactors. In this work some of those more significant hafnium properties are presented like nuclear material. Also there are presented calculations carried out with the HELIOS code for fuel cells of uranium oxide and of uranium and plutonium mixed oxides under controlled conditions with conventional bars of boron carbide and also with similar bars to which are substituted the absorbent material by metallic hafnium, the results are presented in this work. (Author)

Situación del sector nuclear en España

Directory of Open Access Journals (Sweden)

Antonio González Jiménez

2010-12-01

Full Text Available España comenzó a interesarse por la energía nuclear a finales de los años cuarenta. En el año 1964 se aprobó la Ley de Energía Nuclear, y en el año 1972 se desarrolló el Reglamento de Instalaciones Nucleares y Radiactivas. Las actividades de seguridad y regulación se encomendaron al Consejo de Seguridad Nuclear, creado en 1980, y la investigación y gestión de los residuos radiactivos a la Empresa Nacional de Residuos Radiactivos (ENRESA, creada en 1985. La estructura industrial nuclear en España comenzó a desarrollarse en los años sesenta, como consecuencia de la decisión de construir las centrales nucleares de José Cabrera, Santa María de Garoña y Vandellós I. En la siguiente etapa, en la década de los años setenta, se construyeron las centrales de Almaraz, Ascó y Cofrentes. Durante la tercera etapa, en la década de los años ochenta, se construyeron las centrales de Vandellós II y Trillo I.Actualmente, la industria nuclear española, experta y eficaz, es garantía de que la tecnología nuclear se conserva en España no solo para apoyar a las centrales nucleares en operación, sino para atender un mercado nuclear reactivado a nivel internacional. España dispone de la infraestructura necesaria, la capacidad técnica, los recursos financieros y la voluntad de las empresas en el empeño común de proporcionar a los españoles una energía eléctrica fiable, barata y sostenible, con respeto al medio ambiente y seguridad para los ciudadanos. La energía nuclear es, en definitiva, una pieza clave hoy y lo será en el futuro.

Case study - Argentina

International Nuclear Information System (INIS)

Diaz, E.

1986-01-01

Antecedents and experience of nuclear activities in Argentina; the Atomic Energy Commission (CNEA). First development and research activities. Research reactors and radioisotopes plants. Health physics and safety regulations. - Feasibility studies for the first nuclear power plant. Awarding the first plant CNA I (Atucha I). Relevant data related to the different project stages. Plant performance. - Feasibility study for the second nuclear power plant. Awarding the second plant CNE (Central Nuclear Embalse). Relevant data related to established targets. Differences compared with the first station targets. Local participation. Plant performance. (orig./GL)

Nuclear research reactors

International Nuclear Information System (INIS)

1985-01-01

It's presented data about nuclear research reactors in the world, retrieved from the Sien (Nuclear and Energetic Information System) data bank. The information are organized in table forms as follows: research reactors by countries; research reactors by type; research reactors by fuel and research reactors by purpose. (E.G.) [pt

Completely automated nuclear reactors for long-term operation

International Nuclear Information System (INIS)

Teller, E.; Ishikawa, M.; Wood, L.

1996-01-01

The authors discuss new types of nuclear fission reactors optimized for the generation of high-temperature heat for exceedingly safe, economic, and long-duration electricity production in large, long-lived central power stations. These reactors are quite different in design, implementation and operation from conventional light-water-cooled and -moderated reactors (LWRs) currently in widespread use, which were scaled-up from submarine nuclear propulsion reactors. They feature an inexpensive initial fuel loading which lasts the entire 30-year design life of the power-plant. The reactor contains a core comprised of a nuclear ignitor and a nuclear burn-wave propagating region comprised of natural thorium or uranium, a pressure shell for coolant transport purposes, and automatic emergency heat-dumping means to obviate concerns regarding loss-of-coolant accidents during the plant's operational and post-operational life. These reactors are proposed to be situated in suitable environments at ∼100 meter depths underground, and their operation is completely automatic, with no moving parts and no human access during or after its operational lifetime, in order to avoid both error and misuse. The power plant's heat engine and electrical generator subsystems are located above-ground

Nuclear reactors. Introduction

International Nuclear Information System (INIS)

Boiron, P.

1997-01-01

This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

Design optimization of the Laguna Verde nuclear power station fuel recharge; Optimacion del diseno de recargas de combustible para la Central Laguna Verde

Energy Technology Data Exchange (ETDEWEB)

Cortes Campos, Carlos Cristobal [Instituto de Investigaciones Electricas, Cuernavaca (Mexico); Montes Tadeo, Jose Luis [Instituto Nacional de Investigaciones Nucleares (ININ), Salazar (Mexico)

1991-12-31

It is described, in general terms, the procedure that is followed to accomplish the optimization of the recharge design, and an example is shown where this procedure was applied for the analysis of the type BWR reactor of Unit No. 1 of the Laguna Verde Nuclear Power Station. [Espanol] Se describe en terminos generales el procedimiento que se sigue para realizar la optimacion del diseno de recargas, y se muestra un ejemplo en el que se utilizo dicho procedimiento para el analisis del reactor tipo BWR de la unidad 1, de la Central Laguna Verde (CLV).

Design optimization of the Laguna Verde nuclear power station fuel recharge; Optimacion del diseno de recargas de combustible para la Central Laguna Verde

Energy Technology Data Exchange (ETDEWEB)

Cortes Campos, Carlos Cristobal [Instituto de Investigaciones Electricas, Cuernavaca (Mexico); Montes Tadeo, Jose Luis [Instituto Nacional de Investigaciones Nucleares (ININ), Salazar (Mexico)

1992-12-31

It is described, in general terms, the procedure that is followed to accomplish the optimization of the recharge design, and an example is shown where this procedure was applied for the analysis of the type BWR reactor of Unit No. 1 of the Laguna Verde Nuclear Power Station. [Espanol] Se describe en terminos generales el procedimiento que se sigue para realizar la optimacion del diseno de recargas, y se muestra un ejemplo en el que se utilizo dicho procedimiento para el analisis del reactor tipo BWR de la unidad 1, de la Central Laguna Verde (CLV).

Safety Harmonization of Nuclear Reactors in Europe; La armonizacion de la seguridad de los reactores nucleares en Europa

Energy Technology Data Exchange (ETDEWEB)

Estevan Bolea, M. T.

2006-07-01

Dialogue and common tasks over the past few years have intensified among high level figures responsible for regulation organizations involved in nuclear installations from European countries as well as others. The best indicator of this is the harmonization of safety levels for european nuclear reactors. The CSN increasingly participates more intensely in three relevant international associations on these matters: INRA, FORO IBEROAMERICANO and WENDA. A detail study has been carried out by a WENRA work group whose results represent a large contribution made toward harmonization at reference levels on reactor safety issues, both from a legal viewpoint, and for their implantation in nuclear power plants. The President of the CSN explains in detail the most relevant aspects of this Final Report from WENRA. (Author)

Utilization of nuclear research reactors

International Nuclear Information System (INIS)

1980-01-01

prior to the beginning of the course was of particular value. Interesting scientific visits and demonstrations at the Isotope Institute and at the Central Research Institute for Physics (IFKI), both of the Hungarian Academy of Sciences, were also arranged. During the Study Tour at the Central Institute for Nuclear Research in Rossendorf near Dresden, German Democratic Republic, the participants had the opportunity to observe the organization of a 10 MW nuclear reactor where radioisotopes and radiopharmaceuticals are produced on a commercial scale. Lectures were delivered by local scientists on some of their programmes in applied research in solid state physics and material sciences. At the Technical University of Dresden, the group visited the homogeneous solid-moderated zero-power training reactor (AKR), primarily dedicated to nuclear education and training. Studies on different theoretical and experimental aspects of radiation protection (solid state nuclear track and thermoluminescent detectors) are also being carried out. The last day of the Study Tour was devoted to a visit to the College for Advanced Technology at Zittau, where a training reactor with a power of a few watts has been recently installed. (author)

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

Nuclear reactor theory

International Nuclear Information System (INIS)

Sekimoto, Hiroshi

2007-09-01

This textbook is composed of two parts. Part 1 'Elements of Nuclear Reactor Theory' is composed of only elements but the main resource for the lecture of nuclear reactor theory, and should be studied as common knowledge. Much space is therefore devoted to the history of nuclear energy production and to nuclear physics, and the material focuses on the principles of energy production in nuclear reactors. However, considering the heavy workload of students, these subjects are presented concisely, allowing students to read quickly through this textbook. (J.P.N.)

Improved nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

Power Nuclear Reactors: technology and innovation for development in future; Centrales Nucleares de Potencia: tecnologias actuales e innovaciones para el futuro

Energy Technology Data Exchange (ETDEWEB)

Suarez Antola, R [Universidad Catolica del Uruguay, Montevideo(Uruguay); Ministerio de Industria Energia y Minerria, Montevideo(Uruguay)

2009-07-01

The conference is about some historicals task of the fission technology as well as many types of Nuclear Reactors. Enrichment of fuel, wastes, research reactors and power reactors, a brief advertisment about Uruguay electric siystem and power generation, energetic worldwide, proliferation, safety reactors, incidents, accidents, Three-Mile Island accident, Chernobil accident, damages, risks, classification and description of Power reactors steam generation, nuclear reactor cooling systems, future view.

Extending the Candu Nuclear Reactor Concept: The Multi-Spectrum Nuclear Reactor

International Nuclear Information System (INIS)

Allen, Francis; Bonin, Hugues

2008-01-01

The aim of this work is to examine the multi-spectrum nuclear reactor concept as an alternative to fast reactors and accelerator-driven systems for breeding fissile material and reducing the radiotoxicity of spent nuclear fuel. The design characteristics of the CANDU TM nuclear power reactor are shown to provide a basis for a novel approach to this concept. (authors)

Extending the Candu Nuclear Reactor Concept: The Multi-Spectrum Nuclear Reactor

Energy Technology Data Exchange (ETDEWEB)

Allen, Francis [Director General Nuclear Safety, 280 Slater St, Ottawa, K1A OK2 (Canada); Bonin, Hugues [Royal Military College of Canada, 11 General Crerar Cres, Kingston, K7K 7B4 (Canada)

2008-07-01

The aim of this work is to examine the multi-spectrum nuclear reactor concept as an alternative to fast reactors and accelerator-driven systems for breeding fissile material and reducing the radiotoxicity of spent nuclear fuel. The design characteristics of the CANDU{sup TM} nuclear power reactor are shown to provide a basis for a novel approach to this concept. (authors)

Nuclear reactor instrumentation power monitor

International Nuclear Information System (INIS)

Suzuki, Shigeru.

1989-01-01

The present invention concerns a nuclear reactor instrumentation power monitor that can be used in, for example, BWR type nuclear power plants. Signals from multi-channel detectors disposed on field units are converted respectively by LPRM signal circuits. Then, the converted signals are further converted by a multiplexer into digital signals and transmitted as serial data to a central monitor unit. The thus transmitted serial data are converted into parallel data in the signal processing section of the central monitor unit. Then, LPRM signals are taken out from each of channel detectors to conduct mathematical processing such as trip judgment or averaging. Accordingly, the field unit and the central monitor unit can be connected by way of only one data transmission cable thereby enabling to reduce the number of cables. Further, since the data are transmitted on digital form, it less undergoes effect of noises. (I.S.)

Design of the reactor coolant system and associated systems in nuclear power plants. Safety guide (Spanish Edition); Diseno del sistema de refrigeracion del reactor y los sistemas asociados en las centrales nucleares. Guia de seguridad

Energy Technology Data Exchange (ETDEWEB)

NONE

2010-07-15

This Safety Guide was prepared under the IAEA programme for establishing safety standards for nuclear power plants. The basic requirements for the design of safety systems for nuclear power plants are established in the Safety Requirements publication, Safety Standards Series No. NS-R-1 on Safety of Nuclear Power Plants: Design, which it supplements. This Safety Guide describes how the requirements for the design of the reactor coolant system (RCS) and associated systems in nuclear power plants should be met. This publication is a revision and combination of two previous Safety Guides, Safety Series No. 50-SG-D6 on Ultimate Heat Sink and Directly Associated Heat Transport Systems for Nuclear Power Plants (1982), and Safety Series No. 50-SG-D13 on Reactor Coolant and Associated Systems in Nuclear Power Plants (1987), which are superseded by this new Safety Guide. The revision takes account of developments in the design of the RCS and associated systems in nuclear power plants since the earlier Safety Guides were published in 1982 and 1987, respectively. The other objectives of the revision are to ensure consistency with Ref., issued in 2004, and to update the technical content. In addition, an appendix on pressurized heavy water reactors (PHWRs) has been included.

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.

Nuclear safety. Concerns about the nuclear power reactors in Cuba

International Nuclear Information System (INIS)

Wells, Jim; Aloise, Gene; Flaherty, Thomas J.; Fitzgerald, Duane; Zavala, Mario; Hayward, Mary Alice

1992-09-01

In 1976, the Soviet Union and Cuba concluded an agreement to construct two 440-megawatt nuclear power reactors near Cienfuegos on the south central coast of Cuba, about 180 miles south of Key West, Florida. The construction of these reactors, which began around 1983, was a high priority for Cuba because of its heavy dependence on imported oil. Cuba is estimated to need an electrical generation capacity of 3,000 megawatts by the end of the decade. When completed, the first reactor unit would provide a significant percentage (estimated at over 15 percent) of Cuba's need for electricity. It is uncertain when Cuba's nuclear power reactors will become operational. On September 5, 1992, Fidel Castro announced the suspension of construction at both of Cuba's reactors because Cuba could not meet the financial terms set by the Russian government to complete the reactors. Cuban officials had initially planned to start up the first of the two nuclear reactors by the end of 1993. However, before the September 5 announcement, it was estimated that this reactor would not be operational until late 1995 or early 1996. The civil construction (such as floors and walls) of the first reactor is currently estimated to be about 90 percent to 97 percent complete, but only about 37 percent of the reactor equipment (such as pipes, pumps, and motors) has been installed. The civil construction of the second reactor is about 20 percent to 30 percent complete. No information was available about the status of equipment for the second reactor. According to former Cuban nuclear power and electrical engineers and a technician, all of whom worked at the reactor site and have recently emigrated from Cuba, Cuba's nuclear power program suffers from poor construction practices and inadequate training for future reactor operators. One former official has alleged, for example, that the first reactor containment structure, which is designed to prevent the accidental release of radioactive material into

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

On application of CFD codes to problems of nuclear reactor safety

International Nuclear Information System (INIS)

Muehlbauer, Petr

2005-01-01

The 'Exploratory Meeting of Experts to Define an Action Plan on the Application of Computational Fluid Dynamics (CFD) Codes to Nuclear Reactor Safety Problems' held in May 2002 at Aix-en-Province, France, recommended formation of writing groups to report the need of guidelines for use and assessment of CFD in single-phase nuclear reactor safety problems, and on recommended extensions to CFD codes to meet the needs of two-phase problems in nuclear reactor safety. This recommendations was supported also by Working Group on the Analysis and Management of Accidents and led to formation oaf three Writing Groups. The first writing Group prepared a summary of existing best practice guidelines for single phase CFD analysis and made a recommendation on the need for nuclear reactor safety specific guidelines. The second Writing Group selected those nuclear reactor safety applications for which understanding requires or is significantly enhanced by single-phase CFD analysis, and proposed a methodology for establishing assesment matrices relevant to nuclear reactor safety applications. The third writing group performed a classification of nuclear reactor safety problems where extension of CFD to two-phase flow may bring real benefit, a classification of different modeling approaches, and specification and analysis of needs in terms of physical and numerical assessments. This presentation provides a review of these activities with the most important conclusions and recommendations (Authors)

Proceedings of the 2000 International Conference on Nuclear Energy in Central Europe

International Nuclear Information System (INIS)

Mavko, B.; Cizelj, L.; Kovac, M.

2000-01-01

International Conference Nuclear Energy in Central Europe is an annual meeting of the Nuclear Society of Slovenia. The proceedings contain 108 articles from Slovenia, surrounding countries and countries of the Central and Eastern European Region. Topics are: thermal hydraulics, severe accidents, probabilistic safety assessment (PSA), nuclear waste, safety analyses, nuclear power plant operation, structural integrity and aging, nuclear energy and public, other related topics, research reactors, education and training and Monte Carlo transport calculations

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

Nuclear reaction data and nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Paver, N [University of Trieste (Italy); Herman, M [International Atomic Energy Agency, Vienna (Austria); Gandini, A [ENEA, Rome (Italy)

2001-12-15

These two volumes contain the lecture notes of the workshop 'Nuclear Reaction Data and Nuclear Reactors: Physics, Design and Safety', which was held at the Abdus Salam ICTP in the Spring of 2000. The workshop consisted of five weeks of lecture courses followed by practical computer exercises on nuclear data treatment and design of nuclear power systems. The spectrum of topics is wide enough to timely cover the state-of-the-art and the perspectives of this broad field. The first two weeks were devoted to nuclear reaction models and nuclear data evaluation. Nuclear data processing for applications to reactor calculations was the subject of the third week. On the last two weeks reactor physics and on-going projects in nuclear power generation, waste disposal and safety were presented.

Explicatory Dictionary for Exact Sciences. Nuclear Energy, EN2. Nuclear Power. Romanian/English/French

International Nuclear Information System (INIS)

Dragan, Gleb; Rapeanu, S.N.; Comsa, Olivia

2002-01-01

The explicative dictionary for nuclear power, accomplished in the frame of the Commission for Scientific Terminology of the Romanian Academy, represents the second issue in a series running from Nuclear Energy EN 1 through Nuclear Energy EN 10 covering the following fields: EN 1. General terminology; EN 2. Nuclear power; EN 3. Physical protection and nuclear safeguards; EN 4. Nuclear fuel cycle; EN 5. Radioactive wastes; EN 6. Safety of nuclear facilities and materials; EN 7. Radioprotection and dosimetry; EN 8. Nuclear reactors; EN 9. Nuclear sciences and engineering; EN 10. Nuclear law and legislation. The main body of the dictionary's contents was selected by specialists working with the Center of Technology and Engineering for Nuclear Projects - CITON, based on their experience of more than 20 years in introducing and implementing nuclear power in Romania, as well as, on collaboration with nuclear physics and engineering research institutes and physics departments of Romanian universities. The project of a nuclear dictionary in 10 issues aims at supporting the program of nuclear power development in Romania and is at the same time part of nuclear knowledge management policy boosted by IAEA which encourages publication of informative materials highly specialized but also accessible to the public at large. The project aims also to establish the Romanian standardized terminology in the nuclear domain as much in line as possible with the terminologies of the largest communities worldwide most active in nuclear science and technology. Under the guidance of continuos build-up and evolution of nuclear knowledge the present work is intended to be upgraded permanently. The explanation of the terms was based on SR ISO standards, terminology adopted by Organization for Economic and Cooperation Development, OECD/NEA, and IAEA. This series is targeting translators, specialists, students, and the public at large

Determination of activity by gamma spectrometry of radionuclides present in drums of residues generated in nuclear centrals

International Nuclear Information System (INIS)

Aguiar, J.C.; Fernandez, J.

2006-01-01

The generation of radioactive residuals in nuclear centrals as CNA I (Atucha I Nuclear Central) and CNE (Embalse Nuclear Central) makes that the measurement of those radionuclides has been a previous stage to the waste management. A method used in those nuclear centrals it is the gamma spectrometry with HPGe detectors, previous to the immobilization of the residual in a cemented matrix, with this the contact with the external agents and its possible dispersion to the atmosphere in the short term is avoided. The ARN (Nuclear Regulatory Authority) of Argentina it carries out periodically intercomparisons and evaluations of the measurement and procedures systems used in the nuclear power stations for the correct measurement and determination of activity of radioactive residuals by gamma spectrometry. In this work an independent method of measurement is exposed to the nuclear power stations. To determine the activity of the residuals by gamma spectrometry deposited in drums, it is required of the precise knowledge of the efficiency curve for such geometry and matrix. Due to the RNA doesn't have a pattern of these characteristics, a mathematical model has been used to obtain this efficiency curve. For it, it is necessary to determine previously: 1) the geometric efficiency or solid angle sustained by the source-detector system (drum-detector) applying a mathematical model described in this work. 2) To estimate the auto-attenuation factor that present the photons in the cemented matrix, these calculations are carried out with a simple equation and its are verified with the Micro Shield 6.10 program. The container commonly used by these nuclear power stations its are drums for 220 liters constructed with SAE 1010 steel and with a thickness of 0.127 cm, with an approximate weight 7.73 Kg., internal diameter of 57.1 cm, and height: 87 cm. The results obtained until the moment register a discrepancy from 5 to 10% with relationship to the measurements carried out by the

Diseño y evaluación de un procedimiento on-line para la biorremediación de aguas radiactivas en centrales nucleares

OpenAIRE

Belinchón Vergara, Julio Angel

2016-01-01

Recientemente se ha demostrado la existencia de microorganismos en las piscinas de almacenamiento de combustible nuclear gastado en las centrales nucleares utilizando técnicas convencionales de cultivo en el laboratorio. Estudios posteriores han puesto de manifiesto que los microorganismos presentes eran capaces de colonizar las paredes de acero inoxidable de las piscinas formando biopelículas. Adicionalmente se ha observado la capacidad de estas biopelículas de retener radionúclidos, lo que ...

Nuclear power reactors

International Nuclear Information System (INIS)

1982-11-01

After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

Nuclear reactor neutron shielding

Science.gov (United States)

Speaker, Daniel P; Neeley, Gary W; Inman, James B

2017-09-12

A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

Molten salt reactors and possible scenarios for future nuclear power deployment

International Nuclear Information System (INIS)

Merle-Lucotte, E.; Mathieu, L.; Heuer, D.; Loiseaux, J. M.; Billebaud, A.; Brissot, R.; David, S.; Garzenne, C.; Laulan, O.; Le Brun, C.; Lecarpentier, D.; Liatard, E.; Meplan, O.; Michel-Sendis, F.; Nuttin, A.; Perdu, F.

