WorldWideScience

Sample records for central nuclear en atucha reactor

  1. 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)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

  4. Nuclear knowledge preservation of Atucha type Reactor: Practical approaches and lessons learned

    International Nuclear Information System (INIS)

    Full text: The nuclear option is born in Argentina in 1950. Since then and after half century, the Argentine nuclear sector reaches a hierarchy and an important technological - scientific dimension in all the issues related to activities linked to the nuclear sector. The export of research reactors, the nuclear fuel production and the installation of nucleoelectrical power stations constitute one of the most outstanding achievements of this activity. The Atucha I Nuclear Power Plant, of Siemens-KWU technology with 30 years of operation and Embalse (Candu type) with 20 years of operation, register an efficient and safe performance, within the worldwide standards. In addition, and with Siemens technology the Atucha II NPP is in construction with a 82% of work advance. At the moment Siemens, has transferred his place as designer of reactors LWR to the Framatom ANP Company and consequently Argentina will have to make all the efforts to assure the 'knowledge preservation' of the technology of this line of reactors, if it hopes that Atucha I extends its useful life and Atucha II is finalized. Likewise the aging of the personnel in the sector is worrisome being numerous the great number of people who are close to retire what it worsens if we think that there are a few young people able in this specialty by the lack of the entrance of candidates in compatible races This situation generated the necessity to implement a system of Knowledge Management to capture and to capitalize the tacit and explicit knowledge, to spread it and to share by using techniques and tools adapted through the organization. Practical Approaches and Lessons Learned: A - Analysis of the strategy: The strategy was based on recognizing the critical knowledge by means of the use of a methodology that incorporates the technique of the map of the critical knowledge. That is to say, to recognize the patrimony of the knowledge around the reactor. The layout of the knowledge map of reactor together with the

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  7. Argentina: Nuclear power development and Atucha 2

    International Nuclear Information System (INIS)

    In 2014, nuclear energy generated about 5,257 GWh of electricity or a total share of 4.05 % of the total electrical energy of about 129,747.63 GWh kWh produced in Argentina and there has been a trend for this production to increase. Argentina currently has a nuclear production capacity of 1,010 megawatts of electrical energy. However, when the Atucha 2 nuclear power plant is completed and starts commercial operation, it will add 745 megawatts to this electrical production capacity. There are two sites with nuclear power plants in Argentina: Atucha and Embalse. The Embalse nuclear power plant went into operation in 1984. At the Atucha site, the Atucha-1 nuclear power plant started operation in 1974. It was the first nuclear power plant in Latin America. Construction of Atucha-2 started in 1981 but advanced slowly due to funding and was suspended in 1994 when the plant was 81 % built. In 2003, new plans were approved to complete the Atucha 2. I summer 2014 the plant went critical for the first time. The construction was completed under a contract with AECL.

  8. Argentina: Nuclear power development and Atucha 2

    Energy Technology Data Exchange (ETDEWEB)

    Nogarin, Mauro

    2015-08-15

    In 2014, nuclear energy generated about 5,257 GWh of electricity or a total share of 4.05 % of the total electrical energy of about 129,747.63 GWh kWh produced in Argentina and there has been a trend for this production to increase. Argentina currently has a nuclear production capacity of 1,010 megawatts of electrical energy. However, when the Atucha 2 nuclear power plant is completed and starts commercial operation, it will add 745 megawatts to this electrical production capacity. There are two sites with nuclear power plants in Argentina: Atucha and Embalse. The Embalse nuclear power plant went into operation in 1984. At the Atucha site, the Atucha-1 nuclear power plant started operation in 1974. It was the first nuclear power plant in Latin America. Construction of Atucha-2 started in 1981 but advanced slowly due to funding and was suspended in 1994 when the plant was 81 % built. In 2003, new plans were approved to complete the Atucha 2. I summer 2014 the plant went critical for the first time. The construction was completed under a contract with AECL.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  11. Use of remote operated equipment in the inspection and repair of the reactor at Atucha I Nuclear Power Plant

    International Nuclear Information System (INIS)

    On August 11, 1988, failures in a cooling channel of Atucha I Nuclear Power Plant in Argentina were detected. This paper describes the use of remote operated equipment for the subsequent repair. The work on the repair included the following phases: inspection of core intervals with an underwater TV camera, definition of repair methods, design and construction of the necessary tools, repair and finally inspection and evaluation of repair results. (author). 19 figs

  12. Atucha I nuclear power plant surveillance programme

    International Nuclear Information System (INIS)

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

  13. Atucha I nuclear power plant transients analysis

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    Diego Hurtado de Mendoza

    2005-01-01

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

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

    International Nuclear Information System (INIS)

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

  17. Atucha I nuclear power plant achieving a full safeguards approach

    International Nuclear Information System (INIS)

    At Atucha I nuclear power plant (NPP) it has recently been fully attained technical safeguards inspection goals. This is an important accomplishment since the facility had partially attained the safeguards technical goals for some years. Moreover, this situation implied in one occasion that the criterion was not met for the entire State. Nevertheless, it should be stressed that the IAEA, ABACC and ARN have always been able to draw positive safeguards conclusions. According to IAEA Safeguards Criteria issued in 1991 and in particular since 1995, the facility has attained, only partially, one of the technical inspection goals due to the requirement of having a fuel flow monitor to count and verify by gross defect the fuels discharged from the reactor core before being placed under IAEA and ABACC surveillance at the storage ponds. Argentina repeatedly stated in different IAEA instances, including its policy-making organs, its opinion that criteria mainly based on one type of Ol, could not be easily extrapolated to other Ols, in particular those already constructed and in operation. Our country nevertheless engaged fully, since the issuing of such criteria, in finding an integral solution to this technical issue. From ARN viewpoint, the goal has been to find a suitable method to fulfil this requirement without implying substantive construction modifications to the facility, together with having in place appropriate Nda techniques to verify the spent fuel inventory at the ponds of Atucha I before the Vim were left working. Besides, this Nda method is appropriate in case of a C and S failure. The paper describes the various attempts for solving this issue that would imply a convenient design, a testing period and the final implementation for safeguards purposes of the selected technical solutions. It is also shown a comparison between the inspection activities performed before the installation of the Vim (V Irradiated Fuel Monitor) and the ones expected with this system

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

    International Nuclear Information System (INIS)

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

  19. Vibration monitoring of the Atucha-I NPP reactor vessel: experience and future improvements

    International Nuclear Information System (INIS)

    Keeping nuclear plant components in prime operating condition is essential for overall plant safety and availability. One way to assess the good operating condition of nuclear components is an on-line monitoring system. A vibration monitoring system is intended to monitor a component or system to ensure that changes in the mechanical integrity are detected early, even during operation. This paper describes the experience gained during the development of a vibration monitoring system for the Atucha-I Nuclear Power Plant reactor vessel that started up in 1990. High sensitivity velocity sensors were mounted on the closure head of the reactor vessel to monitor the vibrations. Some of the results obtained for different reactor operating conditions are presented in the form of frequency spectra or RMS values for different frequency bands. A description of the system, its evolution and future improvements are also included. (orig.)

  20. Station Black-Out Analysis with MELCOR 1.8.6 Code for Atucha 2 Nuclear Power Plant

    OpenAIRE

    Analia Bonelli; Oscar Mazzantini; Martin Sonnenkalb; Marcelo Caputo; Juan Matias García; Pablo Zanocco; Marcelo Gimenez

    2012-01-01

    A description of the results for a Station Black-Out analysis for Atucha 2 Nuclear Power Plant is presented here. Calculations were performed with MELCOR 1.8.6 YV3165 Code. Atucha 2 is a pressurized heavy water reactor, cooled and moderated with heavy water, by two separate systems, presently under final construction in Argentina. The initiating event is loss of power, accompanied by the failure of four out of four diesel generators. All remaining plant safety systems are supposed to be avail...

  1. Hot functional testing of the pressurized heavy water reactor plant Atucha II with light water

    International Nuclear Information System (INIS)

    The two pressurized heavy water reactors PHWR Atucha I (designed and built by S/KWU, now AREVA), and Atucha II (designed by S/KWU and plant construction now completed by NA-SA) are owned by Nucleoelectrica Argentina S.A. (NA-SA). Atucha II was designed in the 1980'ies in parallel to the two most recent S/KWU PWR generations Prekonvoi and Konvoi. Its basic design has been updated and optimized including also backfitting of components and systems for severe accident management. The gross electric power of the plant is 745 MWe. Construction and commissioning of Atucha II has been resumed by NA-SA after a work stop in the 1990'ies and is now almost completed. Hot functional testing HFT was performed in two phases in September and October 2013 and in March and April 2014. Hot functional testing was performed with light water and the fuel assemblies loaded. The chemistry program for the HFT was derived from practices and experience gathered at other S/KWU designed PWRs during HFTs and consisted of the following main targets and requirements: (1) Low chloride and sulfate concentrations close to normal operation values specified in the VGB water chemistry guideline for power operation of PWR plants; (2) Thorough oxygen removal during heat-up and reducing conditions through N2H4 dosing; (3) High pH value (target range 1.5 to 2 ppm Li); (4) Passivation treatment of the nuclear steam supply system NSSS at temperatures of at least 260°C for a time period of at least 120 hours; (5) Zinc addition at a constant rate of 20 g Zn per day throughout the various HFT phases. Zinc dosing was begun during the first heat-up of the plant at temperatures above approx. 150°C. Daily measurement of the zinc concentration for process control was not necessary and not required due to the elaborated zinc application procedure. The main results of the chemistry program for the HFT of plant are described and evaluated in this contribution. Data shows that all chemistry targets were met

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

    International Nuclear Information System (INIS)

    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)

  3. Commissioning of the water demineralization plant of Atucha II Nuclear Power Plant

    International Nuclear Information System (INIS)

    In Argentina there are two operating Nuclear Power Reactors and a third one is being constructed. Embalse NPP is a 648 Mwe CANDU®-600 type pressurized heavy water reactor (PHWR), designed and built by Atomic Energy of Canada (AECL) and in commercial operation since 1984. Atucha I is a Pressurized Vessel Heavy Water Reactor (PVHWR) of 340 Mwe, in operation since 1974, and Atucha II (also PHWR) of 740 Mwe is in advanced construction state, both of them designed by SIEMENS-KWU. All of these Nuclear Power Plants are operated by Nucleoelectrica Argentina (N.A.S.A.). The Comision Nacional de Energia Atomica (C.N.E.A.) is the R and D nuclear institution in the country that, among many other topics, provides technical support to the plants. Although the Atucha II project has suffered some years of delay, pressure tests are expected to be carried out by the end of the year 2010 and in that sense, water chemistry related activities, specifications, chemistry manuals, laboratories organization and personnel training are acquiring importance. The demineralized water needed for the secondary and auxiliary systems is obtained by means of a demineralization plant, which purifies water from Parana River up to nuclear grade. This plant was designed by Degremont in 1979 and consists of a preliminary treatment by coagulation - flocculation and gravel filters, and subsequent demineralization with ion exchange resins. For the commissioning of the demineralization plant, preliminary tests in the chemical laboratory are performed. The flocculator is simulated using a Jar-Test, different coagulants and coagulation aids are tested with the objective of selecting the best product and defining its optimum dosage. The coagulated water is filtered by means of a funnel with filtration paper and sand. The clarified water thus obtained is treated by ion exchange resins, the train consisting of a cationic, an anionic and a mixed bed. The purpose of the laboratory experiments is to test the resins

  4. Integrated management for aging of Atucha Nuclear Power Plant

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

  6. Slightly enriched uranium in Atucha I reactor: an important technological development of CNEA and the nucleo electrical sector

    International Nuclear Information System (INIS)