2004-01-01

An important fraction of the nature energy demand may be satisfied by nuclear power. In this context, the possibilities of worldwide nuclear deployment are studied. We are convinced that the Molten Salt Reactors may play a central role in this deployment. The Molten Salt Reactor needs to be coupled to a reprocessing unit in order to extract the Fission Products which poison the core. The efficiency of this reprocessing has a crucial influence on reactor behavior especially for the breeding ratio. The Molten Salt Breeder Reactor project was based on an intensive reprocessing for high breeding purposes. A new concept of Thorium Molten Salt Reactor is presented here. Including this new concept in the worldwide nuclear deployment, to satisfy these power needs, we consider three typical scenarios, based on three reactor types: Pressurized Water Reactor, Fast Neutron Reactor and Thorium Molten Salt Reactor. The aim of this paper is to demonstrate, in a first hand that a Thorium Molten Salt Reactor can be realistic, with correct temperature coefficients and at least iso-breeder with slow reprocessing and new geometry; on the other hand that such Molten Salt Reactors enable a successful nuclear deployment, while minimizing fuel and waste management problems. (authors)

Nuclear research reactors in Brazil

Energy Technology Data Exchange (ETDEWEB)

Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2011-07-01

The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

Nuclear research reactors in Brazil

International Nuclear Information System (INIS)

Cota, Anna Paula Leite; Mesquita, Amir Zacarias

2011-01-01

The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

Nuclear reactor coolant channels

International Nuclear Information System (INIS)

Macbeth, R.V.

1978-01-01

A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

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)

Twenty years of Radiology in RP-10 nuclear reactor protection; Veinte anos de proteccion radiologica en el reactor nuclear RP-10

Energy Technology Data Exchange (ETDEWEB)

Zapata, Alejandro L.; Ramos, Fernando T.; Arrieta, Rolando W.B.; Vela Mora, Mariano, E-mail: lzapata@ipen.gob.pe, E-mail: framos@ipen.gob.pe, E-mail: rarrieta@ipen.gob.pe, E-mail: mvela@ipen.gob.pe [Instituto Peruano de Energia Nuclear (IPEN), Lima (Peru)

2013-07-01

In this report we present the results about radiation controls during 1990 - 2010, carried out in the Nuclear Reactor RP-10 of the Nuclear Center of Huarangal. These controls and radiological evaluation are of much utility for the radio personnel protection of this one and other reactors, since it allows to compares these variables with respect to the time. From the results obtained in monitoring and radiation controls, we conclude that in no case it has been reached the limits allowed by the Peruvian Regulating Authority. (author)

Physics of nuclear reactors

International Nuclear Information System (INIS)

Baeten, Peter

2006-01-01

This course gives an introduction to Nuclear Reactor Physics. The first chapter explains the most important parameters and concepts in nuclear reactor physics such as fission, cross sections and the effective multiplication factor. Further on, in the second chapter, the flux distributions in a stationary reactor are derived from the diffusion equation. Reactor kinetics, reactor control and reactor dynamics (feedback effects) are described in the following three chapters. The course concludes with a short description of the different types of existing and future reactors. (author)

La educación de la tecnología de los reactores nucleares en un programa de postgrado de Ingeniería Multidisciplinar

OpenAIRE

Mesquita, Amir Zacarias; Lameiras, Fernando Soares; Martins, Maximiliano Delany

2016-01-01

En este artículo se describen las actividades desarrolladas y las innovaciones realizadas en la línea de investigación en los reactores nucleares en un programa de posgrado multidisciplinario, el Posgrado Programa de CDTN/CNEN, en Belo Horizonte. El programa cuenta con maestría y doctorado (stricto sensu), cubriendo las áreas de la ciencia y la ingeniería de materiales, física y química de la materia condensada, la física y la biología de la salud, ciencias de la tierra y la tecnología minera...

Communication system and spectral analysis for Ge-Li and GeHp detectors; Sistema de comunicaciones y analisis de espectros para detectores Ge-Li y GeHp

Energy Technology Data Exchange (ETDEWEB)

Fernandez, J; Castano, P; Bonino, A D; Righetti, M A

1991-12-31

An integral communication and spectral analysis system (SICADE) was developed and implemented to satisfy the need to optimize and automate the measurement system used in Atucha I nuclear power plant for the activity in the primary loop`s water extracted by the TV system. The importance of these measurements is based on the fact that from the spectrometric analysis of the samples extracted, the Iodines-GN and Iodines-Iodines relations, which allow to detect the presence of deficient fuel elements, are calculated. The system developed is based on two modules integrated in a unique set commanded by the operators through the screen dialogue. (Author). [Espanol] Dada la necesidad de optimizar y automatizar el sistema de medicion utilizado en la Central Nuclear Atucha I para la actividad en el agua del circuito primario extraida a traves del sistema TV se desarrollo e implemento un sistema integral de Comunicacion y Analisis de espectros (SICADE). La importancia de estas mediciones radica en el hecho que del analisis espectrometrico de las muestras extraidas se calculan las relaciones entre Iodos-GN e Iodos-Iodos que permiten detectar la presencia de elementos combustibles fallados. El sistema desarrollado esta basado en dos modulos integrados en un unico conjunto comandado por el operador a traves del dialogo por pantalla. (Autor).

The First Two Years of Operating Experience of the Kahl Nuclear Power Station; Experience acquise pendant les deux premieres annees de fonctionnement de la centrale nucleaire de Kahl; Opyt pervykh dvukh let ehkspluatatsii atomnoj ehlektrostantsii v Kale; Experiencia adquirida en los primeros cuatro anos de funcionamiento de la central nucleoelectrica de Kahl

Energy Technology Data Exchange (ETDEWEB)

Bruchner, H. J. [Aeg-Kernenergieanlagen, Frankfurt-am-Main (Germany); Weckesser, A. [Versuchs-Atomkraftwerk Kahl Gmbh, Kahl (Germany)

1963-10-15

Kahl, being the first European private atomic power station, has been operating at load since June 1961. Kahl is equipped with a boiling-water reactor, operating in an indirect cycle and with natural circulation. Its electrical net capacity is 15 MW, and the power produced until February 1963 amounts to 140 million kWh. In the paper the operating experience, particularly the extensive testing programme such as concerning transient behaviour and gammascanning, will be reviewed. Results about the operating performance of certain plant components such as control-rod-drive system, off-gas system and turbine will be presented. After this testing programme the plant has been operated at base load for some time in order to obtain realistic experience on fuel performance. After completion of the base-load phase of operation a nuclear superheating test loop will be installed and run in the Kahl reactor. (author) [French] La centrale de Kahl, la premiere en Europe dont l'exploitation soit assuree par une entreprise privee, alimente le secteur depuis juin 1961. Elle est equipee d'un reacteur a eau bouillante, en cycle indirect, a circulation naturelle. Avec une puissance electrique nette de 15 MW, elle avait produit, jusqu'a fevrier 1963, 140 millions de kWh. Le memoire rend compte de l'experience acquise concernant le fonctionnement de la centrale, notamment du programme etendu des essais effectues sur son comportement en regime transitoire et sur la detection gamma. Il donne, en particulier, des renseignements sur le fonctionnement de certains elements de la centrale tels que le mecanisme de commande des barres de controle, le dispositif d'echappement des gaz et la turbine. A la suite de ce programme d'essais, on fait fonctionner depuis quelque temps la centrale comme centrale de base afin de rassembler des donnees pratiques sur les performances du combustible. A l'expiration de la periode prevue pour le fonctionnement comme centrale de base, on amenagera, dans le reacteur

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

Nuclear reactor buildings

International Nuclear Information System (INIS)

Nagashima, Shoji; Kato, Ryoichi.

1985-01-01

Purpose: To reduce the cost of reactor buildings and satisfy the severe seismic demands in tank type FBR type reactors. Constitution: In usual nuclear reactor buildings of a flat bottom embedding structure, the flat bottom is entirely embedded into the rock below the soils down to the deck level of the nuclear reactor. As a result, although the weight of the seismic structure can be decreased, the amount of excavating the cavity is significantly increased to inevitably increase the plant construction cost. Cross-like intersecting foundation mats are embedded to the building rock into a thickness capable withstanding to earthquakes while maintaining the arrangement of equipments around the reactor core in the nuclear buildings required by the system design, such as vertical relationship between the equipments, fuel exchange systems and sponteneous drainings. Since the rock is hard and less deformable, the rigidity of the walls and the support structures of the reactor buildings can be increased by the embedding into the rock substrate and floor responsivity can be reduced. This enables to reduce the cost and increasing the seismic proofness. (Kamimura, M.)

Research reactors in Argentina

International Nuclear Information System (INIS)

Carlos Ruben Calabrese

1999-01-01

Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

Reinforcement of planning and response to emergencies inside nuclear power plants; Refuerzo de la planificación y respuesta ante emergencias en el interior de las centrales nucleares

Energy Technology Data Exchange (ETDEWEB)

Cortés, M.A.; Garcés, J.M.; Bolaños, J.; Fernández, J.; Gutiérrez, C.; Pontejo, A.; Rosell, B.

2015-07-01

Fukushima Daichi nuclear accident (11th march of 2011) required a rapid and effective response of nuclear industry, in order to guarantee that current safety margins were enough to cope with events like this or even worse. In Europe, stress tests were carried out to comply with that objective, evaluating if every nuclear plant was able to succeed managing design basis accidents or even beyond. Stress-test results were satisfactory, although several areas for improvement where identified. Improvements in emergency response and management have been developed in several phases. First, increasing emergency response capacity by implementing modifications in existing equipment. Later, installing new equipment (generally portable) which would be used in even more degraded conditions. This must be held by the revision of emergency response personnel and their training. Finally, external support has been reinforced to help in longstanding emergencies, with the establishment of CAE (Regional Center), and collaboration protocols with the Army. These actions, supervised and approved by the Regulator Body (CSN) will lead to increase plants safety margins and their emergency response capacities in order to reduce probability and consequences of a nuclear accident to the minimum. [Spanish] El accidente nuclear ocurrido el 11 de marzo de 2011 en Fukushima Daichi requirió de una pronta y efectiva respuesta de la industria nuclear, que garantizase que los márgenes de seguridad eran los adecuados para hacer frente a sucesos de estas características o incluso peores. Con ese objetivo se planteó, a nivel europeo, la realización de unas pruebas de resistencia que valorasen si las centrales podían hacer frente con garantías a una serie de accidentes dentro y más allá de las bases de diseño. Los resultados fueron satisfactorios, si bien se identificaron varios aspectos de mejora. La mejora en la respuesta y gestión de emergencias de las centrales nucleares se ha planteado en

Da Lat Nuclear Research Reactor. Role and perspective in the development of radioisotope and nuclear technique application in Vietnam

International Nuclear Information System (INIS)

Tran Ha Anh; Tran Khac An; Ngo Phu Khang; Nguyen Mong Sinh

1995-01-01

The Da Lat Nuclear Research Reactor is playing a central role in the development of both the Nuclear Research Institute and nuclear application in our country. Thanks to this main scientific tool, the Nuclear Research Institute nearly 10 years after the completion of its renovation from the previous American-made TRIGA MARK 2 reactor is being able to implement numerous scientific and technological research projects and to develop significant applications of radioisotopes and various nuclear techniques. A general overview of the research and development activities of the Institute based on the Da Lat Nuclear Research Reactor is given as well as those aiming at ensuring its safe, reliable and efficient operation and at enlarging the perspectives of its utilisation in the future. (authors). 5 refs., 1 fig., 1 tab

Present status of space nuclear reactor

International Nuclear Information System (INIS)

Kaneko, Yoshihiko

1996-01-01

USA and former USSR led space development, and had the experience of launching nuclear reactor satellites. In USA, the research and development of space nuclear reactor were advanced mainly by NASA, and in 1965, the nuclear reactor for power source ''SNAP-10A'' was launched and put on the orbit around the earth. Thereafter, the reactor was started up, and the verifying test at 500 We was successfully carried out. Also for developing the reactor for thermal propulsion, NERVA/ROVER project was done till 1973, and the technological basis was established. The space Exploration Initiative for sending mankind to other solar system planets than the earth is the essential point of the future projects. In former USSR, the ground experiment of the reactor for 800 We power source ''Romashka'', the development of the reactor for 10 kWe power source ''Topaz-1 and 2'', the flight of the artificial satellites, Cosmos 954 and Cosmos 1900, on which nuclear reactors were mounted, and the operation of 33 ocean-monitoring satellites ''RORSAT'' using small fast reactors were carried out. The mission of space development and the nuclear reactors as power source, the engineering of space nuclear reactors, the present status and the trend of space nuclear reactor development, and the investigation by the UN working group on the safety problem of space nuclear reactors are described. (K.I.)

Status of the nuclear sector in Spain Situación del sector nuclear en España

Directory of Open Access Journals (Sweden)

Antonio González Jiménez

2010-12-01

ó a desarrollarse en los años sesenta, como consecuencia de la decisión de construir las centrales nucleares de José Cabrera, Santa María de Garoña y Vandellós I. En la siguiente etapa, en la década de los años setenta, se construyeron las centrales de Almaraz, Ascó y Cofrentes. Durante la tercera etapa, en la década de los años ochenta, se construyeron las centrales de Vandellós II y Trillo I. Actualmente, la industria nuclear española, experta y eficaz, es garantía de que la tecnología nuclear se conserva en España no solo para apoyar a las centrales nucleares en operación, sino para atender un mercado nuclear reactivado a nivel internacional. España dispone de la infraestructura necesaria, la capacidad técnica, los recursos financieros y la voluntad de las empresas en el empeño común de proporcionar a los españoles una energía eléctrica fiable, barata y sostenible, con respeto al medio ambiente y seguridad para los ciudadanos. La energía nuclear es, en definitiva, una pieza clave hoy y lo será en el futuro.

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

Fast reactors in nuclear power

Energy Technology Data Exchange (ETDEWEB)

Kazachkovskii, O

1981-02-01

The possible applications are discussed of fast reactor nuclear power plants. Basic differences are explained in fast and thermal reactors, mainly with a view to nuclear fuel utilization. Discussed in more detail are the problems of nuclear fuel reproduction and the nost important technical problems of fast reactors. Flow charts are shown of heat transfer for fast reactors BN-350 (loop design) and BN-600 (integral coolant circuit design). Main specifications are given for demonstration and power fast reactors in operation, under construction and in project-stage.

Nuclear reactors; graphical symbols

International Nuclear Information System (INIS)

1987-11-01

This standard contains graphical symbols that reveal the type of nuclear reactor and is used to design graphical and technical presentations. Distinguishing features for nuclear reactors are laid down in graphical symbols. (orig.) [de

Guidebook to nuclear reactors

International Nuclear Information System (INIS)

Nero, A.V. Jr.

1976-05-01

A general introduction to reactor physics and theory is followed by descriptions of commercial nuclear reactor types. Future directions for nuclear power are also discussed. The technical level of the material is suitable for laymen

Nuclear power plant with several reactors

Energy Technology Data Exchange (ETDEWEB)

Grishanin, E I; Ilyunin, V G; Kuznetsov, I A; Murogov, V M; Shmelev, A N

1972-05-10

A design of a nuclear power plant suggested involves several reactors consequently transmitting heat to a gaseous coolant in the joint thermodynamical circuit. In order to increase the power and the rate of fuel reproduction the low temperature section of the thermodynamical circuit involves a fast nuclear reactor, whereas a thermal nuclear reactor is employed in the high temperature section of the circuit for intermediate heating and for over-heating of the working body. Between the fast nuclear and the thermal nuclear reactors there is a turbine providing for the necessary ratio between pressures in the reactors. Each reactor may employ its own coolant.

Nuclear reactor safety research in Idaho

International Nuclear Information System (INIS)

Zeile, H.J.

1983-01-01

Detailed information about the performance of nuclear reactor systems, and especially about the nuclear fuel, is vital in determining the consequences of a reactor accident. Fission products released from the fuel during accidents are the ultimate safety concern to the general public living in the vicinity of a nuclear reactor plant. Safety research conducted at the Idaho National Engineering Laboratory (INEL) in support of the U.S. Nuclear Regulatory Commission (NRC) has provided the NRC with detailed data relating to most of the postulated nuclear reactor accidents. Engineers and scientists at the INEL are now in the process of gathering data related to the most severe nuclear reactor accident - the core melt accident. This paper describes the focus of the nuclear reactor safety research at the INEL. The key results expected from the severe core damage safety research program are discussed

Nuclear reactor engineering: Reactor design basics. Fourth edition, Volume One

International Nuclear Information System (INIS)

Glasstone, S.; Sesonske, A.

1994-01-01

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in design and operation of nuclear power plants. Extensively updated, the fourth edition includes new material on reactor safety and risk analysis, regulation, fuel management, waste management, and operational aspects of nuclear power. This volume contains the following: energy from nuclear fission; nuclear reactions and radiations; neutron transport; nuclear design basics; nuclear reactor kinetics and control; radiation protection and shielding; and reactor materials

Professional practices: a short introduction of national nuclear activities to university students; Practicas profesionales: breve incursion laboral de alumnos universitarios en las actividades nucleares

Energy Technology Data Exchange (ETDEWEB)

Martin, Hugo R [Comision Nacional de Energia Atomica, Cordoba(Argentina). Regional Centro

2005-07-01

tiene a cargo la produccion del Oxido de Uranio para el combustible de las centrales nucleares en Atucha 1 y Embalse; las cuestiones relacionadas con la opinion publica han sido una constante preocupacion para la institucion. Por tal motivo, a partir de junio de 2003 se comenzo a desarrollar un Programa de Apertura Institucional (PAI), con la intencion de poner al alcance de la sociedad cordobesa un conocimiento mas profundo sobre las actividades que se desarrollan en el lugar. Entre las multiples acciones implementadas se encuentra el desarrollo de relaciones institucionales con las universidades locales, con el objeto de permitir a alumnos, docentes e investigadores, segun los intereses de cada uno, un acercamiento a la realidad del sector nuclear nacional. A partir de entonces la CNEA ha comenzado a participar activamente de las actividades de la comunidad universitaria local, tanto como miembro permanente en comisiones evaluadoras para el otorgamiento de becas y subsidios del sistema cientifico local, como en consejos asesores del gobierno provincial y otras instancias similares. De las actividades desarrolladas, la que ha mostrado ser muy efectiva para los objetivos propuestos y que motiva estas lineas, ha sido la realizacion de Practicas Profesionales de alumnos del ultimo anio de distintas carreras universitarias en instalaciones de la institucion. (autor)

Nuclear reactor engineering: Reactor systems engineering. Fourth edition, Volume Two

International Nuclear Information System (INIS)

Glasstone, S.; Sesonske, A.

1994-01-01

This new edition of this classic reference combines broad yet in-depth coverage of nuclear engineering principles with practical descriptions of their application in the design and operation of nuclear power plants. Extensively updated, the fourth edition includes new materials on reactor safety and risk analysis, regulation, fuel management, waste management and operational aspects of nuclear power. This volume contains the following: the systems concept, design decisions, and information tools; energy transport; reactor fuel management and energy cost considerations; environmental effects of nuclear power and waste management; nuclear reactor safety and regulation; power reactor systems; plant operations; and advanced plants and the future

Nuclear reactor

International Nuclear Information System (INIS)

Batheja, P.; Huber, R.; Rau, P.

1985-01-01

Particularly for nuclear reactors of small output, the reactor pressure vessel contains at least two heat exchangers, which have coolant flowing through them in a circuit through the reactor core. The circuit of at least one heat exchanger is controlled by a slide valve, so that even for low drive forces, particularly in natural circulation, the required even loading of the heat exchanger is possible. (orig./HP) [de

High-temperature and breeder reactors - economic nuclear reactors of the future

International Nuclear Information System (INIS)

Djalilzadeh, A.M.

1977-01-01

The thesis begins with a review of the theory of nuclear fission and sections on the basic technology of nuclear reactors and the development of the first generation of gas-cooled reactors applied to electricity generation. It then deals in some detail with currently available and suggested types of high temperature reactor and with some related subsidiary issues such as the coupling of different reactor systems and various schemes for combining nuclear reactors with chemical processes (hydrogenation, hydrogen production, etc.), going on to discuss breeder reactors and their application. Further sections deal with questions of cost, comparison of nuclear with coal- and oil-fired stations, system analysis of reactor systems and the effect of nuclear generation on electricity supply. (C.J.O.G.)