    The fuel cycle change in Atucha I nuclear power plant (CNA 1) represents a successful and relevant development in the field of nuclear technology, going from the basic studies (CNEA) to the industrial use (CNEA-NASA-CONUAR), with remarkable economic results because it caused an important decrease in generating costs, with potential implications in Embalse nuclear power plant (CNE) and Atucha II nuclear power plant (CNA 2). The new fuel cycle uses low enriched uranium (LEU), with 0.85% in 235U isotope instead of 0.71% as was in the original natural uranium fuel cycle. This produces an increase in burn up of more than 80% and correspondingly a decrease of 45% in fuel use and waste volume for the same energy produced. The economic saving amounts to approximately $ 6,5000,000 per year in fuel. Even if there are records of use of low enriched uranium in reactors moderated with heavy water, CNA 1 is the first power plant in the world that uses it commercially, with important economic and operating advantages. (author)

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

    International Nuclear Information System (INIS)

    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

  8. Environmental management in Atucha-I and Embalse nuclear power plants

    International Nuclear Information System (INIS)

    The operator, Nucleoelectrica Argentina, received the certification of the IRAM (Argentine Institute of Normalization) under the ISO standard 14001:1996 for the System of Environmental Management implemented in the nuclear power plants of Atucha I (Lima, province of Buenos Aires, Argentina) and Embalse (Cordoba, Argentina). The process of certification is outlined as well as its effects on the environmental management

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

    International Nuclear Information System (INIS)

    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)

  10. Sub-arc narrow gap welding of the Atucha 2 reactor pressure vessel closure head

    Energy Technology Data Exchange (ETDEWEB)

    Hantsch, H.; Million, K.; Zimmermann, H.; Cerjak, H.; Pellkofer, D.; Schmidt, J.

    1983-01-01

    Narrow-gap technology was used for reasons of design and fabrication when dome of the Atucha 2 reactor pressure vessel was welded to its flange. Preliminary tests has yielded the necessary improvements to the well-established sub-arc practice. New facilities had to be developed for the welding and the accompanying machining work (finishing in the narrow gap). Special measures were adopted for monitoring the welding process and for recording the welding parameters. The new method was tried out on several large test coupons before the welding of the final product was started. No difficulties were encountered during the work.

  11. Sub-arc narrow gap welding of the Atucha 2 reactor pressure vessel closure head

    International Nuclear Information System (INIS)

    Narrow-gap technology was used for reasons of design and fabrication when dome of the Atucha 2 reactor pressure vessel was welded to its flange. Preliminary tests has yielded the necessary improvements to the well-established sub-arc practice. New facilities had to be developed for the welding and the accompanying machining work (finishing in the narrow gap). Special measures were adopted for monitoring the welding process and for recording the welding parameters. The new method was tried out on several large test coupons before the welding of the final product was started. No difficulties were encountered during the work. (orig.)

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

    International Nuclear Information System (INIS)

    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).

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

    International Nuclear Information System (INIS)

    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

  14. Micrometeorological study of the Atucha Nuclear Power Plant site

    International Nuclear Information System (INIS)

    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.)

  15. Station Black-Out Analysis with MELCOR 1.8.6 Code for Atucha 2 Nuclear Power Plant

    International Nuclear Information System (INIS)

    A description of the results for a Station Black-Out analysis for Atucha 2 Nuclear Power Plant is presented here. Calculations were performed with MELCOR 1.8.6 YV3165 Code. Atucha 2 is a pressurized heavy water reactor, cooled and moderated with heavy water, by two separate systems, presently under final construction in Argentina. The initiating event is loss of power, accompanied by the failure of four out of four diesel generators. All remaining plant safety systems are supposed to be available. It is assumed that during the Station Black-Out sequence the first pressurizer safety valve fails stuck open after 3 cycles of water release, respectively, 17 cycles in total. During the transient, the water in the fuel channels evaporates first while the moderator tank is still partially full. The moderator tank inventory acts as a temporary heat sink for the decay heat, which is evacuated through conduction and radiation heat transfer, delaying core degradation. This feature, together with the large volume of the steel filler pieces in the lower plenum and a high primary system volume to thermal power ratio, derives in a very slow transient in which RPV failure time is four to five times larger than that of other German PWRs.

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

    Directory of Open Access Journals (Sweden)

    DOMINGO QUILICI

    2013-06-01

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

  17. New in-core instrumentation for Atucha I Nuclear Power Plant

    International Nuclear Information System (INIS)

    In-Core Neutron flux Monitoring Assemblies were developed and constructed for Atucha I Nuclear Power Plant (CNA I) in the Argentine National Atomic Energy Commission (CNEA) to replace the German original ones that had to be removed for maintenance reasons. Each assembly contains seven Self-Powered Neutron Detectors (SPND) in different measurement positions, protected with an external Zy-4 tube. Detector's mineral insulated wires pass the reactor pressure vessel (RPV) cover head through seven holes made in a removable steal-sealing plug. A braze welding is made between the plug and the wires to seal such holes. The manufacture of this component was divided into two big processes. One of them was the fabrication and testing of the detectors and the other was the construction and checking of the assembly. a) The SPND were totally developed and constructed in CNEA. A composite bar is assembled at the beginning of the process. Such bar is composed by an Inconel 600 external tube, an Inconel 600 central bar, a small tube of vanadium (neutron sensitive material) placed around the middle of the central bar and tubular OMg pellets, between the external tube and the central bar, as isolation. A sketch of the composite bar is shown. The composite bar is deformed by cold working with intermediate thermal treatments. The external diameter is reduced in several steps with the consequent increase in length. When the final diameter is obtained, the bar is cut in the middle of the vanadium (two detectors are obtained) and the ends are sealed. The fabrication method is patented and allows obtaining integral cable-detector devices. The electric contact between the vanadium and the central conductor is obtained by plastic deformation and the external sheath is continuous with a conic reduction from the detector up to the cable. In this way the internal welding, that is a cause of failure, and the welding between the sheaths are avoid. The main advantage of these detectors is to have

  18. 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)

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

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

    International Nuclear Information System (INIS)

    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 UO2 required. (Author)

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

    International Nuclear Information System (INIS)

    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.)

  1. Loss of main pumps in the Atucha 1 nuclear power station. Modeling with RELAP5/MOD3.2

    International Nuclear Information System (INIS)

    This paper analyzes the temporal evolution of natural circulation at the Atucha 1 nuclear power station when losing its main pumps due to lack of external feeding power. This leads to a temporal stoppage with important depressurization, from the nominal flow of forced circulation to another of natural one. Results are attained with a program called RELAP5/MOD3.2

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

    International Nuclear Information System (INIS)

    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)

  3. Centrales nucleares en España : el parón nuclear

    OpenAIRE

    Sanz Díaz, Benito (1949-)

    1984-01-01

    Este libro constituye una oportuna y valiosa contribución a ese necesario debate nacional sobre la energía nuclear. Tiene como objetivo la central nuclear de Cofrentes, una de las postreras y más singulares —tanto por su modelo como por su emplazamiento e historia de construcción— de las centrales nucleares que entrarán en funcionamiento en nuestro país. El libro, sin embargo, aporta información sobre la totalidad del programa nuclear español y su permanente visión económica garantiza la vía ...

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  7. Integrated ageing management of Atucha NPP

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  8. Integrated ageing management of Atucha NPP

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  10. 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)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  13. 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)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  16. 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)

    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

  17. Modeling with RELAP5/3.2. Thermal-hydraulic behaviour simulation because of the main pumps loss in the Atucha I nuclear power plant

    International Nuclear Information System (INIS)

    Time evolution of natural circulation in the Atucha I nuclear power plant (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 RELAP5/3.2 code's modeling. The study is based on the refrigeration condition 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 recorder values, in the same incident. The results are satisfactory in a way. On the other hand, it has suggested some modeling changes. The RELAP5/3.2 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. (authors)

  18. NUCLEAR REACTOR

    Science.gov (United States)

    Anderson, C.R.

    1962-07-24

    A fluidized bed nuclear reactor and a method of operating such a reactor are described. In the design means are provided for flowing a liquid moderator upwardly through the center of a bed of pellets of a nentron-fissionable material at such a rate as to obtain particulate fluidization while constraining the lower pontion of the bed into a conical shape. A smooth circulation of particles rising in the center and falling at the outside of the bed is thereby established. (AEC)

  19. 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)

    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)

  20. Widespread use of best estimate codes in Atucha II NPP simulator; Uso extendido de codigos Best estimate en el sumulador de Atucha II

    Energy Technology Data Exchange (ETDEWEB)

    Dieguez, G. L.; Garde, D.; Ruiz, J. A.; Exposito, A.

    2016-08-01

    Since the best estimate codes were first used in training simulators in the 90s, the computing power has increased dramatically and today the computing cost has become insignificant in these project. This allows using state of the art process simulation codes, both on detailed design models and also on the simulation of all the potentially biphasic plant systems. The Atucha II NPP simulator sets a good example. The quality of the results has enabled to proactively support the plant start-up, verifying the planned maneuvers and featuring reference behavior. It has also helped to understand unexpected phenomena and optimize the control loops. (Author)

  1. Ultrasonic meters in the feedwater flow to recover thermal power in the reactor of nuclear power plant of Laguna Verde U1 and U2; Medidores ultrasonicos en el flujo de agua de alimentacion para recuperar potencia termica en el reactor de la Central Nuclear Laguna Verde U1 and U2

    Energy Technology Data Exchange (ETDEWEB)

    Tijerina S, F. [CFE, Central Laguna Verde, Km. 42.5 Carretera Cardel-Nautla, Veracruz (Mexico)]. e-mail: francisco.tijerina@cfe.gob.mx

    2008-07-01

    The engineers in nuclear power plants BWRs and PWRs based on the development of the ultrasonic technology for the measurement of the mass, volumetric flow, density and temperature in fluids, have applied this technology in two primary targets approved by the NRC: the use for the recovery of thermal power in the reactor and/or to be able to realize an increase of thermal power licensed in a 2% (MUR) by 1OCFR50 Appendix K. The present article mentions the current problem in the measurement of the feedwater flow with Venturi meters, which affects that the thermal balance of reactor BWRs or PWRs this underestimated. One in broad strokes describes the application of the ultrasonic technology for the ultrasonic measurement in the flow of the feedwater system of the reactor and power to recover thermal power of the reactor. One is to the methodology developed in CFE for a calibration of the temperature transmitters of RTD's and the methodology for a calibration of the venturi flow transmitters using ultrasonic measurement. Are show the measurements in the feedwater of reactor of the temperature with RTD's and ultrasonic measurement, as well as the flow with the venturi and the ultrasonic measurement operating the reactor to the 100% of nominal thermal power, before and after the calibration of the temperature transmitters and flow. Finally, is a plan to be able to realize a recovery of thermal power of the reactor, showing as carrying out their estimations. As a result of the application of ultrasonic technology in the feedwater of reactor BWR-5 in Laguna Verde, in the Unit 1 cycle 13 it was recover an equivalent energy to a thermal power of 25 MWt in the reactor and an exit electrical power of 6 M We in the turbogenerator. Also in the Unit 2 cycle 10 it was recover an equivalent energy to a thermal power of 40 MWt in the reactor and an exit electrical power of 16 M We in the turbogenerator. (Author)

  2. 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.