Nuclear Reactor Sharing Program

International Nuclear Information System (INIS)

1994-01-01

The Ohio State University Research Reactor (OSURR) is licensed to operate at a maximum power level of 500 kW. A pool-type reactor using flat-plate, low enriched fuel elements, the OSURR provides several experimental facilities including two 6-inch i.d. beam ports, a graphite thermal column, several graphite-isotope-irradiation elements, a pneumatic transfer system (Rabbit), various dry tubes, and a Central Irradiation Facility (CIF). The core arrangement and accessibility facilitates research programs involving material activation or core parameter studies. The OSURR control room is large enough to accommodate laboratory groups which can use control instrumentation for monitoring of experiments. The control instrumentation is relatively simple, without a large amount of duplication. This facilitates opportunities for hands-on experience in reactor operation by nuclear engineering students making reactor parameter measurements. For neutron activation analysis and analyses of natural environmental radioactivity, the NRL maintains the gamma ray spectroscopy system (GRSS). It is comprised of two PC-based 8192-channel multichannel analyzers (MCAs) with all the required software for quantitative analysis. A 3 double-prime x 3 double-prime NaI(Tl), a 14 percent Ge(Li), and a High Purity Germanium detector are currently available for use with the spectroscopy system

Core reactor simulation of the Central Laguna Verde (CLV) reactor in stationary state and an example of the application in the recharge options analysis of cycle 3; Simulacion del nucleo del reactor de la Central Laguna Verde (CLV) en estado estacionario y ejemplo de aplicacion en el analisis de alternativas de recarga del ciclo 3

Energy Technology Data Exchange (ETDEWEB)

Ocampo Mansilla, Hector; Francois Lacouture, Juan Luis; Blanco Lara, Jesus; Cortes Campos, Carlos Cristobal; Esquivias Montoya, Jesus; Esquivel Torres, Jose Luis; Martin del Campo Marquez, Cecilia [Instituto de Investigaciones Electricas, Cuernavaca (Mexico); Montes Tadeo, Jose Luis [Instituto Nacional de Investigaciones Nucleares (ININ), Salazar (Mexico); Sanchez Herrera, Luciano; Torres Alvarez, Carlos [Comision Federal de Electricidad (CFE), Mexico, D. F. (Mexico)

1991-12-31

The results are presented of a study requested by Comision Federal de Electricidad (CFE) for the analysis of Cycle 3, of Unit No. 1 of the Laguna Verde Nuclear Power Station (CLNV) and determine the burning effect impact, carried out with the starting tests and the operation of Cycles 1 and 2 on base of the cycle extension known as coastdown. The calculations were realized with the Code Package FMS for fuel managing, using the Code PRESTO-B that analyzes the reactor in detailed form in three dimensions an in stationary state. In the study the schemes of fraction of recharge proposed by General Electric (GE) were analyzed with the effect of cycle extension. The initial design value of 100 assemblies for Cycle 3, GE proposes to increase such fraction from 112 to 120 assemblies. This impacts the cost of the second recharge and the purpose of this investigation is to analyze options with higher fuel enrichment in U-235 to minimize the number of assemblies in this recharge. The analyses effected show that the designs proposed by GE do not fulfill the required energy proposed for the cycle, even using in the recharge only fuel with 3.03% of enrichment. It is proposed, likewise, the fuel enrichment up to 3.25% to satisfy the energy demand with a minimum of assemblies. [Espanol] Se presentan los resultados de un estudio solicitado por la Comision Federal de Electricidad (CFE) para analizar el ciclo 3, de la unidad 1 de la Central Laguna Verde (CLV), y determinar el impacto del efecto de quemado llevado a cabo con las pruebas de arranque y por la operacion de los ciclos 1 y 2 con base en la tecnica de alargamiento del ciclo conocida como coastdown1. Los calculos se realizaron con el paquete de codigos FMS para la administracion de combustible, usando el codigo PRESTO-B que analiza el reactor en forma detallada en tres dimensiones y en estado estacionario. Se analizaron en el estudio los esquemas de fraccion de recarga propuesta por la General Electric (GE) con el efecto de

Core reactor simulation of the Central Laguna Verde (CLV) reactor in stationary state and an example of the application in the recharge options analysis of cycle 3; Simulacion del nucleo del reactor de la Central Laguna Verde (CLV) en estado estacionario y ejemplo de aplicacion en el analisis de alternativas de recarga del ciclo 3

Energy Technology Data Exchange (ETDEWEB)

Ocampo Mansilla, Hector; Francois Lacouture, Juan Luis; Blanco Lara, Jesus; Cortes Campos, Carlos Cristobal; Esquivias Montoya, Jesus; Esquivel Torres, Jose Luis; Martin del Campo Marquez, Cecilia [Instituto de Investigaciones Electricas, Cuernavaca (Mexico); Montes Tadeo, Jose Luis [Instituto Nacional de Investigaciones Nucleares (ININ), Salazar (Mexico); Sanchez Herrera, Luciano; Torres Alvarez, Carlos [Comision Federal de Electricidad (CFE), Mexico, D. F. (Mexico)

1992-12-31

The results are presented of a study requested by Comision Federal de Electricidad (CFE) for the analysis of Cycle 3, of Unit No. 1 of the Laguna Verde Nuclear Power Station (CLNV) and determine the burning effect impact, carried out with the starting tests and the operation of Cycles 1 and 2 on base of the cycle extension known as coastdown. The calculations were realized with the Code Package FMS for fuel managing, using the Code PRESTO-B that analyzes the reactor in detailed form in three dimensions an in stationary state. In the study the schemes of fraction of recharge proposed by General Electric (GE) were analyzed with the effect of cycle extension. The initial design value of 100 assemblies for Cycle 3, GE proposes to increase such fraction from 112 to 120 assemblies. This impacts the cost of the second recharge and the purpose of this investigation is to analyze options with higher fuel enrichment in U-235 to minimize the number of assemblies in this recharge. The analyses effected show that the designs proposed by GE do not fulfill the required energy proposed for the cycle, even using in the recharge only fuel with 3.03% of enrichment. It is proposed, likewise, the fuel enrichment up to 3.25% to satisfy the energy demand with a minimum of assemblies. [Espanol] Se presentan los resultados de un estudio solicitado por la Comision Federal de Electricidad (CFE) para analizar el ciclo 3, de la unidad 1 de la Central Laguna Verde (CLV), y determinar el impacto del efecto de quemado llevado a cabo con las pruebas de arranque y por la operacion de los ciclos 1 y 2 con base en la tecnica de alargamiento del ciclo conocida como coastdown1. Los calculos se realizaron con el paquete de codigos FMS para la administracion de combustible, usando el codigo PRESTO-B que analiza el reactor en forma detallada en tres dimensiones y en estado estacionario. Se analizaron en el estudio los esquemas de fraccion de recarga propuesta por la General Electric (GE) con el efecto de

Refuelling nuclear reactors

International Nuclear Information System (INIS)

Stacey, J.; Webb, J.; White, W.P.; McLaren, N.H.

1981-01-01

An improved nuclear reactor refuelling machine is described which can be left in the reactor vault to reduce the off-load refuelling time for the reactor. The system comprises a gripper device rangeable within a tubular chute, the gripper device being movable by a pantograph. (U.K.)

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)

Method of constructing lower dry well access tunnel for nuclear reactor container

International Nuclear Information System (INIS)

Kume, Tadashi; Furukawa, Hedeyasu.

1993-01-01

The method of the present invention facilitates construction of a lower dry well access tunnel for a nuclear reactor container. The lower dry well access tunnel is constructed across the reactor container and the reactor main body foundation. In this case, the lower dry well access tunnel is divided into three sections, i.e., axial end portions and a central portion. At first, each of the end portions is attached to the walls of the reactor container and the reactor main body foundation respectively. Subsequently, the central portion is attached to each of the end portions. An adjusting margin is previously provided to the central portion upon manufacturing each of the sections for adjusting deviation from a nominal size upon construction. In such a construction method, it is possible to eliminate interference accident during construction between the end portions of the lower dry well access tunnel and the reactor container and the reactor main body foundation, to facilitate the construction. Further, this facilitates the fabricating operation for dimensional alignment between the lower dry well access tunnel, and the reactor container and the reactor main body foundation. (I.S.)

Choosing a standard reactor: International competition and domestic politics in Chinese nuclear policy

International Nuclear Information System (INIS)

Ramana, M.V.; Saikawa, Eri

2011-01-01

China has ambitious plans to expand its nuclear power capacity. One of the policy goals that high-level policymakers have desired is to base the nuclear program on a standardized reactor design. However, this has not materialized so far. By examining its nuclear reactor choices for individual projects, we argue that China’s policymaking process has been greatly influenced by international competition and domestic politics. Multiple international nuclear vendors are intent upon maintaining their respective niches in the expanding Chinese reactor market, and they have used various forms of economic and political pressure to achieve their objectives. On the other hand, China’s policymaking process is fragmented and the shifting power balances among powerful domestic actors do not allow a fixed path to be followed. Further, because of the high costs and potential profits involved, nuclear reactor choices in China have been driven not just by technical considerations but also by foreign and trade policy objectives. All of these make it unlikely that China will standardize the reactor type it constructs in the near future. -- Highlights: ► China’s nuclear power policymaking has been fragmented and without central control. ► Multiple domestic actors have pursued independent agendas. ► International nuclear vendors have intensely competed for Chinese reactor contracts. ► Economic, political and foreign policy goals have driven reactor contract decisions. ► China is unlikely to construct only a standardized reactor design.

Nuclear reactor internals arrangement

International Nuclear Information System (INIS)

Frisch, E.; Andrews, H.N.

1976-01-01

A nuclear reactor internals arrangement is disclosed which facilitates reactor refueling. A reactor vessel and a nuclear core is utilized in conjunction with an upper core support arrangement having means for storing withdrawn control rods therein. The upper core support is mounted to the underside of the reactor vessel closure head so that upon withdrawal of the control rods into the upper core support, the closure head, the upper core support and the control rods are removed as a single unit thereby directly exposing the core for purposes of refueling

Indian advanced nuclear reactors

International Nuclear Information System (INIS)

Saha, D.; Sinha, R.K.

2005-01-01

For sustainable development of nuclear energy, a number of important issues like safety, waste management, economics etc. are to be addressed. To do this, a number of advanced reactor designs as well as fuel cycle technologies are being pursued worldwide. The advanced reactors being developed in India are the AHWR and the CHTR. Both the reactors use thorium based fuel and have many passive features. This paper describes the Indian advanced reactors and gives a brief account of the international initiatives for the sustainable development of nuclear energy. (author)

Licensing of nuclear reactor operators

International Nuclear Information System (INIS)

1979-09-01

Recommendations are presented for the licensing of nuclear reactor operators in units licensed according to the legislation in effect. They apply to all physical persons designated by the Operating Organization of the nuclear reactor or reactors to execute any of the following functional activities: a) to manipulate the controls of a definite reactor b) to direct the authorized activities of the reactor operators licesed according to the present recommendations. (F.E.) [pt

Sodium-cooled nuclear reactor

International Nuclear Information System (INIS)

Hammers, H.W.

1982-01-01

The invention concerns a sodium-cooled nuclear reactor, whose reactor tank contains the primary circuit, shielding surrounding the reactor core and a primary/secondary heat exchanger, particularly a fast breeder reactor on the module principle. In order to achieve this module principle it is proposed to have electromagnetic circulating pumps outside the reactor tank, where the heat exchanger is accomodated in an annular case above the pumps. This case has several openings at the top end to the space above the reactor core, some smaller openings in the middle to the same space and is connected at the bottom to an annular space between the tank wall and the reactor core. As a favoured variant, it is proposed that the annular electromagnetic pumps should be arranged concentrically to the reactor tank, where there is an annual duct on the inside of the reactor tank. In this way the sodium-cooled nuclear reactor is made suitable as a module with a large number of such elements. (orig.) [de

Development of a central PC-based system for reactor signal monitoring and analysis

International Nuclear Information System (INIS)

Karim, A.; Ansari, S.A.; Rauf Baig, A.

1998-01-01

A personal computer based system was developed for on-line monitoring, signal processing and display of important reactor parameters of the Pakistan Research Reactor-1. The system was designed for assistance to both reactor operator and users. It performs three main functions. The first is the centralized radiation monitoring in and around the reactor building. The computer acquires signals from radiation monitoring channels and continuously displays them on distributed monitors. Trend monitoring and alarm generation is also done. In case of any abnormal condition the radiation level data is automatically stored in computer memory for detailed off-line analysis. In the second part the computer does the performance testing of nuclear instrumentation channels by signal statistical analysis, and generates alarm in case the channel standard deviation error exceeds the permissible error. Mean values of important nuclear signals are also displayed on distributed monitors as a part of reactor safety parameters display system. The third function is on-line computation of reactor physics parameters of the core which are important from operational and safety points-of-view. The signals from radiation protection system and nuclear instrumentation channels in the reactor were interfaced with the computer for this purpose. The development work was done under an IAEA research contract as a part of coordinated research programme. (author)

Nuclear reactor simulator

International Nuclear Information System (INIS)

Baptista, Vinicius Damas

1996-01-01

The Nuclear Reactor Simulator was projected to help the basic training in the formation of the Nuclear Power Plants operators. It gives the trainee the opportunity to see the nuclear reactor dynamics. It's specially indicated to be used as the support tool to NPPT (Nuclear Power Preparatory Training) from NUS Corporation. The software was developed to Intel platform (80 x 86, Pentium and compatible ones) working under the Windows operational system from Microsoft. The program language used in development was Object Pascal and the compiler used was Delphi from Borland. During the development, computer algorithms were used, based in numeric methods, to the resolution of the differential equations involved in the process. (author)

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)

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.)

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)

Transmutation of nuclear waste in nuclear reactors

International Nuclear Information System (INIS)

Abrahams, K.; Kloosterman, J.L.; Pilate, S.; Wehmann, U.K.

1996-03-01

The objective of this joint study of ECN, Belgonucleaire, and Siemens is to investigate possibilities for transmutation of nuclear waste in regular nuclear reactors or in special transmutation devices. Studies of possibilities included the limits and technological development steps which would be needed. Burning plutonium in fast reactors, gas-cooled high-temperature reactors and light water reactors (LWR) have been considered. For minor actinides the transmutation rate mainly depends on the content of the minor actinides in the reactor and to a much less degree on the fact whether one uses a homogeneous system (with the actinides mixed into the fuel) or a heterogeneous system. If one wishes to stabilise the amount of actinides from the present LWRs, about 20% of all nuclear power would have to be generated in special burner reactors. It turned out that reactor transmutation of fission products would require considerable recycling efforts and that the time needed for a substantial transmutation would be rather long for the presently available levels of the neutron flux. If one would like to design burner systems which can serve more light water reactors, a large effort would be needed and other burners (possibly driven by accelerators) should be considered. (orig.)

Central eastern Europe approach to the security over nuclear materials

International Nuclear Information System (INIS)

Smagala, G.

2002-01-01

Full text: This paper presents an overview of the national approaches to physical protection of nuclear materials in Central Eastern Europe (CEE), with an emphasis on Poland. Soviet influence in the past led to inadequate safety culture in nuclear activities and insufficient security of nuclear materials and facilities in the region. In the centralized economies all aspects of nuclear activities, including ownership of the nuclear facilities, were the responsibility of the state with no clear separation between regulating and promoting functions. During the last decade a significant progress has been made in the region to clean up the legacy of the past and to improve practices in physical protection of nuclear materials. The countries of Central Eastern Europe have had many similar deficiencies in nuclear field and problems to overcome, but cannot be viewed as a uniform block. There are local variations within the region in a size of nuclear activities, formulated respective regulations and adopted measures to secure nuclear materials and facilities. Nevertheless, all twelve nations, with nuclear reactors and without nuclear facilities, have joined the convention on the physical protection of nuclear material and most of them declare that they have followed the IAEA recommendations INFCIRC/225/Rev.4 to elaborate and implement their physical protection systems of nuclear materials and facilities. The largest request for an international advisory mission (IPPAS) to review states' physical protection systems and to address needs for improvement was received from the countries of Central Eastern Europe. Poland belongs to the beneficiaries where the IPPAS mission and later follow-up consultations resulted in physical protection upgrade of the research reactor under the IAEA/US/UK technical assistance project. A powerful incentive to the progress made in a number of CEE countries was the goal of accession to the European Union. The physical protection of nuclear

HIGH TEMPERATURE, HIGH POWER HETEROGENEOUS NUCLEAR REACTOR

Science.gov (United States)

Hammond, R.P.; Wykoff, W.R.; Busey, H.M.

1960-06-14

A heterogeneous nuclear reactor is designed comprising a stationary housing and a rotatable annular core being supported for rotation about a vertical axis in the housing, the core containing a plurality of radial fuel- element supporting channels, the cylindrical empty space along the axis of the core providing a central plenum for the disposal of spent fuel elements, the core cross section outer periphery being vertically gradated in radius one end from the other to provide a coolant duct between the core and the housing, and means for inserting fresh fuel elements in the supporting channels under pressure and while the reactor is in operation.

Generalities about nuclear reactors

International Nuclear Information System (INIS)

Jaouen, C.; Beroux, P.

2012-01-01

From Zoe, the first nuclear reactor, till the current EPR, the French nuclear industry has always advanced by profiting from the feedback from dozens of years of experience and operations, in particular by drawing lessons from the most significant events in its history, such as the Fukushima accident. The new generations of reactors must improve safety and economic performance so that the industry maintain its legitimacy and its share in the production of electricity. This article draws the history of nuclear power in France, gives a brief description of the pressurized water reactor design, lists the technical features of the different versions of PWR that operate in France and compares them with other types of reactors. The feedback experience concerning safety, learnt from the major nuclear accidents Three Miles Island (1979), Chernobyl (1986) and Fukushima (2011) is also detailed. Today there are 26 third generation reactors being built in the world: 4 EPR (1 in Finland, 1 in France and 2 in China); 2 VVER-1200 in Russia, 8 AP-1000 (4 in China and 4 in the Usa), 8 APR-1400 (4 in Korea and 4 in UAE), and 4 ABWR (2 in Japan and 2 in Taiwan)

Nuclear reactors to come

International Nuclear Information System (INIS)

Lung, M.

2002-01-01

The demand for nuclear energy will continue to grow at least till 2050 because of mainly 6 reasons: 1) the steady increase of the world population, 2) China, India and Indonesia will reach higher social standard and their energy consumption will consequently grow, 3) fossil energy resources are dwindling, 4) coal will be little by little banned because of its major contribution to the emission of green house effect gas, 5) renewable energies need important technological jumps to be really efficient and to take the lead, and 6) fusion energy is not yet ready to take over. All these reasons draw a promising future for nuclear energy. Today 450 nuclear reactors are operating throughout the world producing 17% of the total electrical power demand. In order to benefit fully of this future, nuclear industry has to improve some characteristics of reactors: 1) a more efficient use of uranium (it means higher burnups), 2) a simplification and automation of reprocessing-recycling chain of processes, 3) efficient measures against proliferation and against any misuse for terrorist purposes, and 4) an enhancement of safety for the next generation of reactors. The characteristics of fast reactors and of high-temperature reactors will likely make these kinds of reactors the best tools for energy production in the second half of this century. (A.C.)

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)

Requirements of coolants in nuclear reactors

International Nuclear Information System (INIS)

Abass, O. A. M.

2014-11-01

This study discussed the purposes and types of coolants in nuclear reactors to generate electricity. The major systems and components associated with nuclear reactors are cooling system. There are two major cooling systems utilized to convert the heat generated in the fuel into electrical power. The primary system transfers the heat from the fuel to the steam generator, where the secondary system begins. The steam formed in the steam generator is transferred by the secondary system to the main turbine generator, where it s converted into electricity after passing through the low pressure turbine. There are various coolants used in nuclear reactors-light water, heavy water and liquid metal. The two major types of water-cooled reactors are pressurized water reactors (PWR) and boiling water reactors (BWR) but pressurized water reactors are more in the world. Also discusses this study the reactors and impact of the major nuclear accidents, in the April 1986 disaster at the Chernobyl nuclear power plant in Ukraine was the product operators, and in the March 2011 at the Fukushima nuclear power plant in Japan was the product of earthquake of magnitude 9.0, the accidents caused the largest uncontrolled radioactive release into the environment.(Author)

Guidelines for nuclear reactor equipments safety-analysis

International Nuclear Information System (INIS)

1978-01-01

The safety analysis in approving the applications for nuclear reactor constructions (or alterations) is performed by the Committee on Examination of Reactor Safety in accordance with various guidelines prescribed by the Atomic Energy Commission. In addition, the above Committee set forth its own regulations for the safety analysis on common problems among various types of nuclear reactors. This book has collected and edited those guidelines and regulations. It has two parts: Part I includes the guidelines issued to date by the Atomic Energy Commission: and Part II - regulations of the Committee. Part I has collected 8 categories of guidelines which relate to following matters: nuclear reactor sites analysis guidelines and standards for their applications; standard exposure dose of plutonium; nuclear ship operation guidelines; safety design analysis guidelines for light-water type, electricity generating nuclear reactor equipments; safety evaluation guidelines for emergency reactor core cooling system of light-water type power reactors; guidelines for exposure dose target values around light-water type electricity generating nuclear reactor equipments, and guidelines for evaluation of above target values; and meteorological guidelines for the safety analysis of electricity generating nuclear reactor equipments. Part II includes regulations of the Committee concerning - the fuel assembly used in boiling-water type and in pressurized-water type reactors; techniques of reactor core heat designs, etc. in boiling-water reactors; and others

Nuclear energy in central and eastern Europe; L'energie nucleaire en Europe centrale et orientale

Energy Technology Data Exchange (ETDEWEB)

Cognet, G.; Billaud, E.; Battista, L.; Pascal, J. [CEA/Ambassade de France a Budapest (Hungary)

2010-07-15

Bulgaria, Hungary, Czech Republic, Romania, Slovakia and Slovenia have all nuclear power plants on their soil, their entry to the European Union has counter-balanced the number of member states that are not in favour of nuclear energy. All these countries committed themselves to upgrade their installations to European safety standards. This work is going on and new programs for service life extension and for construction of new units have been launched. All these countries have a strong culture and know-how in nuclear engineering inherited from soviet era. They stay very active in this field, for instance the ALLEGRO project that is carried out by a collaboration between Hungarian, Slovak, Czech research institutes and western partners aims at building the first demonstration plant of the future fourth generation reactor: a fast reactor cooled by gas. The situation of each country plus Croatia (that is a candidate to full E.U. membership) is reviewed. (A.C.)

Development of a central PC-based system for reactor signal monitoring and analysis

International Nuclear Information System (INIS)

Karim, A.; Ansari, S.A.; Baig, A.R.

1996-05-01

A personal computer based system was developed for on-line monitoring, signal processing and display of important parameters of the Pakistan Reactor-1. The system was designed for assistance to both reactor operator and users. It performs three main functions. The first is the centralized radiation monitoring in and around the reactor building. The computer acquires signals from radiation monitoring channels and continuously displays them on distributed monitors. Trend monitoring and alarm generation is also done. In case of any abnormal condition the radiation level data is automatically stored in computer memory for detailed off-line analysis. In the second part the computer does the performance testing of nuclear instrumentation channels by signal statistical analysis and generates alarm in case the channel standard deviation error exceeds the permissible error. Mean values of important nuclear signals are also displayed on distributed monitors as a part of reactor safety parameters display system. The third function is on-line computation of reactor physics parameters of the core which are important from operational and safety point-of-view. The signals from radiation protection system and nuclear instrumentation channels in the reactor were interfaced with the computer for this purpose. The development work was done under an IAEA research contract as a part of coordinated research programme. (author) 12 figs

Historical civilian nuclear accident based Nuclear Reactor Condition Analyzer

Science.gov (United States)

McCoy, Kaylyn Marie

There are significant challenges to successfully monitoring multiple processes within a nuclear reactor facility. The evidence for this observation can be seen in the historical civilian nuclear incidents that have occurred with similar initiating conditions and sequences of events. Because there is a current lack within the nuclear industry, with regards to the monitoring of internal sensors across multiple processes for patterns of failure, this study has developed a program that is directed at accomplishing that charge through an innovation that monitors these systems simultaneously. The inclusion of digital sensor technology within the nuclear industry has appreciably increased computer systems' capabilities to manipulate sensor signals, thus making the satisfaction of these monitoring challenges possible. One such manipulation to signal data has been explored in this study. The Nuclear Reactor Condition Analyzer (NRCA) program that has been developed for this research, with the assistance of the Nuclear Regulatory Commission's Graduate Fellowship, utilizes one-norm distance and kernel weighting equations to normalize all nuclear reactor parameters under the program's analysis. This normalization allows the program to set more consistent parameter value thresholds for a more simplified approach to analyzing the condition of the nuclear reactor under its scrutiny. The product of this research provides a means for the nuclear industry to implement a safety and monitoring program that can oversee the system parameters of a nuclear power reactor facility, like that of a nuclear power plant.

Nuclear reactor

International Nuclear Information System (INIS)

Mysels, K.J.; Shenoy, A.S.

1976-01-01

A nuclear reactor is described in which the core consists of a number of fuel regions through each of which regulated coolant flows. The coolant from neighbouring fuel regions is combined in a manner which results in an averaging of the coolant temperature at the outlet of the core. By this method the presence of hot streaks in the reactor is reduced. (UK)

Nuclear rocket engine reactor

Energy Technology Data Exchange (ETDEWEB)

Lanin, Anatoly

2013-07-01

Covers a new technology of nuclear reactors and the related materials aspects. Integrates physics, materials science and engineering Serves as a basic book for nuclear engineers and nuclear physicists. The development of a nuclear rocket engine reactor (NRER) is presented in this book. The working capacity of an active zone NRER under mechanical and thermal load, intensive neutron fluxes, high energy generation (up to 30 MBT/l) in a working medium (hydrogen) at temperatures up to 3100 K is displayed. Design principles and bearing capacity of reactors area discussed on the basis of simulation experiments and test data of a prototype reactor. Property data of dense constructional, porous thermal insulating and fuel materials like carbide and uranium carbide compounds in the temperatures interval 300 - 3000 K are presented. Technological aspects of strength and thermal strength resistance of materials are considered. The design procedure of possible emergency processes in the NRER is developed and risks for their origination are evaluated. Prospects of the NRER development for pilotless space devices and piloted interplanetary ships are viewed.

Control rod drive of nuclear reactor

International Nuclear Information System (INIS)

Zhuchkov, I.I.; Gorjunov, V.S.; Zaitsev, B.I.

1980-01-01

This invention relates to nuclear reactors and, more particularly, to a drive of a control rod of a nuclear reactor and allows power control, excess reactivity compensation, and emergency shut-down of a reactor. (author)

Nuclear reactor

International Nuclear Information System (INIS)

Tilliette, Z.

1975-01-01

A description is given of a nuclear reactor and especially a high-temperature reactor in which provision is made within a pressure vessel for a main cavity containing the reactor core and a series of vertical cylindrical pods arranged in spaced relation around the main cavity and each adapted to communicate with the cavity through two collector ducts or headers for the primary fluid which flows downwards through the reactor core. Each pod contains two superposed steam-generator and circulator sets disposed in substantially symmetrical relation on each side of the hot primary-fluid header which conveys the primary fluid from the reactor cavity to the pod, the circulators of both sets being mounted respectively at the bottom and top ends of the pod

Installation of a new type of nuclear reactor in Mexico: advantages and disadvantages; Instalacion de un nuevo tipo de reactor nuclear en Mexico: ventajas y desventajas

Energy Technology Data Exchange (ETDEWEB)

Jurado P, M.; Martin del Campo M, C. [FI-UNAM, 04510 Mexico D.F. (Mexico)]. e-mail: mjp_green@hotmail.com

2005-07-01

In this work the main advantages and disadvantages of the installation of a new type of nuclear reactor different to the BWR type reactor in Mexico are presented. A revision of the advanced reactors is made that are at the moment in operation and of the advanced reactors that are in construction or one has already planned its construction in the short term. Specifically the A BWR and EPR reactors are analyzed. (Author)

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1993-11-01

This document is a compilation and source list of nuclear safety criteria that the Nuclear Regulatory Commission (NRC) applies to licensed reactors; it can be used by DOE and DOE contractors to identify NRC criteria to be evaluated for application to the DOE reactors under their cognizance. The criteria listed are those that are applied to the areas of nuclear safety addressed in the safety analysis report of a licensed reactor. They are derived from federal regulations, USNRC regulatory guides, Standard Review Plan (SRP) branch technical positions and appendices, and industry codes and standards

Potential of light water reactors for future nuclear power plants

International Nuclear Information System (INIS)

Gueldner, R.

2003-01-01

Energy consumption worldwide is going to increase further in the next few decades. Reliable supplies of electricity can be achieved only by centralized power plant structures. In this scenario, nuclear power plants are going to play a leading role as reliable and competitive plants, also under deregulated market conditions. Today, light water reactors have achieved a leading position, both technically and economically, contributing 85% to worldwide electricity generation in nuclear plants. They will continue to be a proven technology in power generation. In many countries, activities therefore are concentrated on extending the service life of plants beyond a period of forty years. New nuclear generating capacities are expected to be created and added from the end of this decade onward. Most of this capacity will be in light water reactors. The concepts of third-generation reactors will meet all economic and technical safety requirements of the 21st century and will offer considerable potential for further development. Probably some thirty years from now, fourth-generation nuclear power plants will be ready for commercial application. These plants will penetrate especially new sectors of the energy markets. Public acceptance of new nuclear power plants is not a matter of reactor lines, provided that safety requirements are met. The important issue is the management of radioactive waste. The construction of new nuclear power plants in Western Europe and North America mainly hinges on the ability to explain to the public that there is a need for new plants and that nuclear power is fundamental to assuring sustainable development. (orig.)

Nuclear reactor PBMR and cogeneration; Reactor nuclear PBMR y cogeneracion

Energy Technology Data Exchange (ETDEWEB)

Ramirez S, J. R.; Alonso V, G., E-mail: ramon.ramirez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2013-10-15

In recent years the nuclear reactor designs for the electricity generation have increased their costs, so that at the moment costs are managed of around the 5000 US D for installed kw, reason for which a big nuclear plant requires of investments of the order of billions of dollars, the designed reactors as modular of low power seek to lighten the initial investment of a big reactor dividing the power in parts and dividing in modules the components to lower the production costs, this way it can begin to build a module and finished this to build other, differing the long term investment, getting less risk therefore in the investment. On the other hand the reactors of low power can be very useful in regions where is difficult to have access to the electric net being able to take advantage of the thermal energy of the reactor to feed other processes like the water desalination or the vapor generation for the processes industry like the petrochemical, or even more the possible hydrogen production to be used as fuel. In this work the possibility to generate vapor of high quality for the petrochemical industry is described using a spheres bed reactor of high temperature. (Author)

Towards nuclear fusion reactors

International Nuclear Information System (INIS)

1993-11-01

The results of nuclear fusion researches in JAERI are summarized. In this report, following themes are collected: the concept of fusion reactor (including ITER), fusion reactor safety, plasma confinement, fusion reactor equipment, and so on. Includes glossary. (J.P.N.)

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

Nuclear Reactor Engineering Analysis Laboratory

International Nuclear Information System (INIS)

Carlos Chavez-Mercado; Jaime B. Morales-Sandoval; Benjamin E. Zayas-Perez

1998-01-01

The Nuclear Reactor Engineering Analysis Laboratory (NREAL) is a sophisticated computer system with state-of-the-art analytical tools and technology for analysis of light water reactors. Multiple application software tools can be activated to carry out different analyses and studies such as nuclear fuel reload evaluation, safety operation margin measurement, transient and severe accident analysis, nuclear reactor instability, operator training, normal and emergency procedures optimization, and human factors engineering studies. An advanced graphic interface, driven through touch-sensitive screens, provides the means to interact with specialized software and nuclear codes. The interface allows the visualization and control of all observable variables in a nuclear power plant (NPP), as well as a selected set of nonobservable or not directly controllable variables from conventional control panels

Nuclear Capacity Building through Research Reactors

International Nuclear Information System (INIS)

2017-01-01

Four Instruments: •The IAEA has recently developed a specific scheme of services for Nuclear Capacity Building in support of the Member States cooperating research reactors (RR) willing to use RRs as a primary facility to develop nuclear competences as a supporting step to embark into a national nuclear programme. •The scheme is composed of four complementary instruments, each of them being targeted to specific objective and audience: Distance Training: Internet Reactor Laboratory (IRL); Basic Training: Regional Research Reactor Schools; Intermediate Training: East European Research Reactor Initiative (EERRI); Group Fellowship Course Advanced Training: International Centres based on Research Reactors (ICERR)

Artificial intelligence in nuclear reactor operation

International Nuclear Information System (INIS)

Da Ruan; Benitez-Read, J.S.

2005-01-01

Assessment of four real fuzzy control applications at the MIT research reactor in the US, the FUGEN heavy water reactor in Japan, the BR1 research reactor in Belgium, and a TRIGA Mark III reactor in Mexico will be examined through a SWOT analysis (strengths, weakness, opportunities, and threats). Special attention will be paid to the current cooperation between the Belgian Nuclear Research Centre (SCK·CEN) and the Mexican Nuclear Centre (ININ) on AI-based intelligent control for nuclear reactor operation under the partial support of the National Council for Science and Technology of Mexico (CONACYT). (authors)

Physical start up of the Dalat nuclear research reactor with the core configuration having a central neutron trap

International Nuclear Information System (INIS)

Pham Duy Hien; Ngo Quang Huy; Vu Hai Long; Tran Khanh Mai

1994-01-01

After the reactor has reached physical criticality with the core configuration exempt from central neutron trap on 1 November 1983, the core configuration with a central neutron trap has been arranged in the reactor and the reactor has reached physical criticality with this core configuration at 17h48 on 18 December 1983. The integral worths of different control rods are determined with accuracy. 2 refs., 24 figs., 18 tabs

Graphite materials for nuclear reactors

International Nuclear Information System (INIS)

Oku, Tatsuo

1991-01-01

Graphite materials have been used in the nuclear fission reactors from the beginning of the reactor development for the speed reduction and reflection of neutron. Graphite materials are used both as a moderator and as a reflector in the core of high temperature gas-cooled reactors, and both as a radiation shielding material and as a reflector in the surrounding of the core for the fast breeder reactor. On the other hand, graphite materials are being positively used as a first wall of plasma as it is known that low Z materials are useful for holding high temperature plasma in the nuclear fusion devices. In this paper the present status of the application of graphite materials to the nuclear fission reactors and fusion devices (reactors) is presented. In addition, a part of results on the related properties to the structural design and safety evaluation and results examined on the subjects that should be done in the future are also described. (author)

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

Brazilian nuclear power plants decommissioning plan for a multiple reactor site

Energy Technology Data Exchange (ETDEWEB)

Monteiro, Deiglys B.; Moreira, Joao M.L.; Maiorino, Jose R., E-mail: deiglys.monteiro@ufabc.edu.br, E-mail: joao.moreira@ufabc.edu.br, E-mail: joserubens.maiorino@ufabc.edu.br [Universidade Federal do ABC (CECS/UFABC), Santo Andre, SP (Brazil). Centro de Engenharia, Modelagem e Ciencias Aplicadas. Programa de Pos-Graduacao em Energia e Engenharia da Energia

2015-07-01

Actually, Brazil has two operating Nuclear Power Plants and a third one under construction, all at Central Nuclear Almirante Alvaro Alberto - CNAAA. To comply with regulatory aspects the power plants operator, Eletronuclear, must present to Brazilian Nuclear Regulatory Agency, CNEN, a decommissioning plan. Brazilian experience with decommissioning is limited because none of any nuclear reactor at the country was decommissioned. In literature, decommissioning process is well described despite few nuclear power reactors have been decommissioned around the world. Some different approach is desirable for multiple reactors sites, case of CNAAA site. During the decommissioning, a great amount of wastes will be produced and have to be properly managed. Particularly, the construction of Auxiliary Services on the site could be a good choice due to the possibility of reducing costs. The present work intends to present to the Eletronuclear some aspects of the decommissioning concept and decommissioning management, storage and disposal de wastes, based on the available literature, regulatory standards of CNEN and international experience as well as to suggest some solutions to be implemented at CNAAA site before starts the decommissioning project in order to maximize the benefits. (author)

Brazilian nuclear power plants decommissioning plan for a multiple reactor site

International Nuclear Information System (INIS)

Monteiro, Deiglys B.; Moreira, Joao M.L.; Maiorino, Jose R.

2015-01-01

Actually, Brazil has two operating Nuclear Power Plants and a third one under construction, all at Central Nuclear Almirante Alvaro Alberto - CNAAA. To comply with regulatory aspects the power plants operator, Eletronuclear, must present to Brazilian Nuclear Regulatory Agency, CNEN, a decommissioning plan. Brazilian experience with decommissioning is limited because none of any nuclear reactor at the country was decommissioned. In literature, decommissioning process is well described despite few nuclear power reactors have been decommissioned around the world. Some different approach is desirable for multiple reactors sites, case of CNAAA site. During the decommissioning, a great amount of wastes will be produced and have to be properly managed. Particularly, the construction of Auxiliary Services on the site could be a good choice due to the possibility of reducing costs. The present work intends to present to the Eletronuclear some aspects of the decommissioning concept and decommissioning management, storage and disposal de wastes, based on the available literature, regulatory standards of CNEN and international experience as well as to suggest some solutions to be implemented at CNAAA site before starts the decommissioning project in order to maximize the benefits. (author)

Neutron spectra calculation and doses in a subcritical nuclear reactor based on thorium; Calculo de espectros de neutrones y dosis en un reactor nuclear subcritico a base de Torio

Energy Technology Data Exchange (ETDEWEB)

Medina C, D.; Hernandez A, P. L.; Hernandez D, V. M.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico); Sajo B, L., E-mail: dmedina_c@hotmail.com [Universidad Simon Bolivar, Laboratorio de Fisica Nuclear, Apdo. Postal 89000, Caracas 1080A (Venezuela, Bolivarian Republic of)

2015-10-15

This paper describes a heterogeneous subcritical nuclear reactor with molten salts based on thorium, with graphite moderator and a source of {sup 252}Cf, whose dose levels in the periphery allows its use in teaching and research activities. The design was done by the Monte Carlo method with the code MCNP5 where the geometry, dimensions and fuel was varied in order to obtain the best design. The result is a cubic reactor of 110 cm side with graphite moderator and reflector. In the central part they have 9 ducts that were placed in the direction of axis Y. The central duct contains the source of {sup 252}Cf, of 8 other ducts, are two irradiation ducts and the other six contain a molten salt ({sup 7}LiF - BeF{sub 2} - ThF{sub 4} - UF{sub 4}) as fuel. For design the k{sub eff}, neutron spectra and ambient dose equivalent was calculated. In the first instance the above calculation for a virgin fuel was called case 1, then a percentage of {sup 233}U was used and the percentage of Th was decreased and was called case 2. This with the purpose to compare two different fuels working inside the reactor. In the case 1 a value was obtained for the k{sub eff} of 0.13 and case 2 of 0.28, maintaining the subcriticality in both cases. In the dose levels the higher value is in case 2 in the axis Y with a value of 3.31 e-3 ±1.6% p Sv/Q this value is reported in for one. With this we can calculate the exposure time of personnel working in the reactor. (Author)

Reactor core of nuclear reactor

International Nuclear Information System (INIS)

Sasagawa, Masaru; Masuda, Hiroyuki; Mogi, Toshihiko; Kanazawa, Nobuhiro.

1994-01-01

In a reactor core, a fuel inventory at an outer peripheral region is made smaller than that at a central region. Fuel assemblies comprising a small number of large-diameter fuel rods are used at the central region and fuel assemblies comprising a great number of smalldiameter fuel rods are used at the outer peripheral region. Since a burning degradation rate of the fuels at the outer peripheral region can be increased, the burning degradation rate at the infinite multiplication factor of fuels at the outer region can substantially be made identical with that of the fuels in the inner region. As a result, the power distribution in the direction of the reactor core can be flattened throughout the entire period of the burning cycle. Further, it is also possible to make the degradation rate of fuels at the outer region substantially identical with that of fuels at the inner side. A power peak formed at the outer circumferential portion of the reactor core of advanced burning can be lowered to improve the fuel integrity, and also improve the reactor safety and operation efficiency. (N.H.)

Nuclear reactors and fuel cycle

International Nuclear Information System (INIS)

2014-01-01

The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100 th nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U 3 O 8 were replaced by U 3 Si 2 -based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to fulfill its mission that is

Nuclear reactors and fuel cycle

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-07-01

The Nuclear Fuel Center (CCN) of IPEN produces nuclear fuel for the continuous operation of the IEA-R1 research reactor of IPEN. The serial production started in 1988, when the first nuclear fuel element was delivered for IEA-R1. In 2011, CCN proudly presents the 100{sup th} nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U{sub 3}O{sub 8} were replaced by U{sub 3}Si{sub 2}-based fuels, and the research of U Mo is currently under investigation. Additionally, the Brazilian Multipurpose Research Reactor (RMB), whose project will rely on the CCN for supplying fuel and uranium targets. Evolving from an annual production from 10 to 70 nuclear fuel elements, plus a thousand uranium targets, is a huge and challenging task. To accomplish it, a new and modern Nuclear Fuel Factory is being concluded, and it will provide not only structure for scaling up, but also a safer and greener production. The Nuclear Engineering Center has shown, along several years, expertise in the field of nuclear, energy systems and correlated areas. Due to the experience obtained during decades in research and technological development at Brazilian Nuclear Program, personnel has been trained and started to actively participate in design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in production of radiopharmaceuticals. The institution has participated in the monitoring and technical support concerning the safety, licensing and modernization of the research reactors IPEN/MB-01 and IEA-R1. Along the last two decades, numerous specialized services of engineering for the Brazilian nuclear power plants Angra 1 and Angra 2 have been carried out. The contribution in service, research, training, and teaching in addition to the development of many related technologies applied to nuclear engineering and correlated areas enable the institution to

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

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)

Availability analysis of the nuclear instrumentation of a research reactor

International Nuclear Information System (INIS)

Vianna Filho, Alfredo Marques

2016-01-01

The maintenance of systems and equipment is a central question related to Production Engineering. Although systems are not fully reliable, it is often necessary to minimize the failure occurrence likelihood. The failures occurrences can have disastrous consequences during a plane flight or operation of a nuclear power plant. The elaboration of a maintenance plan has as objective the prevention and recovery from system failures, increasing reliability and reducing the cost of unplanned shutdowns. It is also important to consider the issues related to organizations safety, especially those dealing with dangerous technologies. The objective of this thesis is to propose a method for maintenance analysis of a nuclear research reactor, using a socio-technical approach, and focused on existing conditions in Brazil. The research reactor studied belongs to the federal government and it is located in the city of Rio de Janeiro. The specific objective of this thesis is to develop the availability analysis of one of the principal systems of the research reactor, the nuclear instrumentation system. In this analysis, were taken into account not only the technical aspects of the modules related to nuclear instrumentation system, but also the human and organizational factors that could affect the availability of the nuclear instrumentation system. The results showed the influence of these factors on the availability of the nuclear instrumentation system. (author)

Accelerator driven reactors and nuclear waste management projects in the Czech Republic

Energy Technology Data Exchange (ETDEWEB)

Janouch, F. [Royal Institute of Technology, Stockholm (Sweden); Mach, R. [Institute of Nuclear Physics, Rez near Prague (Czechoslovakia)

1995-10-01

The Czech Republic is almost the only country in the central Europe which continues with the construction of nuclear power reactors. Its small territory and dense population causes public worries concerning the disposal of the spent nuclear fuel. The Czech nuclear scientists and the power companies and the nuclear industries are therefore looking for alternative solutions. The Los Alamos ATW project had received a positive response in the Czech mass-media and even in the industrial and governmental quarters. The recent scientific symposium {open_quotes}Accelerator driven reactors and nuclear waste management{close_quotes} convened at the Liblice castle near Prague, 27-29. 6. 1994 and sponsored by the Czech Energy Company CEZ, reviewed the competencies and experimental basis in the Czech republic and made the first attempt to formulate the national approach and to establish international collaboration in this area.

Reactors physics. Bases of nuclear physics

International Nuclear Information System (INIS)

Diop, Ch.M.