  3. Nuclear reactors

    International Nuclear Information System (INIS)

    This draft chart contains graphical symbols from which the type of (nuclear) reactor can be seen. They will serve as illustrations for graphical sketches. Important features of the individual reactor types are marked out graphically. The user can combine these symbols to characterize a specific reactor type. The basic graphical symbol is a square with a point in the centre. Functional groups can be depicted for closer specification. If two functional groups are not clearly separated, this is symbolized by a dotted line or a channel. Supply and discharge lines for coolant, moderator and fuel are specified in accordance with DIN 2481 and can be further specified by additional symbols if necessary. The examples in the paper show several different reactor types. (orig./AK)

  4. Nuclear reactors

    International Nuclear Information System (INIS)

    A nuclear reactor has a large prompt negative temperature coefficient of reactivity. A reactor core assembly of a plurality of fluid-tight fuel elements is located within a water-filled tank. Each fuel element contains a solid homogeneous mixture of 50-79 w/o zirconium hydride, 20-50 w/o uranium and 0.5-1.5 W erbium. The uranium is not more than 20 percent enriched, and the ratio of hydrogen atoms to zirconium atoms is between 1.5:1 and 7:1. The core has a long lifetime, E.G., at least about 1200 days

  5. Nuclear reactors

    International Nuclear Information System (INIS)

    In a liquid cooled nuclear reactor, the combination is described for a single-walled vessel containing liquid coolant in which the reactor core is submerged, and a containment structure, primarily of material for shielding against radioactivity, surrounding at least the liquid-containing part of the vessel with clearance therebetween and having that surface thereof which faces the vessel make compatible with the liquid, thereby providing a leak jacket for the vessel. The structure is preferably a metal-lined concrete vault, and cooling means are provided for protecting the concrete against reaching a temperature at which damage would occur. (U.S.)

  6. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  7. 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

  8. Validation of updated neutronic calculation models proposed for Atucha-II PHWR. Part II: Benchmark comparisons of PUMA core parameters with MCNP5 and improvements due to a simple cell heterogeneity correction

    International Nuclear Information System (INIS)

    In 2005 the Argentine Government took the decision to complete the construction of the Atucha-II nuclear power plant, which has been progressing slowly during the last ten years. Atucha-II is a 745 MWe nuclear station moderated and cooled with heavy water, of German (Siemens) design located in Argentina. It has a pressure vessel design with 451 vertical coolant channels and the fuel assemblies (FA) are clusters of 37 natural UO2 rods with an active length of 530 cm. For the reactor physics area, a revision and update of reactor physics calculation methods and models was recently carried out covering cell, supercell (control rod) and core calculations. This paper presents benchmark comparisons of core parameters of a slightly idealized model of the Atucha-I core obtained with the PUMA reactor code with MCNP5. The Atucha-I core was selected because it is smaller, similar from a neutronic point of view, more symmetric than Atucha-II, and has some experimental data available. To validate the new models benchmark comparisons of k-effective, channel power and axial power distributions obtained with PUMA and MCNP5 have been performed. In addition, a simple cell heterogeneity correction recently introduced in PUMA is presented, which improves significantly the agreement of calculated channel powers with MCNP5. To complete the validation, the calculation of some of the critical configurations of the Atucha-I reactor measured during the experiments performed at first criticality is also presented. (authors)

  9. The reactor ALLEGRO and the sustainable nuclear energy in Central Europe

    Directory of Open Access Journals (Sweden)

    Gadó János

    2014-01-01

    Full Text Available The Visegrád-4 countries (CZ, HU, PL and SK would like to use nuclear energy on the long run. The construction of new Generation 3+ nuclear units probably belong in each country to this realm. These reactors will provide safe and cheap electric energy approximately until the end of the 21st century. In order to use nuclear energy in the 22nd century, sustainability of fuel supply shall be achieved by applying Generation 4 breeder reactors with fast spectrum. The corresponding research and development is organized now in the framework of the V4G4 Centre of Excellence establshed by the nuclear research institutes of the region with a strong technical support from the French CEA. The most important milestone of these efforts is the start-up of the ALLEGRO reactor that shall demonstrate the viability of the gas cooled fast reactor technology.

  10. The reactor ALLEGRO and the sustainable nuclear energy in Central Europe

    Science.gov (United States)

    Gadó, János

    2014-09-01

    The Visegrád-4 countries (CZ, HU, PL and SK) would like to use nuclear energy on the long run. The construction of new Generation 3+ nuclear units probably belong in each country to this realm. These reactors will provide safe and cheap electric energy approximately until the end of the 21st century. In order to use nuclear energy in the 22nd century, sustainability of fuel supply shall be achieved by applying Generation 4 breeder reactors with fast spectrum. The corresponding research and development is organized now in the framework of the V4G4 Centre of Excellence establshed by the nuclear research institutes of the region with a strong technical support from the French CEA. The most important milestone of these efforts is the start-up of the ALLEGRO reactor that shall demonstrate the viability of the gas cooled fast reactor technology.

  11. Nuclear reactors

    International Nuclear Information System (INIS)

    Disclosed is a nuclear reactor cooled by a freezable liquid has a vessel for containing said liquid and comprising a structure shaped as a container, and cooling means in the region of the surface of said structure for effecting freezing of said liquid coolant at and for a finite distance from said surface for providing a layer of frozen coolant on and supported by said surface for containing said liquid coolant. In a specific example, where the reactor is sodium-cooled, the said structure is a metal-lined concrete vault, cooling is effected by closed cooling loops containing NaK, the loops extending over the lined surface of the concrete vault with outward and reverse pipe runs of each loop separated by thermal insulation, and air is flowed through cooling pipes embedded in the concrete behind the metal lining. 7 claims, 3 figures

  12. The reactor ALLEGRO and the sustainable nuclear energy in Central Europe

    OpenAIRE

    Gadó János

    2014-01-01

    The Visegrád-4 countries (CZ, HU, PL and SK) would like to use nuclear energy on the long run. The construction of new Generation 3+ nuclear units probably belong in each country to this realm. These reactors will provide safe and cheap electric energy approximately until the end of the 21st century. In order to use nuclear energy in the 22nd century, sustainability of fuel supply shall be achieved by applying Generation 4 breeder reactors with fast spectrum. The corresponding research and de...