2006-01-01

The aim of nuclear reactor physics is to quantify the relevant macroscopic data for the characterization of the neutronic state of a reactor core and to evaluate the effects of radiations (neutrons and gamma radiations) on organic matter and on inorganic materials. This first article presents the bases of nuclear physics in the context of nuclear reactors: 1 - reactor physics and nuclear physics; 2 - atomic nucleus - basic definitions: nucleus constituents, dimensions and mass of the atomic nucleus, mass defect, binding energy and stability of the nucleus, strong interaction, nuclear momentums of nucleons and nucleus; 3 - nucleus stability and radioactivity: equation of evolution with time - radioactive decay law; alpha decay, stability limit of spontaneous fission, beta decay, electronic capture, gamma emission, internal conversion, radioactivity, two-body problem and notion of radioactive equilibrium. (J.S.)

Hysteresis phenomenon in nuclear reactor dynamics

Energy Technology Data Exchange (ETDEWEB)

Pirayesh, Behnam; Pazirandeh, Ali [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch; Akbari, Monireh [Shahid Rajaee Teacher Training Univ., Tehran (Iran, Islamic Republic of). Dept. of Mathematics

2017-05-15

This paper applies a nonlinear analysis method to show that hysteresis phenomenon, due to the Saddle-node bifurcation, may occur in the nuclear reactor. This phenomenon may have significant effects on nuclear reactor dynamics and can even be the beginning of a nuclear reactor accident. A system of four dimensional nonlinear ordinary differential equations was considered to study the hysteresis phenomenon in a typical nuclear reactor. It should be noted that the reactivity was considered as a nonlinear function of state variables. The condition for emerging hysteresis was investigated using Routh-Hurwitz criterion and Sotomayor's theorem for saddle node bifurcation. A numerical analysis is also provided to illustrate the analytical results.

Nuclear reactor control column

International Nuclear Information System (INIS)

Bachovchin, D.M.

1982-01-01

The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest crosssectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor

NUCLEAR REACTOR

Science.gov (United States)

Miller, H.I.; Smith, R.C.

1958-01-21

This patent relates to nuclear reactors of the type which use a liquid fuel, such as a solution of uranyl sulfate in ordinary water which acts as the moderator. The reactor is comprised of a spherical vessel having a diameter of about 12 inches substantially surrounded by a reflector of beryllium oxide. Conventionnl control rods and safety rods are operated in slots in the reflector outside the vessel to control the operation of the reactor. An additional means for increasing the safety factor of the reactor by raising the ratio of delayed neutrons to prompt neutrons, is provided and consists of a soluble sulfate salt of beryllium dissolved in the liquid fuel in the proper proportion to obtain the result desired.

Innovative designs of nuclear reactors

International Nuclear Information System (INIS)

Gabaraev, B.A.; Cherepnin, Y.S.

2010-01-01

The world development scenarios predict at least a 2.5 time increase in the global consumption of primary energy in the first half of the twenty-first century. Much of this growth can be provided by the nuclear power which possesses important advantages over other energy technologies. However, the large deployment of nuclear sources may take place only when the new generation of reactors appears on the market and will be free of the shortcomings found in the existing nuclear power installations. The public will be more inclined to accept nuclear plants that have better economics; higher safety; more efficient management of the radioactive waste; lower risk of nuclear weapons proliferation, and provided that the focus is made on the energy option free of ∇ e 2 generation. Currently, the future of nuclear power is trusted to the technology based on fast reactors and closed fuel cycle. The latter implies reprocessing of the spent nuclear fuel of the nuclear plants and re-use of plutonium produced in power reactors

Nuclear characteristic simulation device for reactor core

International Nuclear Information System (INIS)

Arakawa, Akio; Kobayashi, Yuji.

1994-01-01

In a simulation device for nuclear characteristic of a PWR type reactor, there are provided a one-dimensional reactor core dynamic characteristic model for simulating one-dimensional neutron flux distribution in the axial direction of the reactor core and average reactor power based on each of inputted signals of control rod pattern, a reactor core flow rate, reactor core pressure and reactor core inlet enthalphy, and a three-dimensional reactor core dynamic characteristic mode for simulating three-dimensional power distribution of the reactor core, and a nuclear instrumentation model for calculating read value of the nuclear instrumentation disposed in the reactor based on the average reactor core power and the reactor core three-dimensional power distribution. A one-dimensional neutron flux distribution in the axial direction of the reactor core, a reactor core average power, a reactor core three-dimensional power distribution and a nuclear instrumentation read value are calculated. As a result, the three-dimensional power distribution and the power level are continuously calculated. Further, since the transient change of the three-dimensional neutron flux distribution is calculated accurately on real time, more actual response relative to a power monitoring device of the reactor core and operation performance can be simulated. (N.H.)

Nuclear Reactors and Technology; (USA)

Energy Technology Data Exchange (ETDEWEB)

Cason, D.L.; Hicks, S.C. (eds.)

1991-01-01

Nuclear Reactors and Technology (NRT) announces on a monthly basis the current worldwide information available from the open literature on nuclear reactors and technology, including all aspects of power reactors, components and accessories, fuel elements, control systems, and materials. This publication contains the abstracts of DOE reports, journal articles, conference papers, patents, theses, and monographs added to the Energy Science and Technology Database (EDB) during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international information obtained through the International Energy Agency's Energy Technology Data Exchange or government-to-government agreements. The digests in NRT and other citations to information on nuclear reactors back to 1948 are available for online searching and retrieval on EDB and Nuclear Science Abstracts (NSA) database. Current information, added daily to EDB, is available to DOE and its contractors through the DOE integrated Technical Information System. Customized profiles can be developed to provide current information to meet each user's needs.

Nuclear power in eastern and central Europe. Background paper

International Nuclear Information System (INIS)

Myers, L.C.

1993-11-01

The breakup of the former Soviet Union and other political changes in eastern and central Europe have opened up the area to closer scrutiny than was previously possible. Because of the accident at Chernobyl, nuclear power is one of the subjects that western nations have had a great deal of interest in exploring. The former Soviet Union designed and/or helped build more than 60 civilian reactors in the region. Most of these reactors follow one of two distinctly different designs: the VVER, or pressurized water reactor series; and the RBMK, which is a graphite-moderated, multi-channel reactor (the so-called Chernobyl type). In addition, there are two fast-breeder reactors and four graphite-moderated boiling water reactors for combined heat and power in operation in Russia. These last two designs are not widely distributed and so are not discussed in detail in this report. As noted above, the safety of Soviet-designed reactors has been of great concern around the world since the catastrophic events at Chernobyl in 1986. This paper will briefly describe the technology involved. It will also examine the main safety concerns, both technical and organizational, associated with each reactor type. In addition, the paper will review the nuclear power programs in the new countries emerging from the former Soviet Union and its satellites and discuss the international efforts underway to address the most pressing problems. (author). 1 tab

Nuclear power in eastern and central Europe. Background paper

Energy Technology Data Exchange (ETDEWEB)

Myers, L C [Library of Parliament, Ottawa, ON (Canada). Science and Technology Div.

1993-11-01

The breakup of the former Soviet Union and other political changes in eastern and central Europe have opened up the area to closer scrutiny than was previously possible. Because of the accident at Chernobyl, nuclear power is one of the subjects that western nations have had a great deal of interest in exploring. The former Soviet Union designed and/or helped build more than 60 civilian reactors in the region. Most of these reactors follow one of two distinctly different designs: the VVER, or pressurized water reactor series; and the RBMK, which is a graphite-moderated, multi-channel reactor (the so-called Chernobyl type). In addition, there are two fast-breeder reactors and four graphite-moderated boiling water reactors for combined heat and power in operation in Russia. These last two designs are not widely distributed and so are not discussed in detail in this report. As noted above, the safety of Soviet-designed reactors has been of great concern around the world since the catastrophic events at Chernobyl in 1986. This paper will briefly describe the technology involved. It will also examine the main safety concerns, both technical and organizational, associated with each reactor type. In addition, the paper will review the nuclear power programs in the new countries emerging from the former Soviet Union and its satellites and discuss the international efforts underway to address the most pressing problems. (author). 1 tab.

The case for innovation in nuclear reactor and fuel cycle designs

International Nuclear Information System (INIS)

Mourogov, V. M.

2000-01-01

Nuclear power is an important contributor to the world's electricity needs. In 1999 it supplied more than one-sixth of global electricity. Due to the fact that it is a capital intensive and sophisticated technology, the larger part of world nuclear capacity, i.e. 83 %, is concentrated in the industrialized countries, mainly OECD countries and economies in transition in Central and Eastern Europe (see table I and table II). Nuclear Power plays an important role in those industrialized countries by ensuring their energy independence, helping to keep the air clean and reducing carbon emissions. The development of nuclear power in industrialized countries was reached through intensive development of reactor manufacturing capabilities and a sophisticated fuel cycle infrastructure. The competitiveness of nuclear power was reached through developing large scale reactors units, between 1000-1600 MWe and aggressive NPP construction programmes. It is understood that waste management becomes economical within national waste repository concepts when the total national nuclear capacity reaches several GWe. Nuclear power plants are introduced in large electricity grids and they are used mainly for base load electricity generation

U.S. Nuclear Power Reactor Plant Status

Data.gov (United States)

Nuclear Regulatory Commission — Demographic data on U.S. commercial nuclear power reactors, including: plant name/unit number, docket number, location, licensee, reactor/containment type, nuclear...

VALIDATION OF SIMBAT-PWR USING STANDARD CODE OF COBRA-EN ON REACTOR TRANSIENT CONDITION

Directory of Open Access Journals (Sweden)

Muhammad Darwis Isnaini

2016-03-01

Full Text Available The validation of Pressurized Water Reactor typed Nuclear Power Plant simulator developed by BATAN (SIMBAT-PWR using standard code of COBRA-EN on reactor transient condition has been done. The development of SIMBAT-PWR has accomplished several neutronics and thermal-hydraulic calculation modules. Therefore, the validation of the simulator is needed, especially in transient reactor operation condition. The research purpose is for characterizing the thermal-hydraulic parameters of PWR1000 core, which be able to be applied or as a comparison in developing the SIMBAT-PWR. The validation involves the calculation of the thermal-hydraulic parameters using COBRA-EN code. Furthermore, the calculation schemes are based on COBRA-EN with fixed material properties and dynamic properties that calculated by MATPRO subroutine (COBRA-EN+MATPRO for reactor condition of startup, power rise and power fluctuation from nominal to over power. The comparison of the temperature distribution at nominal 100% power shows that the fuel centerline temperature calculated by SIMBAT-PWR has 8.76% higher result than COBRA-EN result and 7.70% lower result than COBRA-EN+MATPRO. In general, SIMBAT-PWR calculation results on fuel temperature distribution are mostly between COBRA-EN and COBRA-EN+MATPRO results. The deviations of the fuel centerline, fuel surface, inner and outer cladding as well as coolant bulk temperature in the SIMBAT-PWR and the COBRA-EN calculation, are due to the value difference of the gap heat transfer coefficient and the cladding thermal conductivity.

The fuel of nuclear reactors

International Nuclear Information System (INIS)

1995-03-01

This booklet is a presentation of the different steps of the preparation of nuclear fuels performed by Cogema. The documents starts with a presentation of the different French reactor types: graphite moderated reactors, PWRs using MOX fuel, fast breeder reactors and research reactors. The second part describes the fuel manufacturing process: conditioning of nuclear materials and fabrication of fuel assemblies. The third part lists the different companies involved in the French nuclear fuel industry while part 4 gives a short presentation of the two Cogema's fuel fabrication plants at Cadarache and Marcoule. Part 5 and 6 concern the quality assurance, the safety and reliability aspects of fuel elements and the R and D programs. The last part presents some aspects of the environmental and personnel protection performed by Cogema. (J.S.)

Directory of Nuclear Research Reactors 1994

International Nuclear Information System (INIS)

1995-08-01

The Directory of Nuclear Research Reactors is an output of the Agency's computerized Research Reactor Data Base (RRDB). It contains administrative, technical and utilization information on research reactors known to the Agency at the end of December 1994. The data base converted from mainframe to PC is written in Clipper 5.0 and the publication generation system uses Excel 4. The information was collected by the Agency through questionnaires sent to research reactor owners. All data on research reactors, training reactors, test reactors, prototype reactors and critical assemblies are stored in the RRDB. This system contains all the information and data previously published in the Agency's publication, Directory of Nuclear Research Reactor, as well as updated information

1980 nuclear plant survey: no reactors sold; more cancellations

International Nuclear Information System (INIS)

Friedlander, G.D.

1980-01-01

No sales were reported in 1979 by any of the big four reactor suppliers. Three cancellations were reported and construction was suspended on the Jersey Central Power and Light Co.'s Forked River unit. Since last year's survey, the commercial operation dates of about 80 units have been postponed from one year to indefinitely, and nuclear commitments are down from last year's 195 units to 193 units. Presently, there are 72 plants on line, with a capacity of more than 53,000 MW. A resumption of new reactor orders is expected in either late 1980 or early 1981

Nuclear reactor built, being built, or planned

International Nuclear Information System (INIS)

1991-06-01

This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1990. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington headquarters and field offices of DOE, from the US Nuclear Regulatory Commission, from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations, from US and foreign embassies, and from foreign governmental nuclear departments. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables; Section 2 includes nuclear reactors that are operating, being built, or planned; and Section 3 includes reactors that have been shut down permanently or dismantled. Sections 2 and 3 contain the following classification of reactors: Civilian, Production, Military, Export, and Critical Assembly

Preliminary concept of a zero power nuclear reactor core

International Nuclear Information System (INIS)

Mai, Luiz Antonio; Siqueira, Paulo de Tarso D.

2011-01-01

The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

Preliminary concept of a zero power nuclear reactor core

Energy Technology Data Exchange (ETDEWEB)

Mai, Luiz Antonio; Siqueira, Paulo de Tarso D., E-mail: lamai@ipen.b, E-mail: ptsiquei@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2011-07-01

The purpose of this work is to define a zero power core to study the neutronic behavior of a modern research reactor as the future RMB (Brazilian Nuclear Multipurpose reactor). The platform used was the IPEN/MB-01 nuclear reactor, installed at the Nuclear and Energy Research Institute (IPEN-CNEN/SP). Equilibrium among minimal changes in the current reactor facilities and an arrangement that will be as representative as possible of a future core were taken into account. The active parts of the elements (fuel and control/safety) were determined to be exactly equal the elements of a future reactor. After several technical discussions, a basic configuration for the zero power core was defined. This reactor will validate the neutronic calculations and will allow the execution of countless future experiments aiming a real core. Of all possible alternative configurations for the zero power core representative of a future reactor - named ZPC-MRR (Zero Power Core - Modern Research Reactor), it was concluded, through technical and practical arguments, that the core will have an array of 4 x 5 positions, with 19 fuel elements, identical in its active part to a standard MTR (Material Test Reactor), 4 control/safety elements having a unique flat surface and a central position of irradiation. The specifications of the fuel elements (FEs) are the same as defined to standard MTR in its active part, but the inferior nozzles are differentiated because ZPC-MRR will be a set without heat generation. A study of reactivity was performed using MCNP code, and it was estimated that it will have around 2700 pcm reactivity excess in its 19 FEs configuration (alike the present IPEN/MB-01 reactivity). The effective change in the IPEN/MB-01 reactor will be made only in the control rods drive mechanism. It will be necessary to modify the center of this mechanism. Major modifications in the facility will not be necessary. (author)

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)

Nuclear reactor

International Nuclear Information System (INIS)

Rau, P.

1980-01-01

The reactor core of nuclear reactors usually is composed of individual elongated fuel elements that may be vertically arranged and through which coolant flows in axial direction, preferably from bottom to top. With their lower end the fuel elements gear in an opening of a lower support grid forming part of the core structure. According to the invention a locking is provided there, part of which is a control element that is movable along the fuel element axis. The corresponding locking element is engaged behind a lateral projection in the opening of the support grid. The invention is particularly suitable for breeder or converter reactors. (orig.) [de

Safety of nuclear power reactors

International Nuclear Information System (INIS)

MacPherson, H.G.

1982-01-01

Safety is the major public issue to be resolved or accommodated if nuclear power is to have a future. Probabilistic Risk Analysis (PRA) of accidental releases of low-level radiation, the spread and activity of radiation in populated areas, and the impacts on public health from exposure evolved from the earlier Rasmussen Reactor Safety Study. Applications of the PRA technique have identified design peculiarities in specific reactors, thus increasing reactor safety and establishing a quide for evaluating reactor regulations. The Nuclear Regulatory Commission and reactor vendors must share with utilities the responsibility for reactor safety in the US and for providing reasonable assurance to the public. This entails persuasive public education and information that with safety a top priority, changes now being made in light water reactor hardware and operations will be adequate. 17 references, 2 figures, 2 tables

DOE fundamentals handbook: Nuclear physics and reactor theory

International Nuclear Information System (INIS)

1993-01-01

The Nuclear Physics and Reactor Theory Handbook was developed to assist nuclear facility operating contractors in providing operators, maintenance personnel, and the technical staff with the necessary fundamentals training to ensure a basic understanding of nuclear physics and reactor theory. The handbook includes information on atomic and nuclear physics; neutron characteristics; reactor theory and nuclear parameters; and the theory of reactor operation. This information will provide personnel with a foundation for understanding the scientific principles that are associated with various DOE nuclear facility operations and maintenance

Nuclear reactor cavity floor passive heat removal system

Science.gov (United States)

Edwards, Tyler A.; Neeley, Gary W.; Inman, James B.

2018-03-06

A nuclear reactor includes a reactor core disposed in a reactor pressure vessel. A radiological containment contains the nuclear reactor and includes a concrete floor located underneath the nuclear reactor. An ex vessel corium retention system includes flow channels embedded in the concrete floor located underneath the nuclear reactor, an inlet in fluid communication with first ends of the flow channels, and an outlet in fluid communication with second ends of the flow channels. In some embodiments the inlet is in fluid communication with the interior of the radiological containment at a first elevation and the outlet is in fluid communication with the interior of the radiological containment at a second elevation higher than the first elevation. The radiological containment may include a reactor cavity containing a lower portion of the pressure vessel, wherein the concrete floor located underneath the nuclear reactor is the reactor cavity floor.

The future of nuclear reactors

International Nuclear Information System (INIS)

Teller, E.

1989-01-01

The Atomic Energy Commission Advisory Committee on Reactor Safeguards began work in early 1948 with the firm and unanimous conviction that nuclear power could not survive a significant damaging accident. They as a committee felt that their job was to make reactors so safe that no such event would ever occur. However, ambitious reactor planners did not like all the buts and cautions that the committee was raising. They seemed to delay unduly their setting sail into the brave new world of clean, cheap, safe nuclear energy. The committee was soon nicknamed the Committee on Reactor Prevention. Reactors, of course, represented a tremendous step into the future. To an unprecedented extent, they were based on theory. But the committee did not have the luxury of putting a preliminary model into operation and waiting for difficulties to show up. In assessing new designs and developments, they had to anticipate future difficulties. Their proposals in good part were accepted, but their deep emphasis on safety did not become a part of the program. Today, forty years later, the author still believes both in the need for nuclear reactors and in the need of a thorough-going, pervasive emphasis on their safety. Real, understandable safety can be achieved, and that achievement is the key to our nuclear future. The details he gives are only examples. The need for reactors that are not only safe but obviously safe can be ignored only at our peril

Health requirements for nuclear reactor operators

International Nuclear Information System (INIS)

1980-05-01

The health prerequisites established for the qualification of nuclear reactor operators according to CNEN-NE-1.01 Guidelines Licensing of nuclear reactor operators, CNEN-12/79 Resolution, are described. (M.A.) [pt

Technique of nuclear reactors controls; Technique des controles des reacteurs nucleaires

Energy Technology Data Exchange (ETDEWEB)

Weill, J [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

1953-12-15

This report deal about 'Techniques of control of the nuclear reactors' in the goal to achieve the control of natural uranium reactors and especially the one of Saclay. This work is mainly about the measurement into nuclear parameters and go further in the measurement of thermodynamic variables,etc... putting in relief the new features required on behalf of the detectors because of their use in the thermal neutrons flux. In the domain of nuclear measurement, we indicate the realizations and the results obtained with thermal neutron detectors and for the measurement of ionizations currents. We also treat the technical problem of the start-up of a reactor and of the reactivity measurement. We give the necessary details for the comprehension of all essential diagrams and plans put on, in particular, for the reactor of Saclay. (author) [French] Nous avons aborde le probleme de la ''Technique du Controle des reacteurs nucleaires'' dans le but de realiser le controle du reacteur de Saclay. C'est ainsi que nous avons ete amene a etudier le probleme dans son ensemble, tel qu'il se pose pour tout reacteur a uranium naturel. Ce travail traite principalement du domaine des mesures a caractere nucleaire et s'etend dans le domaine des mesures thermodynamque de niveaux, etc... mettant en relief les caracteristiques nouvelles exigees de la part des detecteurs du fait de leur utilisation dans le flux de neutrons thermiques. Dans le domaine de mesures nucleaires, nous indiquons principalement les realisations et les resultats obtenus pour les detecteurs de neutrons thermiques et pour la mesure de courants d'ionisations. Nous traitons egalement du probleme technique du demarrage d'un reacteur et du probleme de la mesure de la reactivite. Nous donnons les details necessaires a la comrehension de tous les schemas et plans de cablages essentiels mis au point, en particulier, pour le reacteur de Saclay. (auteur)

System for bearing a nuclear reactor vessel cooled by liquid metal

International Nuclear Information System (INIS)

Mahe, A.; Jullien, G.

1976-01-01

The invention relates to a bearing system for supporting a nuclear reactor vessel of the kind which is suspended from the reactor closure slab. The bearing system comprises a ring connected at one end to a collar and at the other end to two collars. The collar connected to the bottom end of the ring forms the top part of the vessel to be supported while the other two collars fit into the slab at two separate places. The ring and collars are disposed in an annular space formed in the slab and dividing it into two parts, i.e., a central part and a peripheral part surrounding the central part of the slab

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)

Improvements in streaking nuclear reactors

International Nuclear Information System (INIS)

Pedrick, A.P.