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    A nuclear reactor is described in which the core components, including fuel-rod assemblies, control-rod assemblies, fertile rod-assemblies, and removable shielding assemblies, are supported by a plurality of separate inlet modular units. These units are referred to as inlet module units to distinguish them from the modules of the upper internals of the reactor. The modular units are supported, each removable independently of the others, in liners in the supporting structure for the lower internals of the reactor. The core assemblies are removably supported in integral receptacles or sockets of the modular units. The liners, units, sockets and assemblies have inlet openings for entry of the fluid. The modular units are each removably mounted in the liners with fluid seals interposed between the opening in the liner and inlet module into which the fluid enters in the upper and lower portion of the liner. Each assembly is similarly mounted in a corresponding receptacle with fluid seals interposed between the openings where the fluid enters in the lower portion of the receptacle or fitting closely in these regions. As fluid flows along each core assembly a pressure drop is produced along the fluid so that the fluid which emerges from each core assembly is at a lower pressure than the fluid which enters the core assembly. However because of the seals interposed in the mountings of the units and assemblies the pressures above and below the units and assemblies are balanced and the units are held in the liners and the assemblies are held in the receptacles by their weights as they have a higher specific gravity than the fluid. The low-pressure spaces between each module and its liner and between each core assembly and its module is vented to the low-pressure regions of the vessel to assure that fluid which leaks through the seals does not accumulate and destroy the hydraulic balance

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

    International Nuclear Information System (INIS)

    The present storage facilities for spent resins in Atucha I NPP would be full within the next 5 years, accordingly some tasks are being planned to conditioning these waste and so generate enough storage capacity for the remaining life of the plant. Among this tasks is the characterization of spent resins that has several objectives: to know their chemical and radiochemical composition; the homogeneity of these parameters in the total volume of spent resins; the existence or not of compact zones; the proportion of 'crud'; the breakage degree of the micro-spheres; etc. The first step was to analyse the criterion to follow for sampling resins in the storage deposit of 40 m3. In order to take some samples from different points, a special device was required. It had to be introduced closed in the resin bed, opened to take the sample and then closed again to return to the surface. A device used in cereal industry for sampling silos to different depths was modified in its internal capacity for reducing operator dose and increasing the length of rod in order to reaching the bottom of the pit. The device was tested in cold mock up before to taking actual samples. Active resins samples, five in total up to now, were taken from deposit to different depths and kept in lead containers. After analysing the samples, the following average results were extracted: 1.7 x 105 Bq/g of Co-60, 9.7 x 105 Bq/g of Cs-137 and 774 Bq/g of total alpha, which corresponds to intermediate activity waste. The differences between the values of activity of the different samples are of up to 310 % for Co-60 and of up to 788 % for the Cs-137 what indicates a great inhomogeneity. The direct observation of resin grains, placed in a transparent glass burette, did not demonstrate an important proportion of broken or divided resins. (author)

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

    International Nuclear Information System (INIS)

    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.

  18. 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.

  19. 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)

    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)

  20. Argentinian president Cristina de Kirchner starts commissioning of Atucha II mechanical systems; Inbetriebsetzung der Maschinentechnik in Atucha II durch die Staatspraesidenten Argentiniens

    Energy Technology Data Exchange (ETDEWEB)

    Fabian, Hermann O. [NA-SA-Beratungskomitee zu Designfragen, Erlangen (Germany); Mazzantini, Oscar A. [Nuclearelectrica Argentina SA, Atucha (Argentina). Licensing and Safety

    2012-01-15

    On September 28, 2011, construction of the Atucha II heavy-water reactor plant (CNA II, 745 gross MWe) in Argentina reached another important milestone on its way to completion of the project. In the presence of Argentinian State President Cristina Fernandez de Kirchner, commissioning of the mechanical systems was initiated. The technology of the Atucha reactor facilities, which are fueled with natural uranium, moderated by heavy water (D{sub 2}O) and cooled by light water (H{sub 2}O) (Pressurized Heavy Water Reactor, PHWR), originally was developed by Siemens. These plants were designed alongside the pressurized water reactors using light water. The strict requirements applying to advanced designs in Germany as well as current findings in technical implementation were employed. Major construction and mechanical assembly work has been completed. The primary system has been installed completely. The first core was manufactured at the Argentinian fuel element factory of Conuar and placed into the plant's dry store, which also contains the control and shutdown rods. The heavy water (D{sub 2}O) needed for nuclear operation has been produced and is being stored on the plant premises. Extensive commissioning work is going on in the building for the switching systems including the main and the emergency control rooms. The turbine hall is largely complete, as are the facilities for cooling water supply. The commissioning phase of the plant has been started involving considerable manpower. Some procedures for the later part of the commissioning phase still need to be completed. As a consequence of its sound and qualified basic design and the incorporation of other, new findings, Atucha II in its optimized design will prove to be a safe and reliable plant in operation. The start of commercial operation of Atucha II has been announced for probably late 2012. (orig.)

  1. Visual interface for the automation of the instrumented pendulum of Charpy tests used in the surveillance program of reactors vessel of nuclear power plants; Interfase visual para la automatizacion del pendulo instrumentado de pruebas Charpy utilizado en el programa de vigilancia de la vasija de reactores de centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Rojas S, A.S.; Sainz M, E.; Ruiz E, J.A. [ININ, Carretera Mexico-Toluca Km.36.5, Mpio. de Ocoyoacac, Estado de Mexico (Mexico)]. E-mail: asrs@nuclear.inin.mx; esm@nuclear.inin.mx; jare@nuclear.inin.mx

    2004-07-01

    Inside the Programs of Surveillance of the nuclear power stations periodic information is required on the state that keep the materials with those that builds the vessel of the reactor. This information is obtained through some samples or test tubes that are introduced inside the core of the reactor and it is observed if its physical characteristics remain after having been subjected to the radiation changes and temperature. The rehearsal with the instrumented Charpy pendulum offers information on the behavior of fracture dynamics of a material. In the National Institute of Nuclear Research (ININ) it has an instrumented Charpy pendulum. The operation of this instrument is manual, having inconveniences to carry out rehearsals with radioactive material, handling of high and low temperatures, to fulfill the normative ones for the realization of the rehearsals, etc. In this work the development of a computational program is presented (virtual instrument), for the automation of the instrumented pendulum. The system has modules like: Card of data acquisition, signal processing, positioning system, tempered system, pneumatic system, compute programs like it is the visual interface for the operation of the instrumented Charpy pendulum and the acquisition of impact signals. This system shows that given the characteristics of the nuclear industry with radioactive environments, the virtual instrumentation and the automation of processes can contribute to diminish the risks to the personnel occupationally exposed. (Author)

  2. Nuclear research reactors

    International Nuclear Information System (INIS)

    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.)

  3. Nuclear reactor

    International Nuclear Information System (INIS)

    In order to reduce neutron embrittlement of the pressue vessel of an LWR, blanked off elements are fitted at the edge of the reactor core, with the same dimensions as the fuel elements. They are parallel to each other, and to the edge of the reactor taking the place of fuel rods, and are plates of neutron-absorbing material (stainless steel, boron steel, borated Al). (HP)

  4. Nuclear Energy in Central Europe 98, Proceedings

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  6. ARGOS PHWR 380. Argentine offer of a safer pressurized heavy-water reactor of 380 MW. '...a many-eyed guardian...' concerned about nuclear power plant safety

    International Nuclear Information System (INIS)

    Reactor vendors in most countries have had lean pickings for the past decade, and ordering seems unlikely to show much growth until the shock wave from the Chernobyl accident has died away. Paradoxically, however, at least one firm sees a niche in the market. ENACE - the Empresa Nuclear Argentina de Centrales Electricas, or Argentine Nuclear Power Plant Corporation - is stepping out into the market place with a newly-designed 380 MWe nuclear power plant. The plant is equipped with a pressurized heavy-water reactor of the pressure vessel type (PHWR). ENACE has adopted new boundary design conditions and has embodied a number of special features to assure safety and economy in operation. The major shareholder in ENACE is the Argentine National Atomic Energy Commission (CNEA). ENACE is the architect-engineer for the NPP projects of the Argentine nuclear programme. It has a licensing agreement with Siemens AG's Kraftwerk Union AG, which is its minor shareholder. Under this agreement, ENACE has the right to use the Siemens-KWU PHWR technology, which was originally developed for the MZFR reactor in the Federal Republic of Germany, as well as their know-how in pressurized (light-) water reactors (PWRs) design and construction. The CNEA also has agreements with Atomic Energy of Canada Ltd. for the transfer of technology related to CANDU-type PTHWRs. The CNEA and ENACE have acquired considerable practical experience from the construction and operation of the 367 MWe Atucha I PHWR and the 648 MWe Embalse PTHWR; ENACE is currently building Argentina's third nuclear power plant, Atucha II, a 745 MWe PHWR. (author)

  7. Types of Nuclear Reactors

    International Nuclear Information System (INIS)

    The presentation is based on the following areas: Types of Nuclear Reactors, coolant, moderator, neutron spectrum, fuel type, pressurized water reactor (PWR), boiling water reactor (BWR) reactor pressurized heavy water (PHWR), gas-cooled reactor, RBMK , Nuclear Electricity Generation,Challenges in Nuclear Technology Deployment,EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR.

  8. 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)

    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.)

  9. Nuclear reactor

    International Nuclear Information System (INIS)

    In an improved reactor core for a high conversion BWR reactor, Pu-breeding type BWR type reactor, Pu-breeding type BWR type rector, FEBR type reactor, etc., two types of fuel assemblies are loaded such that fuel assemblies using a channel box of a smaller irradiation deformation ratio are loaded in a high conversion region, while other fuel assemblies are loaded in a burner region. This enables to suppress the irradiation deformation within an allowable limit in the high conversion region where the fast neutron flux is high and the load weight from the inside of the channel box due to the pressure loss is large. At the same time, the irradiation deformation can be restricted within an allowable limit without deteriorating the neutron economy in the burner region in which fast neutron flux is low and the load weight from the inside of the channel box is small since a channel box with smaller neutron absorption cross section or reduced wall thickness is charged. As a result, it is possible to prevent structural deformations such as swelling of the channel box, bending of the entire assemblies, bending of fuel rods, etc. (K.M.)

  10. Nuclear reactor

    International Nuclear Information System (INIS)

    The liquid metal (sodium) cooled fast breeder reactor has got fuel subassemblies which are bundled and enclosed by a common can. In order to reduce bending of the sides of the can because of the load caused by the coolant pressure the can has got a dodecagon-shaped crosssection. The surfaces of the can may be of equal width. One out of two surfaces may also be convex towards the center. (RW)

  11. Nuclear reactor

    International Nuclear Information System (INIS)

    A detector having high sensitivity to fast neutrons and having low sensitivity to thermal neutrons is disposed for reducing influences of neutron detector signals on detection values of neutron fluxes when the upper end of control rod pass in the vicinity of the neutron flux detector. Namely, the change of the neutron fluxes is greater in the thermal neutron energy region while it is smaller in the fast neutron energy region. This is because the neutron absorbing cross section of B-10 used as neutron absorbers of control rods is greater in the thermal neutron region and it is smaller in the fast neutron region. As a result, increase of the neutron detection signals along with the local neutron flux change can be reduced, and detection signals corresponding to the reactor power can be obtained. Even when gang withdrawal of operating a plurality of control rods at the same time is performed, the reactor operation cycle can be measured accurately, thereby enabling to shorten the reactor startup time. (N.H.)

  12. Argentinian president Cristina de Kirchner starts commissioning of Atucha II mechanical systems

    International Nuclear Information System (INIS)

    On September 28, 2011, construction of the Atucha II heavy-water reactor plant (CNA II, 745 gross MWe) in Argentina reached another important milestone on its way to completion of the project. In the presence of Argentinian State President Cristina Fernandez de Kirchner, commissioning of the mechanical systems was initiated. The technology of the Atucha reactor facilities, which are fueled with natural uranium, moderated by heavy water (D2O) and cooled by light water (H2O) (Pressurized Heavy Water Reactor, PHWR), originally was developed by Siemens. These plants were designed alongside the pressurized water reactors using light water. The strict requirements applying to advanced designs in Germany as well as current findings in technical implementation were employed. Major construction and mechanical assembly work has been completed. The primary system has been installed completely. The first core was manufactured at the Argentinian fuel element factory of Conuar and placed into the plant's dry store, which also contains the control and shutdown rods. The heavy water (D2O) needed for nuclear operation has been produced and is being stored on the plant premises. Extensive commissioning work is going on in the building for the switching systems including the main and the emergency control rooms. The turbine hall is largely complete, as are the facilities for cooling water supply. The commissioning phase of the plant has been started involving considerable manpower. Some procedures for the later part of the commissioning phase still need to be completed. As a consequence of its sound and qualified basic design and the incorporation of other, new findings, Atucha II in its optimized design will prove to be a safe and reliable plant in operation. The start of commercial operation of Atucha II has been announced for probably late 2012. (orig.)