1976-01-01

In this type of reactor atomic nuclei are stripped of their electron shells by heating to form a very high temperature plasma which is passed at high speed through a chamber in which they are forced into contact with a 'wall' formed by a unidirectional stream of photons from continuous laser beams. In this way it should be possible to brush off from the surface of the nuclei protons and neutrons, with release of their binding energy. The energy thus produced can be subjected to much more gentle control than with a fission or fusion reactor. Furthermore, if this concept can be successfully applied to elements of high atomic number which are normally regarded as stable and unfissionable, a vast new source of nuclear energy release will have been made available. It also seems possible that an atomic nucleus might be spun sufficiently in such a reactor to disintegrate it completely into nucleons by simple centrifugal action, with great release of binding energy. The reactor described has a central body with radial ducts through which the nuclei are passed, and a number of lasers are provided whose beams are arranged so that the nuclei are discharged at the cross-over point of two or more laser beams which form a corner at the junction of two or more photon walls. (U.K.)

Technique of nuclear reactors controls

International Nuclear Information System (INIS)

Weill, J.

1953-12-01

This report deal about 'Techniques of control of the nuclear reactors' in the goal to achieve the control of natural uranium reactors and especially the one of Saclay. This work is mainly about the measurement into nuclear parameters and go further in the measurement of thermodynamic variables,etc... putting in relief the new features required on behalf of the detectors because of their use in the thermal neutrons flux. In the domain of nuclear measurement, we indicate the realizations and the results obtained with thermal neutron detectors and for the measurement of ionizations currents. We also treat the technical problem of the start-up of a reactor and of the reactivity measurement. We give the necessary details for the comprehension of all essential diagrams and plans put on, in particular, for the reactor of Saclay. (author) [fr

Selection of nuclear reactor coolant materials

International Nuclear Information System (INIS)

Shi Lisheng; Wang Bairong

2012-01-01

Nuclear material is nuclear material or materials used in nuclear industry, the general term, it is the material basis for the construction of nuclear power, but also a leader in nuclear energy development, the two interdependent and mutually reinforcing. At the same time, nuclear materials research, development and application of the depth and breadth of science and technology reflects a nation and the level of the nuclear power industry. Coolant also known as heat-carrier agent, is an important part of the heart nuclear reactor, its role is to secure as much as possible to the economic output in the form fission energy to heat the reactor to be used: the same time cooling the core, is controlled by the various structural components allowable temperature. This paper described the definition of nuclear reactor coolant and characteristics, and then addressed the requirements of the coolant material, and finally were introduced several useful properties of the coolant and chemical control. (authors)

Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

International Nuclear Information System (INIS)

Friess, Friederike Renate

2017-01-01

According to a many publications and discussions, fast reactors hold promises to improve safety, non-proliferation, economic aspects, and reduce the nuclear waste problems. Consequently, several reactor designs advocated by the Generation IV Forum are fast reactors. In reality, however, after decades of research and development and billions of dollars investment worldwide, there are only two fast breeders currently operational on a commercial basis: the Russian reactors BN-600 and BN-800. Energy generation alone is apparently not a sufficient selling point for fast breeder reactors. Therefore, other possible applications for fast nuclear reactors are advocated. Three relevant examples are investigated in this thesis. The first one is the disposition of excess weapon-grade plutonium. Unlike for high enriched uranium that can be downblended for use in light water reactors, there exists no scientifically accepted solution for the disposition of weapon-grade plutonium. One option is the use in fast reactors that are operated for energy production. In the course of burn-up, the plutonium is irradiated which intends to fulfill two objectives: the resulting isotopic composition of the plutonium is less suitable for nuclear weapons, while at the same time the build-up of fission products results in a radiation barrier. Appropriate reprocessing technology is in order to extract the plutonium from the spent fuel. The second application is the use as so-called nuclear batteries, a special type of small modular reactors (SMRs). Nuclear batteries offer very long core lifetimes and have a very small energy output of sometimes only 10 MWe. They can supposedly be placed (almost) everywhere and supply energy without the need for refueling or shuffling of fuel elements for long periods. Since their cores remain sealed for several decades, nuclear batteries are claimed to have a higher proliferation resistance. The small output and the reduced maintenance and operating requirements

Neutron-physical simulation of fast nuclear reactor cores. Investigation of new and emerging nuclear reactor systems

Energy Technology Data Exchange (ETDEWEB)

Friess, Friederike Renate

2017-07-12

According to a many publications and discussions, fast reactors hold promises to improve safety, non-proliferation, economic aspects, and reduce the nuclear waste problems. Consequently, several reactor designs advocated by the Generation IV Forum are fast reactors. In reality, however, after decades of research and development and billions of dollars investment worldwide, there are only two fast breeders currently operational on a commercial basis: the Russian reactors BN-600 and BN-800. Energy generation alone is apparently not a sufficient selling point for fast breeder reactors. Therefore, other possible applications for fast nuclear reactors are advocated. Three relevant examples are investigated in this thesis. The first one is the disposition of excess weapon-grade plutonium. Unlike for high enriched uranium that can be downblended for use in light water reactors, there exists no scientifically accepted solution for the disposition of weapon-grade plutonium. One option is the use in fast reactors that are operated for energy production. In the course of burn-up, the plutonium is irradiated which intends to fulfill two objectives: the resulting isotopic composition of the plutonium is less suitable for nuclear weapons, while at the same time the build-up of fission products results in a radiation barrier. Appropriate reprocessing technology is in order to extract the plutonium from the spent fuel. The second application is the use as so-called nuclear batteries, a special type of small modular reactors (SMRs). Nuclear batteries offer very long core lifetimes and have a very small energy output of sometimes only 10 MWe. They can supposedly be placed (almost) everywhere and supply energy without the need for refueling or shuffling of fuel elements for long periods. Since their cores remain sealed for several decades, nuclear batteries are claimed to have a higher proliferation resistance. The small output and the reduced maintenance and operating requirements

Technology of nuclear reactors

International Nuclear Information System (INIS)

Ravelet, F.

2016-01-01

This academic report for graduation in engineering first presents operation principles of a nuclear reactor core. It presents core components, atomic nuclei, the notions of transmutation and radioactivity, quantities used to characterize ionizing radiations, the nuclear fission, statistical aspects of fission and differences between fast and slow neutrons, a comparison between various heat transfer fluids, the uranium enrichment process, and different types of reactor (boiling water, natural uranium and heavy water, pressurized water, and fourth generation). Then, after having recalled the French installed power, the author proposes an analysis of a typical 900 MWe nuclear power plant: primary circuit, reactor, fuel, spent fuel, pressurizer and primary pump, secondary circuit, aspects related to control-command, regulation, safety and exploitation. The last part proposes a modelling of the thermodynamic cycle of a pressurized water plant by using an equivalent Carnot cycle, a Rankine cycle, and a two-phase expansion cycle with drying-overheating

Reactors. Nuclear propulsion ships

International Nuclear Information System (INIS)

Fribourg, Ch.

2001-01-01

This article has for object the development of nuclear-powered ships and the conception of the nuclear-powered ship. The technology of the naval propulsion P.W.R. type reactor is described in the article B.N.3 141 'Nuclear Boilers ships'. (N.C.)

Neutron dosimetry. Environmental monitoring in a BWR type reactor; Dosimetria de neutrones. Monitoreo ambiental en un reactor del tipo BWR

Energy Technology Data Exchange (ETDEWEB)

Tavera D, L; Camacho L, M E

1991-01-15

The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

The failure diagnoses of nuclear reactor systems

International Nuclear Information System (INIS)

Sheng Huanxing.

1986-01-01

The earlier period failure diagnoses can raise the safety and efficiency of nuclear reactors. This paper first describes the process abnormality monitoring of core barrel vibration in PWR, inherent noise sources in BWR, sodium boiling in LMFBR and nuclear reactor stability. And then, describes the plant failure diagnoses of primary coolant pumps, loose parts in nuclear reactors, coolant leakage and relief valve location

Applications in nuclear data and reactor physics

International Nuclear Information System (INIS)

Cullen, D.E.; Muranaka, R.; Schmidt, J.

1986-01-01

This book presents the papers given at a conference on reactor kinetics and nuclear data collections. Topics considered at the conference included nuclear data processing, PWR core design calculations, reactor neutron dosimetry, in-core fuel management, reactor safety analysis, transients, two-phase flow, fuel cycles of research reactors, slightly enriched uranium, highly enriched uranium, reactor start-up, computer codes, and the transport of spent fuel elements

Three dimensional diffusion calculations of nuclear reactors

International Nuclear Information System (INIS)

Caspo, N.

1981-07-01

This work deals with the three dimensional calculation of nuclear reactors using the code TRITON. The purposes of the work were to perform three-dimensional computations of the core of the Soreq nuclear reactor and of the power reactor ZION and to validate the TRITON code. Possible applications of the TRITON code in Soreq reactor calculations and in power reactor research are suggested. (H.K.)

Nuclear waste management, reactor decommisioning, nuclear liability and public attitudes

International Nuclear Information System (INIS)

Green, R.E.

1982-01-01

This paper deals with several issues that are frequently raised by the public in any discussion of nuclear energy, and explores some aspects of public attitudes towards nuclear-related activities. The characteristics of the three types of waste associated with the nuclear fuel cycle, i.e. mine/mill tailings, reactor wastes and nuclear fuel wastes, are defined, and the methods currently being proposed for their safe handling and disposal are outlined. The activities associated with reactor decommissioning are also described, as well as the Canadian approach to nuclear liability. The costs associated with nuclear waste management, reactor decommissioning and nuclear liability are also discussed. Finally, the issue of public attitudes towards nuclear energy is addressed. It is concluded that a simple and comprehensive information program is needed to overcome many of the misconceptions that exist about nuclear energy and to provide the public with a more balanced information base on which to make decisions

Nuclear reactor with a suspended vessel

International Nuclear Information System (INIS)

Lemercier, Guy.

1977-01-01

This invention relates to a nuclear reactor with a suspended vessel and applies in particular when this is a fast reactor, the core or active part of the reactor being inside the vessel and immersed under a suitable volume of flowing liquid metal to cool it by extracting the calories released by the nuclear fission in the fuel assemblies forming this core [fr

Control for nuclear reactor

International Nuclear Information System (INIS)

Ash, E.B.; Bernath, L.; Facha, J.V.

1980-01-01

A nuclear reactor is provided with several hydraulically-supported spherical bodies having a high neutron absorption cross section, which fall by gravity into the core region of the reactor when the flow of supporting fluid is shut off. (auth)

Overview of Nuclear Reactor Technologies Portfolio

International Nuclear Information System (INIS)

O’Connor, Thomas J.

2012-01-01

Office of Nuclear Energy Roadmap R&D Objectives: • Develop technologies and other solutions that can improve the reliability, sustain the safety, and extend the life of current reactors; • Develop improvements in the affordability of new reactors to enable nuclear energy to help meet the Administration's energy security and climate change goals; • Develop sustainable nuclear fuel cycles; • Develop capabilities to reduce the risks of nuclear proliferation and terrorism

GE's advanced nuclear reactor designs

International Nuclear Information System (INIS)

Berglund, R.C.

1993-01-01

The excess of US electrical generating capacity which has existed for the past 15 years is coming to an end as we enter the 1990s. Environmental and energy security issues associated with fossil fuels are kindling renewed interest in the nuclear option. The importance of these issues are underscored by the National Energy Strategy (NES) which calls for actions which open-quotes are designed to ensure that the nuclear power option is available to utilities.close quotes Utilities, utility associations, and nuclear suppliers, under the leadership of the Nuclear Power Oversight Committee (NPOC), have jointly developed a 14-point strategic plan aimed at establishing a predictable regulatory environment, standardized and pre-licensed Advanced Light Water Reactor (ALWR) nuclear plants, resolving the long-term waste management issue, and other open-quotes enabling conditions.close quotes GE is participating in this national effort and GE's family of advanced nuclear power plants feature two reactor designs, developed on a common technology base, aimed at providing a new generation of nuclear plants to provide safe, clean, economical electricity to the world's utilities in the 1990s and beyond. Together, the large-size (1300 MWe) Advanced Boiling Water Reactor (ABWR) and the small-size (600 MWe) Simplified Boiling Water Reactor (SBWR) are innovative, near-term candidates for expanding electrical generating capacity in the US and worldwide. Both possess the features necessary to do so safety, reliably, and economically

Reactor Dosimetry State of the Art 2008

Science.gov (United States)

Voorbraak, Wim; Debarberis, Luigi; D'Hondt, Pierre; Wagemans, Jan

2009-08-01

Oral session 1: Retrospective dosimetry. Retrospective dosimetry of VVER 440 reactor pressure vessel at the 3rd unit of Dukovany NPP / M. Marek ... [et al.]. Retrospective dosimetry study at the RPV of NPP Greifswald unit 1 / J. Konheiser ... [et al.]. Test of prototype detector for retrospective neutron dosimetry of reactor internals and vessel / K. Hayashi ... [et al.]. Neutron doses to the concrete vessel and tendons of a magnox reactor using retrospective dosimetry / D. A. Allen ... [et al.]. A retrospective dosimetry feasibility study for Atucha I / J. Wagemans ... [et al.]. Retrospective reactor dosimetry with zirconium alloy samples in a PWR / L. R. Greenwood and J. P. Foster -- Oral session 2: Experimental techniques. Characterizing the Time-dependent components of reactor n/y environments / P. J. Griffin, S. M. Luker and A. J. Suo-Anttila. Measurements of the recoil-ion response of silicon carbide detectors to fast neutrons / F. H. Ruddy, J. G. Seidel and F. Franceschini. Measurement of the neutron spectrum of the HB-4 cold source at the high flux isotope reactor at Oak Ridge National Laboratory / J. L. Robertson and E. B. Iverson. Feasibility of cavity ring-down laser spectroscopy for dose rate monitoring on nuclear reactor / H. Tomita ... [et al.]. Measuring transistor damage factors in a non-stable defect environment / D. B. King ... [et al.]. Neutron-detection based monitoring of void effects in boiling water reactors / J. Loberg ... [et al.] -- Poster session 1: Power reactor surveillance, retrospective dosimetry, benchmarks and inter-comparisons, adjustment methods, experimental techniques, transport calculations. Improved diagnostics for analysis of a reactor pulse radiation environment / S. M. Luker ... [et al.]. Simulation of the response of silicon carbide fast neutron detectors / F. Franceschini, F. H. Ruddy and B. Petrović. NSV A-3: a computer code for least-squares adjustment of neutron spectra and measured dosimeter responses / J. G

Neutron noise in nuclear reactors

International Nuclear Information System (INIS)

Blaquiere, A.; Pachowska, R.

1961-06-01

The power of a nuclear reactor, in the operating conditions, presents fluctuations due to various causes. This random behaviour can be included in the study of 'noises'. Among other sources of noise, we analyse hereafter the fluctuations due: a) to the discontinuous emissions of neutrons from an independent source; b) to the multiplication of neutrons inside the reactor. The method which we present makes use of the analogies between the rules governing a nuclear reactor in operation and a number of radio-electrical systems, in particular the feed-back loops. The reactor can be characterized by its 'passing band' and is described as a system submitted to a sequence of random pulses. In non linear operating condition, the effect of neutron noise is defined by means of a non-linear functional, this theory is thus related to previous works the references of which are given at the end of the present report. This leads us in particular in the case of nuclear reactors to some results given by A. Blaquiere in the case of radio-electrical loops. (author) [fr

New materials in nuclear fusion reactors

International Nuclear Information System (INIS)

Iwata, Shuichi

1988-01-01

In the autumn of 1987, the critical condition was attained in the JET in Europe and Japanese JT-60, thus the first subject in the physical verification of nuclear fusion reactors was resolved, and the challenge to the next attainment of self ignition condition started. As the development process of nuclear fusion reactors, there are the steps of engineering, economical and social verifications after this physical verification, and in respective steps, there are the critical problems related to materials, therefore the development of new materials must be advanced. The condition of using nuclear fusion reactors is characterized by high fluence, high thermal flux and strong magnetic field, and under such extreme condition, the microscopic structures of materials change, and they behave much differently from usual case. The subjects of material development for nuclear fusion reactors, the material data base being built up, the materials for facing plasma and high thermal flux, first walls, blanket structures, electric insulators and others are described. The serious effect of irradiation and the rate of defect inducement must be taken in consideration in the structural materials for nuclear fusion reactors. (Kako, I.)

Liquid metal cooled nuclear reactor

International Nuclear Information System (INIS)

Guidez, Joel; Jarriand, Paul.

1975-01-01

The invention concerns a fast neutron nuclear reactor cooled by a liquid metal driven through by a primary pump of the vertical drive shaft type fitted at its lower end with a blade wheel. To each pump is associated an exchanger, annular in shape, fitted with a central bore through which passes the vertical drive shaft of the pump, its wheel being mounted under the exchanger. A collector placed under the wheel comprises an open upward suction bell for the liquid metal. A hydrostatic bearing is located above the wheel to guide the drive shaft and a non detachable diffuser into which at least one delivery pipe gives, envelopes the wheel [fr

Chapter 12. Nullification of nuclear reactors

International Nuclear Information System (INIS)

Toelgyessy, J.; Harangozo, M.

2000-01-01

This is a chapter of textbook of radioecology for university students. In this chapter authors deal with problems connected with nullification of nuclear reactors. There are tree basic methods of nullification of nuclear reactors: (1) conservation, (2) safe close (wall up, embed in concrete), (3) direct dismantlement and remotion and two combined ways: (1) combination of mothball with subsequent dismantlement and remotion and (2) combination of safe close with subsequent dismantlement and remotion. Activity levels as well as volumes of radioactive wastes connected with decommissioning of nuclear reactors are reviewed

EURATOM's Programme of Participation in Power Reactor Construction; Le programme de participation d'Euratom aux reacteurs de puissance; Programma uchastiya v razrabotke ehnergeticheskikh reaktorov Evratoma; El programa de participacion de la Euratom en la construccion y explotacion de reactores de potencia

Energy Technology Data Exchange (ETDEWEB)

Ramadier, R. C.; Parker, E. [Communaute Europoenne de l' Energie Atomique, Bruxelles (Belgium)

1963-10-15

annees de leur utilisation. Cette participation de l'Euratom a eu pour effet d'encourager la construction de certaines de ces centrales nucleaires. Elle a permis en outre et permettra encore d'acquerir des informations extremement utiles au cours des annees qui viennent, annees au cours desquelles les problemes de fonctionnement se poseront d'une facon decisive pour le developpement de l'energie atomique. (author) [Spanish] Para fomentar e l desarrollo de una industria nuclear europea, la Comision de la EURATOM utiliza, entre otros medios, su programa de ''participacion mancomunada'', que le permite cooperar con aportaciones de 32 millones de unidades de cuenta AME como maximo, en la construccion y explotacion de reactores de potencia. Como contrapartida, obtiene informaciones sobre el diseno, la construccion, la puesta en marcha y el funcionamiento de esos reactores. Hasta hoy la Comision ha firmado contratos con tres empresas: a) la Societa Elettronucleare Nazionale (SENN), que hace construir en Italia una central de 150 MW(e) netos equipada con un reactor de agua hirviente de ciclo doble; b) la Societa Italiana Meridionale Energia Atomica (SIMEA), que ha emprendido en Italia la construccion de una central de 200 MW(e) netos equipada con un reactor de uranio natural-grafito-anhidrido carbonico; c) la Societe d'Energie Nucleaire Franco-Belge des Ardennes (SENA), que ha emprendido en la frontera franco-belga la construccion de una central equipada con un reactor de agua a presion que podra alcanzar - y probablemente superar - una potencia neta de 242 MW(e). La Comision ha recibido, ademas, una solicitud del grupo Rheinisch-Westfalisches Elektrizitetswerk- Bayernwerke (RWE-BW) y otra de la N.V. Samenwerkende Electriciteits-Productiebedrijven para participar en la construccion y explotacion de otros dos reactores de potencia: la primera se refiere a un reactor de 237 MW(e) de agua hirviente y de ciclo doble, y la segunda a un reactor de 50 MW(e) de agua hirviente, de ciclo

Nuclear power reactors of new generation

International Nuclear Information System (INIS)

Ponomarev-Stepnoi, N.N.; Slesarev, I.S.

1988-01-01

The paper presents discussions on the following topics: fuel supply for nuclear power; expansion of the sphere of nuclear power applications, such as district heating; comparative estimates of power reactor efficiencies; safety philosophy of advanced nuclear plants, including passive protection and inherent safety concepts; nuclear power unit of enhanced safety for the new generation of nuclear power plants. The emphasis is that designers of new generation reactors face a complicated but technically solvable task of developing highly safe, efficient, and economical nuclear power sources having a wide sphere of application

Nuclear reactor shutdown system

International Nuclear Information System (INIS)

Mangus, J.D.; Cooper, M.H.

1982-01-01

An improved nuclear reactor shutdown system is described comprising a temperature sensitive device connected to control the electric power supply to a magnetic latch holding a body of a neutron absorbing material. The temperature sensitive device is exposed to the reactor coolant so that when the reactor coolant temperature rises above a specific level, the temperature sensitive device will cause deenergization of the magnetic latch to allow the body of neutron absorbing material to enter the reactor core. (author)

Nuclear reactor design

CERN Document Server

2014-01-01

This book focuses on core design and methods for design and analysis. It is based on advances made in nuclear power utilization and computational methods over the past 40 years, covering core design of boiling water reactors and pressurized water reactors, as well as fast reactors and high-temperature gas-cooled reactors. The objectives of this book are to help graduate and advanced undergraduate students to understand core design and analysis, and to serve as a background reference for engineers actively working in light water reactors. Methodologies for core design and analysis, together with physical descriptions, are emphasized. The book also covers coupled thermal hydraulic core calculations, plant dynamics, and safety analysis, allowing readers to understand core design in relation to plant control and safety.

Materials for nuclear reactors

International Nuclear Information System (INIS)

Banerjee, S.; Kamath, H.S.

2005-01-01

The improved performance of present generation nuclear reactors and the realization of advanced reactor concepts, both, require development of better materials. Physical metallurgy/materials science principles which have been exploited in meeting the exacting requirements of nuclear reactor materials (fuels and structural materials), are outlined citing a few specific examples. While the incentive for improvement of traditional fuels (e.g., UO 2 fuel) is primarily for increasing the average core burn up, the development of advanced fuels (e.g., MOX, mixed carbide, nitride, silicide and dispersion fuels) are directed towards better utilization of fissile and fertile inventories through adaptation of innovative fuel cycles. As the burn up of UO 2 fuel reaches higher levels, a more detailed and quantitative understanding of the phenomena such as fission gas release, fuel restructuring induced by radiation and thermal gradients and pellet-clad interaction is being achieved. Development of zirconium based alloys for both cladding and pressure tube applications is discussed with reference to their physical metallurgy, fabrication techniques and in-reactor degradation mechanisms. The issue of radiation embrittlement of reactor pressure vessels (RPVs) is covered drawing a comparison between the western and eastern specifications of RPV steels. The search for new materials which can stand higher rates of atomic displacement due to radiation has led to the development of swelling resistant austenitic and ferritic stainless steels for fast reactor applications as exemplified by the development of the D-9 steel for Indian fast breeder reactor. The presentation will conclude by listing various materials related phenomena, which have a strong bearing on the successful development of future nuclear energy systems. (author)

Reactor use in nuclear engineering programs

International Nuclear Information System (INIS)

Murray, R.L.

1975-01-01

Nuclear reactors for dual use in training and research were established at about 50 universities in the period since 1950, with assistance by the U. S. Atomic Energy Commission and the National Science Foundation. Most of the reactors are in active use for a variety of educational functions--laboratory teaching of undergraduates and graduate students, graduate research, orientation of visitors, and nuclear power plant reactor operator training, along with service to the technical community. As expected, the higher power reactors enjoy a larger average weekly use. Among special programs are reactor sharing and high-school teachers' workshops

Random processes in nuclear reactors

CERN Document Server

Williams, M M R

1974-01-01

Random Processes in Nuclear Reactors describes the problems that a nuclear engineer may meet which involve random fluctuations and sets out in detail how they may be interpreted in terms of various models of the reactor system. Chapters set out to discuss topics on the origins of random processes and sources; the general technique to zero-power problems and bring out the basic effect of fission, and fluctuations in the lifetime of neutrons, on the measured response; the interpretation of power reactor noise; and associated problems connected with mechanical, hydraulic and thermal noise sources

Design of radiation shields in nuclear reactor core

International Nuclear Information System (INIS)

Mousavi Shirazi, A.; Daneshvar, Sh.; Aghanajafi, C.; Jahanfarnia, Gh.; Rahgoshay, M.

2008-01-01

This article consists of designing radiation shields in the core of nuclear reactors to control and restrain the harmful nuclear radiations in the nuclear reactor cores. The radiation shields protect the loss of energy. caused by nuclear radiation in a nuclear reactor core and consequently, they cause to increase the efficiency of the reactor and decrease the risk of being under harmful radiations for the staff. In order to design these shields, by making advantages of the O ppenheim Electrical Network m ethod, the structure of the shields are physically simulated and by obtaining a special algorithm, the amount of optimized energy caused by nuclear radiations, is calculated

Dangerous liaisons: Western involvement in the nuclear power industry of central and eastern Europe

International Nuclear Information System (INIS)

1992-12-01

The state of the nuclear industry in central and eastern Europe is an issue of global concern. However, despite all the political talk and corporate hype since the collapse of communist regimes in the region, this study demonstrates that little has so far been done to change the situation. Moreover, the limited level of finance and support which has been offered has tended to support the expansion of nuclear power programmes in central and eastern Europe, rather than address immediately safety concerns relating to existing nuclear reactors and develop more environmentally acceptable and economically efficient energy systems. (author)

Method for operating nuclear reactor

International Nuclear Information System (INIS)

Utamura, Motoaki; Urata, Megumu; Uchida, Shunsuke

1978-01-01

Purpose: In order to judge the fuel failures, if any, without opening a reactor container for BWR type reactors, a method has been described for measuring the difference between the temperature dependent iodine spike value and the pressure dependent iodine spike value in the pressure vessel. Method: After the scram of a nuclear reactor, steam generated by decay heat is condensed in a remaining heat exchanger and cooling water is returned through a recycling pipe line to a reactor core. At the same time, a control rod drive system pump is operated, the reactor core is filled with the cooling water. Then, the coolant is taken from the recycling pipe line to cool the reactor core. After applying the temperature fluctuation, the cooling water is sampled at a predetermined time interval at a sampling point to determine the changes with time in the radioactive concentration of iodine. When the radioactivity of iodine in the cooling water is lowered sufficiently by a reactor purifying system, the nuclear reactor vessel is depressurized. After applying pressure fluctuation, iodine spike value is determined. (Kawakami, Y.)

The Design of a Nuclear Reactor

Indian Academy of Sciences (India)

The aim of this largely pedagogical article is toemploy pre-college physics to arrive at an understanding of a system as complex as a nuclear reactor. We focus on three key issues: the fuelpin, the moderator, and lastly the dimensions ofthe nuclear reactor.

Applications of computational intelligence in nuclear reactors

International Nuclear Information System (INIS)

Jayalal, M.L.; Jehadeesan, R.

2016-01-01

Computational intelligence techniques have been successfully employed in a wide range of applications which include the domains of medical, bioinformatics, electronics, communications and business. There has been progress in applying of computational intelligence in the nuclear reactor domain during the last two decades. The stringent nuclear safety regulations pertaining to reactor environment present challenges in the application of computational intelligence in various nuclear sub-systems. The applications of various methods of computational intelligence in the domain of nuclear reactors are discussed in this paper. (author)

Nuclear reactor

International Nuclear Information System (INIS)

Sasaki, Tomozo.

1987-01-01

Purpose: To improve the nuclear reactor availability by enabling to continuously exchange fuels in the natural-slightly enriched uranium region during operation. Constitution: A control rod is withdrawn to the midway of a highly enriched uranium region by means of control rod drives and the highly enriched uranium region is burnt to maintain the nuclear reactor always at a critical state. At the same time, fresh uranium-slightly enriched uranium is continuously supplied gravitationally from a fresh fuel reservoir through fuel reservoir to each of fuel pipes in the natural-slightly enriched uranium region. Then, spent fuels reduced with the reactivity by the burn up are successively taken out from the bottom of each of the fuel pipes through an exit duct and a solenoid valve to the inside of a spent fuel reservoir and the burn up in the natural-slightly enriched uranium region is conducted continuously. (Kawakami, Y.)

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)

Nuclear reactor

International Nuclear Information System (INIS)

Scholz, M.

1976-01-01

An improvement of the accessibility of that part of a nuclear reactor serving for biological shield is proposed. It is intended to provide within the biological shield, distributed around the circumference of the reactor pressure vessel, several shielding chambers filled with shielding material, which are isolated gastight from the outside by means of glass panes with a given bursting strength. It is advantageous that, on the one hand, inspection and maintenance will be possible without great effort and, on the other, a large relief cross section will be at desposal if required. (UWI) [de

Nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Prescott, R F; George, B V; Baglin, C J

1978-05-10

Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given.

Nuclear reactors

International Nuclear Information System (INIS)

Prescott, R.F.; George, B.V.; Baglin, C.J.

1978-01-01

Reference is made to thermal insulation on the inner surfaces of containment vessels of fluid cooled nuclear reactors and particularly in situations where the thermal insulation must also serve a structural function and transmit substantial load forces to the surface which it covers. An arrangement is described that meets this requirement and also provides for core support means that favourably influences the flow of hot coolant from the lower end of the core into a plenum space in the hearth of the reactor. The arrangement comprises a course of thermally insulating bricks arranged as a mosaic covering a wall of the reactor and a course of thermally insulating tiles arranged as a mosaic covering the course of bricks. Full constructional details are given. (UK)

On exposure of workers in nuclear reactor facilities for test and in nuclear reactor facilities in research and development stage in fiscal 1988

International Nuclear Information System (INIS)

1989-01-01

The Law for Regulation on Nuclear Reactor requires the operators of nuclear reactors that the exposure dose of workers engaged in work for nuclear reactors should not exceed the limits specified in official notices that are issued based on the Law. The present article summarizes the contents of the Report on Radiation Management in 1988 submitted by the operators of nuclear reactor facilities for test and those of nuclear reactor facilities in research and development stage based on the Law, and the Report on Management of Exposure Dose of Workers submitted by them based on administrative notices. The reports demonstrate that the exposure of workers was below the permissible exposure dose in 1988 in all nuclear reactor facilities. The article presents data on the distribution of exposure dose among workers in all facilities with a nuclear reactor for test, and data on personal exposure of employees and non-employees and overall exposure of all workers in the facilities of Japan Atomic Energy Research Institute and Power Reactor and Nuclear Fuel Development Corporation. (N.K.)

Nuclear Power Reactors in the World. 2013 Ed

International Nuclear Information System (INIS)

2013-01-01

Nuclear Power Reactors in the World is an annual publication that presents the most recent data pertaining to nuclear power reactors in IAEA Member States. This thirty-third edition of Reference Data Series No. 2 provides a detailed comparison of various statistics through 31 December 2012. The tables and figures contain the following information: - General statistics on nuclear reactors in IAEA Member States; - Technical data on specific reactors that are either planned, under construction or operational, or that have been shut down or decommissioned; - Performance data on reactors operating in IAEA Member States, as reported to the IAEA. The data compiled in this publication is a product of the IAEA's Power Reactor Information System (PRIS). The PRIS database is a comprehensive source of data on all nuclear power reactors in the world. It includes specification and performance history data on operational reactors as well as on reactors under construction or in the decommissioning process. The IAEA collects data through designated national correspondents in Member States

Nuclear instrumentation for research reactors; Instrumentacion nuclear para reactores nucleares de investigacion

Energy Technology Data Exchange (ETDEWEB)

Hofer, Carlos G.; Pita, Antonio; Verrastro, Claudio A.; Maino, Eduardo J. [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Unidad de Actividades de Reactores y Centrales Nucleares. Sector Instrumentacion y Control

1997-10-01

The nuclear instrumentation for research reactors in Argentina was developed in 70`. A gradual modernization of all the nuclear instrumentation is planned. It includes start-up and power range instrumentation, as well as field monitors, clamp, scram and rod movement control logic. The new instrumentation is linked to a computer network, based on real time operating system for data acquisition, display and logging. This paper describes the modules and whole system aspects. (author). 2 refs.

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

Desalination of seawater with nuclear power reactors in cogeneration; Desalacion de agua de mar con reactores nucleares de potencia en cogeneracion

Energy Technology Data Exchange (ETDEWEB)

Flores E, R M

2004-07-01

The growing demand for energy and hydraulic resources for satisfy the domestic, industrial, agricultural activities, etc. has wakened up the interest to carry out concerning investigations to study the diverse technologies guided to increase the available hydraulic resources, as well as to the search of alternatives of electric power generation, economic and socially profitable. In this sense the possible use of the nuclear energy is examined in cogeneration to obtain electricity and drinkable water for desalination of seawater. The technologies are analysed involved in the nuclear cogeneration (desalination technology, nuclear and desalination-nuclear joining) available in the world. At the same time it is exemplified the coupling of a nuclear reactor and a process of hybrid desalination that today in day the adult offers and economic advantages. Finally, the nuclear desalination is presented as a technical and economically viable solution in regions where necessities of drinkable water are had for the urban, agricultural consumption and industrial in great scale and that for local situations it is possible to satisfy it desalinating seawater. (Author)

Nuclear reactor kinetics and plant control

CERN Document Server

Oka, Yoshiaki

2013-01-01

Understanding time-dependent behaviors of nuclear reactors and the methods of their control is essential to the operation and safety of nuclear power plants. This book provides graduate students, researchers, and engineers in nuclear engineering comprehensive information on both the fundamental theory of nuclear reactor kinetics and control and the state-of-the-art practice in actual plants, as well as the idea of how to bridge the two. The first part focuses on understanding fundamental nuclear kinetics. It introduces delayed neutrons, fission chain reactions, point kinetics theory, reactivit

Change of nuclear reactor installation in the first nuclear ship of Japan Nuclear Ship Development Agency

International Nuclear Information System (INIS)

1979-01-01

The written application concerning the change of nuclear reactor installation in the first nuclear ship was presented from the JNSDA to the prime minister on January 10, 1979. The contents of the change are the repair of the primary and secondary shields of the reactor, the additional installation of a storage tank for liquid wastes, and the extension of the period to stop the reactor in cold state. The inquiry from the prime minister to the Nuclear Safety Commission was made on June 9, 1979, through the examination of safety in the Nuclear Safety Bureau, Science and Technology Agency. The Nuclear Safety Commission instructed to the Committee for the Examination of Nuclear Reactor Safety on June 11, 1979, about the application of criteria stipulated in the law. The relevant letters and the drafts of examination papers concerning the technical capability and the safety in case of the change of nuclear reactor installation in the first nuclear ship are cited. The JNSDA and Sasebo Heavy Industries, Ltd. seem to have the sufficient technical capability to carry out this change. As the result of examination, it is recognized that the application presented by the JNSDA is in compliance with the criteria stipulated in the law concerning the regulation of nuclear raw materials, nuclear fuel materials and nuclear reactors. (Kako, I.)

Nuclear reactors built, being built, or planned, 1988

International Nuclear Information System (INIS)

1989-08-01

This document contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1988. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually from Washington Headquarters and field offices of DOE, from the US Nuclear Regulatory Commission, from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations, from US and foreign embassies, and from foreign governmental nuclear departments. The book is divided into three major sections: Section 1 consists of a reactor locator map and reactor tables. Section 2 includes nuclear reactors that are operating, being built, or planned. Section 3 includes reactors that have been shut down permanently or dismantled

Introduction to the neutron kinetics of nuclear power reactors

CERN Document Server

Tyror, J G; Grant, P J

2013-01-01

An Introduction to the Neutron Kinetics of Nuclear Power Reactors introduces the reader to the neutron kinetics of nuclear power reactors. Topics covered include the neutron physics of reactor kinetics, feedback effects, water-moderated reactors, fast reactors, and methods of plant control. The reactor transients following faults are also discussed, along with the use of computers in the study of power reactor kinetics. This book is comprised of eight chapters and begins with an overview of the reactor physics characteristics of a nuclear power reactor and their influence on system design and

Fail-safe reactivity compensation method for a nuclear reactor

Science.gov (United States)

Nygaard, Erik T.; Angelo, Peter L.; Aase, Scott B.

2018-01-23

The present invention relates generally to the field of compensation methods for nuclear reactors and, in particular to a method for fail-safe reactivity compensation in solution-type nuclear reactors. In one embodiment, the fail-safe reactivity compensation method of the present invention augments other control methods for a nuclear reactor. In still another embodiment, the fail-safe reactivity compensation method of the present invention permits one to control a nuclear reaction in a nuclear reactor through a method that does not rely on moving components into or out of a reactor core, nor does the method of the present invention rely on the constant repositioning of control rods within a nuclear reactor in order to maintain a critical state.

Fundamentals of Nuclear Reactor Physics

CERN Document Server

Lewis, E E

2008-01-01

This new streamlined text offers a one-semester treatment of the essentials of how the fission nuclear reactor works, the various approaches to the design of reactors, and their safe and efficient operation. The book includes numerous worked-out examples and end-of-chapter questions to help reinforce the knowledge presented. This textbook offers an engineering-oriented introduction to nuclear physics, with a particular focus on how those physics are put to work in the service of generating nuclear-based power, particularly the importance of neutron reactions and neutron behavior. Engin

Space Nuclear Reactor Engineering

Energy Technology Data Exchange (ETDEWEB)

Poston, David Irvin [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2017-03-06

We needed to find a space reactor concept that could be attractive to NASA for flight and proven with a rapid turnaround, low-cost nuclear test. Heat-pipe-cooled reactors coupled to Stirling engines long identified as the easiest path to near-term, low-cost concept.

Nuclear reactor kinetics and control

International Nuclear Information System (INIS)

Lewins, J.

1978-01-01

A consistent, integrated account of modern developments in the study of nuclear reactor kinetics and the problem of their efficient and safe control. It aims to prepare the student for advanced study and research or practical work in the field. Special features include treatments of noise theory, reliability theory and safety related studies. It covers all aspects of the operation and control of nuclear reactors, power and research and is complete in providing physical data methods of calculation and solution including questions of equipment reliability. The work uses illustrations of the main types of reactors in use in the UK, USA and Europe. Each chapter contains problems and worked examples suitable for course work and study. The subject is covered in chapters, entitled: introductory review; neutron and precursor equations; elementary solutions at low power; linear reactor process dynamics with feedback; power reactor control systems; fluctuations and reactor noise; safety and reliability; nonlinear systems (safety and control); analogue computing. (author)

Nuclear propulsion apparatus with alternate reactor segments

International Nuclear Information System (INIS)

Szekely, T.

1979-01-01

Nuclear propulsion apparatus comprising: (a) means for compressing incoming air; (b) nuclear fission reactor means for heating said air; (c) means for expanding a portion of the heated air to drive said compressing means; (d) said nuclear fission reactor means being divided into a plurality of radially extending segments; (e) means for directing a portion of the compressed air for heating through alternate segments of said reactor means and another portion of the compressed air for heating through the remaining segments of said reactor means; and (f) means for further expanding the heated air from said drive means and the remaining heated air from said reactor means through nozzle means to effect reactive thrust on said apparatus. 12 claims

Prospect of realizing nuclear fusion reactors

International Nuclear Information System (INIS)

1989-01-01

This Report describes the results of the research work on nuclear fusion, which CRIEPI has carried out for about ten years from the standpoint of electric power utilities, potential user of its energy. The principal points are; (a) economic analysis (calculation of costs) based on Japanese analysis procedures and database of commercial fusion reactors, including fusion-fission hybrid reactors, and (b) conceptual design of two types of hybrid reactors, that is, fission-fuel producing DMHR (Demonstration Molten-Salt Hybrid Reactor) and electric-power producing THPR (Tokamak Hybrid Power Reactor). The Report consists of the following chapters: 1. Introduction. 2. Conceptual Design of Hybrid Reactors. 3. Economic Analysis of Commercial Fusion Reactors. 4. Basic Studies Applicable Also to Nuclear Fusion Technology. 5. List of Published Reports and Papers; 6. Conclusion. Appendices. (author)

Small reactors and the 'second nuclear era'

International Nuclear Information System (INIS)

Egan, J.R.

1984-01-01

Predictions of the nuclear industry's demise are premature and distort both history and politics. The industry is reemerging in a form commensurate with the priorities of those people and nations controlling the global forces of production. The current lull in plant orders is due primarily to the world recession and to factors related specifically to reactor size. Traditional economies of scale for nuclear plants have been greatly exaggerated. Reactor vendors and governments in Great Britain, France, West Germany, Japan, the United States, Sweden, Canada, and the Soviet Union are developing small reactors for both domestic applications and export to the Third World. The prefabricated, factory-assembled plants under 500 MWe may alleviate many of the existing socioeconomic constraints on nuclear manufacturing, construction, and operation. In the industrialized world, small reactors could furnish a qualitatively new energy option for utilities. But developing nations hold the largest potential market for small reactors due to the modest size of their electrical systems. These units could double or triple the market potential for nuclear power in this century. Small reactors will both qualitatively and quantitatively change the nature of nuclear technology transfers, offering unique advantages and problems vis-a-vis conventional arrangements. (author)

The Design of a Nuclear Reactor

Indian Academy of Sciences (India)

2016-08-26

Aug 26, 2016 ... The aim of this largely pedagogical article is toemploy pre-college physics to arrive at an understanding of a system as complex as a nuclear reactor. We focus on three key issues: the fuelpin, the moderator, and lastly the dimensions ofthe nuclear reactor.

HOMOGENEOUS NUCLEAR POWER REACTOR

Science.gov (United States)

King, L.D.P.

1959-09-01

A homogeneous nuclear power reactor utilizing forced circulation of the liquid fuel is described. The reactor does not require fuel handling outside of the reactor vessel during any normal operation including complete shutdown to room temperature, the reactor being selfregulating under extreme operating conditions and controlled by the thermal expansion of the liquid fuel. The liquid fuel utilized is a uranium, phosphoric acid, and water solution which requires no gus exhaust system or independent gas recombining system, thereby eliminating the handling of radioiytic gas.

Gas-cooled nuclear reactor

International Nuclear Information System (INIS)

1974-01-01

The invention aims at simplying gas-cooled nuclear reactors. For the cooling gas, the reactor is provided with a main circulation system comprising one or several energy conversion main groups such as gas turbines, and an auxiliary circulation system comprising at least one steam-generating boiler heated by the gas after its passage through the reactor core and adapted to feed a steam turbine with motive steam. The invention can be applied to reactors the main groups of which are direct-cycle gas turbines [fr

Nuclear reactor cavity streaming shield

International Nuclear Information System (INIS)

Klotz, R.J.; Stephen, D.W.

1978-01-01

The upper portion of a nuclear reactor vessel supported in a concrete reactor cavity has a structure mounted below the top of the vessel between the outer vessel wall and the reactor cavity wall which contains hydrogenous material which will attenuate radiation streaming upward between vessel and the reactor cavity wall while preventing pressure buildup during a loss of coolant accident

Radioactive nuclides in nuclear reactors

International Nuclear Information System (INIS)

Akatsu, Eiko

1982-12-01

In the Nuclear Engineering School of JAERI, many courses are presented for the people working in and around nuclear reactors. The curricula of the courses contain also chemical subject materials. With reference to the foreign curricula, a plan of educational subject material of chemistry was considered for students of the school in the previous report (JAERI-M 9827), where the first part of the plan, ''Fundamentals of Reactor Chemistry'', was reviewed. This report is a review of the second part of the plan containing fission products chemistry, actinoids elements chemistry and activated reactor materials chemistry. (author)

Digital computer control of a research nuclear reactor

International Nuclear Information System (INIS)

Crawford, Kevan

1986-01-01

Currently, the use of digital computers in energy producing systems has been limited to data acquisition functions. These computers have greatly reduced human involvement in the moment to moment decision process and the crisis decision process, thereby improving the safety of the dynamic energy producing systems. However, in addition to data acquisition, control of energy producing systems also includes data comparison, decision making, and control actions. The majority of the later functions are accomplished through the use of analog computers in a distributed configuration. The lack of cooperation and hence, inefficiency in distributed control, and the extent of human interaction in critical phases of control have provided the incentive to improve the later three functions of energy systems control. Properly applied, centralized control by digital computers can increase efficiency by making the system react as a single unit and by implementing efficient power changes to match demand. Additionally, safety will be improved by further limiting human involvement to action only in the case of a failure of the centralized control system. This paper presents a hardware and software design for the centralized control of a research nuclear reactor by a digital computer. Current nuclear reactor control philosophies which include redundancy, inherent safety in failure, and conservative yet operational scram initiation were used as the bases of the design. The control philosophies were applied to the power monitoring system, the fuel temperature monitoring system, the area radiation monitoring system, and the overall system interaction. Unlike the single function analog computers that are currently used to control research and commercial reactors, this system will be driven by a multifunction digital computer. Specifically, the system will perform control rod movements to conform with operator requests, automatically log the required physical parameters during reactor

Nuclear reactor

International Nuclear Information System (INIS)

Miyashita, Akio.

1981-01-01

Purpose: To facilitate and accelerate a leakage test of valves of a main steam pipe by adding a leakage test partition valve thereto. Constitution: A leakage testing partition valve is provided between a pressure vessel for a nuclear reactor and the most upstream side valve of a plurality of valves to be tested for leakage, a testing branch pipe is communicated with the downstream side of the partition valve, and the testing water for preventing leakage is introduced thereto through the branch pipe. Since main steam pipe can be simply isolated by closing the partition valve in the leakage test, the leakage test can be conducted without raising or lowering the water level in the pressure vessel, and since interference with other work in the reactor can be eliminated, the leakage test can be readily conducted parallel with other work in the reactor in a short time. Clean water can be used without using reactor water as the test water. (Yoshihara, H.)

Nuclear reactors

International Nuclear Information System (INIS)

Prescott, R.F.

1976-01-01

A nuclear reactor containment vessel faced internally with a metal liner is provided with thermal insulation for the liner, comprising one or more layers of compressible material such as ceramic fiber, such as would be conventional in an advanced gas-cooled reactor and also a superposed layer of ceramic bricks or tiles in combination with retention means therefor, the retention means (comprising studs projecting from the liner, and bolts or nuts in threaded engagement with the studs) being themselves insulated from the vessel interior so that the coolant temperatures achieved in a High-Temperature Reactor or a Fast Reactor can be tolerated with the vessel. The layer(s) of compressible material is held under a degree of compression either by the ceramic bricks or tiles themselves or by cover plates held on the studs, in which case the bricks or tiles are preferably bedded on a yielding layer (for example of carbon fibers) rather than directly on the cover plates

Calculation models for a nuclear reactor

International Nuclear Information System (INIS)

Tashanii, Ahmed Ali

2010-01-01

Determination of different parameters of nuclear reactors requires neutron transport calculations. Due to complicity of geometry and material composition of the reactor core, neutron calculations were performed for simplified models of the real arrangement. In frame of the present work two models were used for calculations. First, an elementary cell model was used to prepare cross section data set for a homogenized-core reactor model. The homogenized-core reactor model was then used to perform neutron transport calculation. The nuclear reactor is a tank-shaped thermal reactor. The semi-cylindrical core arrangement consists of aluminum made fuel bundles immersed in water which acts as a moderator as well as a coolant. Each fuel bundle consists of aluminum cladded fuel rods arranged in square lattices. (author)

Assistance in chemistry and chemical processes related to primary, secondary and ancillary systems of nuclear power plants

International Nuclear Information System (INIS)

Chocron, Mauricio A.; Becquart, Elena T.; Iglesias, Alberto M.; La Gamma, Ana M.; Villegas, Marina

2003-01-01

Argentina is currently running two nuclear power plants: Atucha I (CNA I) and Embalse (CNE) operated by Nucleoelectrica Argentina (NASA) whereas the National Atomic Energy Commission (CNEA), among other activities, is responsible for research and development in the nuclear field, operates research reactors and carries out projects related to them. In particular, the Reactor Chemistry Section personnel (currently part of the Chemistry Dept.) has been working on the field of reactor water chemistry for more than 25 years, on research and support to the NPPs chemistry department. Though the most relevant tasks have been connected to primary and secondary circuits chemistry, ancillary systems show along the time unexpected problems or feasible improvements originated in the undergoing operating time as well as in phenomena not foreseen by the constructors. In the present paper are presented the tasks performed in relation to the following systems of Embalse NPP: 1) Heavy water upgrade column preliminary water treatment; 2) Liquid waste system preliminary water treatment; and 3) Primary heat transport system coolant crud composition. (author)

Nuclear data for nuclear reactor analyses

International Nuclear Information System (INIS)

Pearlstein, S.

1984-01-01

A discussion of nuclear data is presented emphasizing to what extent data are known and to what accuracy. The principal data of interest is that for neutron cross-sections. The changing status of data, evaluated nuclear data files and data validation and improvement are described. Although the discussion relates to nuclear data for reactor analysis may of the results also apply to fusion, accelerator, shielding, biomedical, space and defense studies. (U.K.)

Nuclear reactors

International Nuclear Information System (INIS)

Matheson, J.E.

1983-01-01

A nuclear reactor has an upper and a lower grid plate. Protrusions project from the upper grid plate. Fuel assemblies having end fittings fit between the grid plates. An arrangement is provided for accepting axial forces generated during the operation of the nuclear reactor by the flow of the cooling medium and thermal expansion and irradiation-induced growth of the fuel assembly, which comprises rods. Each fuel assembly rests on the lower grid plate and its upper end is elastically supported against the upper grid plate by the above-mentioned arrangement. The arrangement comprises four (for example) torsion springs each having a torsion tube and a torsion bar nested within the torsion tube and connected at one end thereto. The other end of the torsion bar is connected to an associated one of four lever arms. The torsion tube is rigidly connected to the other end fitting and the springs are disposed such that the lever arms are biassed against the protrusions. (author)

Exporting apocalypse: CANDU reactors and nuclear proliferation

International Nuclear Information System (INIS)

McKay, Paul.

The author believes that the peaceful use of nuclear technology leads inevitably to the production of nuclear weapons, and that CANDU reactors are being bought by countries that are likely to build bombs. He states that exports of reactors and nuclear materials cannot be defended and must be stopped

Nuclear power reactor technology

International Nuclear Information System (INIS)

1978-09-01

Risoe National Laboratory was established more than twenty years ago with research and development of nuclear reactor technology as its main objective. The Laboratory has by now accumulated many years of experience in a number of areas vital to nuclear reactor technology. The work and experience of, and services offered by the Laboratory within the following fields are described: Health physics site supervision; Treatment of low and medium level radioactive waste; Core performance evaluation; Transient analysis; Accident analysis; Fuel management; Fuel element design, fabrication and performance evaluation; Non-destructive testing of nuclear fuel; Theoretical and experimental structural analysis; Reliability analysis; Site evaluation. Environmental risk and hazard calculation; Review and analysis of safety documentation. Risoe has already given much assistance to the authorities, utilities and industries in such fields, carrying out work on both light and heavy water reactors. The Laboratory now offers its services to others as a consultant, in education and training of staff, in planning, in qualitative and quantitative analysis, and for the development and specification of fabrication techniques. (author)

Dynamics of nuclear reactor operational cycles

International Nuclear Information System (INIS)

Chapman, L.D.; Wayland, J.R.

With this system dynamics computer model, one can explore the long term effects of a nuclear reactor program. Given an input demand for reactors, the consequences on each sector and the interactions among sectors can be simulated to provide a better understanding of the time development of a nuclear reactor program. The model permits the determination of various levels of activity as a function of time for plant enrichment, fuel fabrication, fuel reprocessing and storage of waste products. In addition, the rates of construction of reactors, spent fuel transit, disposal of waste, mining, shipping, recycling and enrichment can be investigated for optimal planning purposes. The model has been written in a very general manner so that it can be used to simulate any nuclear reactor program. It is an easy task to relate the amount of accidental or operational release of radioactive contaminants into our environment to the activity levels of each of the above sectors. (U.S.)

Nuclear reactor safety systems

International Nuclear Information System (INIS)

Ball, R.M.; Roberts, R.C.

1980-01-01

A safety system for shutting down a nuclear reactor under overload conditions is described. The system includes a series of parallel-connected computer memory type look-up tables each of which receives data on a particular reactor parameter and in each of which a precalculated functional value for that parameter is stored indicative of the percentage of maximum reactor load that the parameter contributes. The various functional values corresponding to the actual measured parameters are added together to provide a control signal used to shut down the reactor under overload conditions. (U.K.)

How power is generated in a nuclear reactor

International Nuclear Information System (INIS)

Swaminathan, V.

1978-01-01

Power generation by nuclear fission as a result of chain reaction caused by neutrons interacting with fissile material such as 235 U, 233 U and 239 Pu is explained. Electric power production by reactor is schematically illustrated. Materials used in thermal reactor and breeder reactor are compared. Fuel reprocessing and disposal of radioactive waste coming from reprocessing plant is briefly described. Nuclear activities in India are reviewed. Four heavy water plants and two power reactors are under construction and will be operative in the near future. Two power reactors are already in operation. Nuclear Fuel Complex at Hyderabad supplies fuel element to the reactors. Fuel reprocessing and waste management facility has been set up at Tarapur. Bhabha Atomic Research Centre at Bombay and Reactor Research Centre at Kalpakkam near Madras are engaged in applied and basic research in nuclear science and engineering. (B.G.W.)

Preparation fo nuclear research reactors operators

International Nuclear Information System (INIS)

Roedel, G.

1986-01-01

The experience obtained with the training of operators of nuclear research reactors is presented. The main tool used in the experiments is the IPR-R1 reactor, a TRIGA MARK I type, owned by Nuclear Technology Development Centre (CDTN) of NUCLEBRAS. The structures of the Research Reactors Operators Training Course and of the Radiological Protection Course, as well as the Operators Qualifying and Requalifying Program, all of them prepared at CDTN are also presented. Mention is made of the application of similar experiments to other groups, such as students coming from Nuclear Sciences and Techniques Course of the Federal University of Minas Gerais. (Author) [pt

Preparation of nuclear research reactors operators

International Nuclear Information System (INIS)

Roedel, G.

1986-01-01

The experience obtained with the training of operators of nuclear research reactors is presented. The main tool used in the experiments is the IPR-R1 reactor, a TRIGA MARK I type, owned by Nuclear Technology Development Centre (CDTN) of NUCLEBRAS. The structures of the Research Reactors Operators Training Course and of the Radiological Protection Course, as well as the Operators Qualifying and Requalifying Program, all of them prepared at CDTN, are also presented. Mention is made of the application of similar experiments to other groups, such as students coming from Nuclear Sciences and Techniques Course of the Federal University of Minas Gerais. (Author) [pt

Proposal of space reactor for nuclear electric propulsion system

International Nuclear Information System (INIS)

Nishiyama, Takaaki; Nagata, Hidetaka; Nakashima, Hideki

2009-01-01

A nuclear reactor installed in spacecrafts is considered here. The nuclear reactor could stably provide an enough amount of electric power in deep space missions. Most of the nuclear reactors that have been developed up to now in the United States and the former Soviet Union have used uranium with 90% enrichment of 235 U as a fuel. On the other hand, in Japan, because the uranium that can be used is enriched to below 20%, the miniaturization of the reactor core is difficult. A Light-water nuclear reactor is an exception that could make the reactor core small. Then, the reactor core composition and characteristic are evaluated for the cases with the enrichment of the uranium fuel as 20%. We take up here Graphite reactor, Light-water reactor, and Sodium-cooled one. (author)

Energy from nuclear reactors

International Nuclear Information System (INIS)

Hospe, J.

1977-01-01

This VDI-Nachrichten series has the target to provide a technical-objective basis for the discussion of the pros and cons of nuclear power. The first part deals with LWR-type reactors which so far have prevailed in nuclear power generation. (orig.) [de

Nuclear Power Reactors in the World. 2014 Ed

International Nuclear Information System (INIS)

2014-01-01

Nuclear Power Reactors in the World is an annual publication that presents the most recent data pertaining to nuclear power reactors in IAEA Member States. This thirty-fourth edition of Reference Data Series No. 2 provides a detailed comparison of various statistics up to and including 31 December 2013. The tables and figures contain the following information: — General statistics on nuclear reactors in IAEA Member States; — Technical data on specific reactors that are either planned, under construction or operational, or that have been shut down or decommissioned; — Performance data on reactors operating in IAEA Member States, as reported to the IAEA. The data compiled in this publication is a product of the IAEA’s Power Reactor Information System (PRIS). The PRIS database is a comprehensive source of data on all nuclear power reactors in the world. It includes specification and performance history data on operational reactors as well as on reactors under construction or in the decommissioning process. The IAEA collects this data through designated national correspondents in Member States

Design of an organic simplified nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Shirvan, Koroush [Dept. of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge (United States); Forrest, Eric [Primary Standards Laboratory, Sandia National Laboratories, Albuquerque (United States)

2016-08-15

Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

Design of an Organic Simplified Nuclear Reactor

Directory of Open Access Journals (Sweden)

Koroush Shirvan

2016-08-01

Full Text Available Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.

ENS RRFM 2005: 9th international topical meeting on research reactor fuel management. Transactions

International Nuclear Information System (INIS)

2005-01-01

The ENS topical meeting on research reactor fuel management is an annual conference launched successfully in 1997. It has since then grown into well established international forum for the exchange and expertise on all significant aspects of the nuclear fuel cycle of research reactors. Oral presentations at this meeting were divided in the following four sessions: International Topics; Fuel Development, Qualification, Fabrication and Licensing; Reactor Operation, Fuel Safety and Core Conversion; Spent Fuel Management, Back-end Options, Transportation. The three poster sessions were devoted to fuel development, qualification, fabrication and licensing; reactor operation, fuel safety, core conversion, spent fuel; spent fuel management, fuel cycle back-end options, transportation

Seals in nuclear reactors

International Nuclear Information System (INIS)

1979-01-01

The aim of this invention is the provision of improved seals for reactor vessels in which fuel assemblies are located together with inlets and outlets for the circulation of a coolant. The object is to provide a seal arrangement for the rotatable plugs of nuclear reactor closure heads which has good sealing capacities over a wide gap during operation of the reactor but which also permits uninhibited rotation of the plugs for maintenance. (U.K.)

Nuclear reactors for the future

International Nuclear Information System (INIS)

Vijayan, P.K.; Kamble, M.T.; Dulera, I.V.

2013-01-01

For the sustainable development of nuclear power plants with enhanced safety features, economic competitiveness, proliferation resistance and physical protection, several advanced reactor developments have been initiated world-wide. The major advanced reactor initiatives and the proposed advanced reactor concepts have been briefly reviewed along with their advantages and challenges. Various advanced reactor designs being pursued in India have also been briefly described in the paper. (author)

78 FR 64028 - Decommissioning of Nuclear Power Reactors

Science.gov (United States)

2013-10-25

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0035] Decommissioning of Nuclear Power Reactors AGENCY... the NRC's regulations relating to the decommissioning process for nuclear power reactors. The revision... Commission (NRC) is issuing Revision 1 of regulatory guide (RG) 1.184 ``Decommissioning of Nuclear Power...

Vessel supporting structure for liquid metal cooled nuclear reactors

International Nuclear Information System (INIS)

Mahe, Armel; Jullien, Georges

1974-01-01

The supporting structure described is for a liquid metal cooled nuclear reactor, the vessel being of the type suspended to the end slab of the reactor. It includes a ring connected at one of its two ends to a single shell and at the other end to two shells. One of these three shells connected to the lower end of the ring forms the upper part of the vessel to be supported. The two other shells are embedded in two sperate parts of the slab. The ring and shell assembly is housed in an annular space provided in the end slab and separating it into two parts, namely a central part and a peripheral part [fr

Thermal-hydraulic analysis of nuclear reactors

CERN Document Server

Zohuri, Bahman

2015-01-01

This text covers the fundamentals of thermodynamics required to understand electrical power generation systems and the application of these principles to nuclear reactor power plant systems. It is not a traditional general thermodynamics text, per se, but a practical thermodynamics volume intended to explain the fundamentals and apply them to the challenges facing actual nuclear power plants systems, where thermal hydraulics comes to play.  Written in a lucid, straight-forward style while retaining scientific rigor, the content is accessible to upper division undergraduate students and aimed at practicing engineers in nuclear power facilities and engineering scientists and technicians in industry, academic research groups, and national laboratories. The book is also a valuable resource for students and faculty in various engineering programs concerned with nuclear reactors. This book also: Provides extensive coverage of thermal hydraulics with thermodynamics in nuclear reactors, beginning with fundamental ...

Nuclear reactor assembly

International Nuclear Information System (INIS)

Dorner, H.; Scholz, M.; Jungmann, A.

1975-01-01

A nuclear reactor assembly includes a reactor pressure tank having a substantially cylindrical side wall surrounded by the wall of a cylindrical cavity formed by a biological shield. A rotative cylindrical wall is interposed between the walls and has means for rotating it from outside of the shield, and a probe is carried by the rotative wall for monitoring the pressure tank's wall. The probe is vertically movable relative to the rotative cylindrical wall, so that by the probe's vertical movement and rotation of the rotative cylinder, the reactor's wall can be very extensively monitored. If the reactor pressure tank's wall fails, it is contained by the rotative wall which is backed-up by the shield cavity wall. (Official Gazette)

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

Nuclear reactor container

International Nuclear Information System (INIS)

Hosaka, Seiichi.

1988-01-01

Cables coverd with non-halogen covering material are used as electric wire cables wired for supplying electric power to a reactor recycling pump. Silicone rubber having specified molecular formula is used for the non-halogen covering material. As a result, formation of chlorine in a nuclear reactor container can be eliminated and increase in the deposited salts to SUS pipeways, etc. can be prevented, to avoid the occurrence of stress corrosion cracks. (H.T.)

Order for execution of the law concerning regulations of nuclear source materials, nuclear fuel materials and reactors

International Nuclear Information System (INIS)

1987-01-01

Chapeter 1 specifies regulations concerning business management for refining and processing, which cover application for designation of refining operation, application for permission for processing operation, and approval of personnel responsible for handling nuclear fuel. Chapter 2 specifies regulations concerning construction and operation of nuclear reactors, which cover application for construction of nuclear reactors, reactors in a research and development stage, application for permission concerning nuclear reactors mounted on foreign nuclear powered ships, application for permission for alteration concerning construction of nuclear reactors, application for permission for alteration concerning nuclear reactors mounted on foreign nuclear powered ships, nuclear reactor facilities to be subjected to regular inspection, nuclear reactor for which submission of operation plan is not required, and application for permission for transfer of nuclear reactor. Chapter 2 also specifies regulations concerning business management for reprocessing and waste disposal. Chapter 3 stipulates regulations concerning use of nuclear fuel substances, nuclear material substances and other substances covered by international regulations, which include rules for application for permission for use of nuclear fuel substances, etc. Supplementary provisions are provided in Chapter 4. (Nogami, K.)

Nuclear Power Reactor simulator - based training program

International Nuclear Information System (INIS)

Abdelwahab, S.A.S.

2009-01-01

nuclear power stations will continue playing a major role as an energy source for electric generation and heat production in the world. in this paper, a nuclear power reactor simulator- based training program will be presented . this program is designed to aid in training of the reactor operators about the principles of operation of the plant. also it could help the researchers and the designers to analyze and to estimate the performance of the nuclear reactors and facilitate further studies for selection of the proper controller and its optimization process as it is difficult and time consuming to do all experiments in the real nuclear environment.this program is written in MATLAB code as MATLAB software provides sophisticated tools comparable to those in other software such as visual basic for the creation of graphical user interface (GUI). moreover MATLAB is available for all major operating systems. the used SIMULINK reactor model for the nuclear reactor can be used to model different types by adopting appropriate parameters. the model of each component of the reactor is based on physical laws rather than the use of look up tables or curve fitting.this simulation based training program will improve acquisition and retention knowledge also trainee will learn faster and will have better attitude

Nuclear reactors built, being built, or planned: 1987

International Nuclear Information System (INIS)

1988-06-01

Nuclear Reactors Built, Being Built, or Planned contains unclassified information about facilities built, being built, or planned in the United States for domestic use or export as of December 31, 1987. The Office of Scientific and Technical Information, US Department of Energy, gathers this information annually for Washington headquarters and field offices of DOE; from the US Nuclear regulatory Commission; from the US reactor manufacturers who are the principal nuclear contractors for foreign reactor locations; from US and foreign embassies; and from foreign governmental nuclear departments. The major change in this revision involves the data related to shutdown and dismantled facilities. Because this information serves substantially different purposes, it has been accumulated in a separate section, ''Reactors and Facilities Shutdown or Dismantled.'' Cancelled reactors or reactors whose progress has been terminated at some stage before operation are included in this section