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

    Directory of Open Access Journals (Sweden)

    Oscar Mazzantini

    2011-01-01

    Full Text Available 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.

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

    International Nuclear Information System (INIS)

    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)

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    Cover gas spaces for primary coolant vessel, such as a reactor container, a pump vessel and an intermediate heat exchanger vessel are in communication with each other by an inverted U-shaped pressure conduit. A transmitter and a receiver are disposed to the pressure conduit at appropriate positions. If vibration frequencies (pressure vibration) from low frequency to high frequency are generated continuously from the transmitter to the inside of the communication pipe, a resonance phenomenon (air-column resonance oscillation) is caused by the inherent frequency or the like of the communication pipe. The frequency of the air-column resonance oscillation is changed by the inner diameter and the clogged state of the pipelines. Accordingly, by detecting the change of the air-column oscillation characteristics by the receiver, the clogged state of the flow channels in the pipelines can be detected even during the reactor operation. With such procedures, steams of coolants flowing entrained by the cover gases can be prevented from condensation and coagulation at a low temperature portion of the pipelines, otherwise it would lead clogging in the pipelines. (I.N.)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  18. Introduction of Nuclear Reactor Engineering

    International Nuclear Information System (INIS)

    This book introduces development, status, supply and demand and resource of nuclear reactor. It deals with basic knowledge of nuclear reactor, which are reactor system, heat recovery in reactor core, structural feature in reactor, materials of structure in reactor, shielding of gamma ray, shielding of reactor, safety and environmental problem of nuclear power plant, nuclear fuel and economical efficiency of nuclear energy.

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

    International Nuclear Information System (INIS)

    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)

  20. 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

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

    International Nuclear Information System (INIS)

    The plant response to the group of events 'large coolant loss' in order to evaluate the associated risk is analyzed. The event that covers all events of similar sequence due to its evolution features, being also the most demanded, is selected as starting event. The representative event is the 'guillotine type rupture of cold primary branch'. An annual occurrence frequency of 10/year is assumed for this event. The safety systems, when the event occurs, must assure the reactor shutdown and the core cooling, creating a heat sink to remove the decay heat. The annual frequency of core meltdown due to great loss of coolant is obtained multiplying the annual frequency of the starting event by the probability of failure of involved safety systems. By means of failure trees, the following is obtained: a) probability of failure to demand of the boron injection shutdown system = 4 x 10-2; b) probability of failure to demand of the high pressure safety injection = 3 x 10-3; c) probability of emergency cooling system failure = 4.4 x 10-2. Therefore, the three possible sequences of core meltdown have the following frequencies: λ1 = 4 x 10-6/year λ2 = 3 x 10-7/year λ3 = 4.4 x 10-6/year. (Author)

  2. El reactor nuclear colombiano y la agencia de actores no humanos en los estudios sociales de la ciencia

    OpenAIRE

    Juan Andrés León Gómez

    2009-01-01

    Basándose en su trabajo de grado de historiador sobre los primeros años del programa nuclear colombiano, el autor explora la relevancia del concepto de actor no humano originario de las teorías de actores-red. Se muestra cómo el análisis de la consolidación del Instituto de Asuntos Nucleares debe incluir un objeto inanimado, el reactor nuclear, como actor social fundamental. PALABRAS CLAVES. actores no humanos, comunidad científica, energía nuclear

  3. 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)

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

    International Nuclear Information System (INIS)

    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

  5. 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.

  6. Hydrogen Pick up in Zircaloy-4: Effects in the Dimensional Stability of Structural Components under Nuclear Reactor Operating Conditions

    OpenAIRE

    Abraham David Banchik; Cintia Paola Fagundez; Pablo Vizcaíno

    2010-01-01

    In the present work, the expansion coefficient due to hydrogen incorporation was measured for the axial direction of a Zircaloy-4 cooling channel, similar to that installed in the Atucha I PHWR, Argentina, trying to simulate the nuclear power reactor operating conditions. As a first step, the solubility curve of hydrogen in Zircloy-4 was determined by two techniques: differential scanning calorimetry and differential dilatometry. The comparison with classical literature curves showed a good a...

  7. Report on the Fourth Reactor Refueling. Laguna Verde Nuclear Central. Unit 1. April-May 1995

    International Nuclear Information System (INIS)

    The fourth refueling of the Unit 1 of Laguna Verde Nuclear Central was executed in the period of April 17 to May 31 of 1995 with the participation of a task group of 358 persons, included technicians and radiation protection officials and auxiliaries.The radiation monitoring and radiological surveillance to the workers was present length ways the refueling process and always attached to the ALARA criteria. The check points for radiation levels were set at: primary container or dry well, reloading floor, decontamination room (level 10.5), turbine building and radioactive waste building. To take advantage of the refueling process, rooms 203 and 213 of the turbine buildings were subject to inspection and maintenance work in valves, heaters and drains of heaters. Management aspects as personnel selection and training, costs, and countable are also presented in this report. Owing to the high cost of man-hour of the members of the ININ staff, its participation in the refueling process was in smaller number than years before. (Author)

  8. Compatibility of sodium with ceramic oxides employed in nuclear reactors; Compatibilidad del sodio con oxidos ceramicos utilizados en reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Acena Moreno, V.

    1981-07-01

    This work is a review of experiments carried out up to the present time on the corrosion and compatibility of ceramic oxides with liquid sodium at temperatures corresponding to those in fast breeder reactors. The review also includes the results of a thermo-dynamic/liquid sodium reactions. The exercise has been conducted with a view to effecting experimental studies in the future. (Author)

  9. Nuclear reactor theory

    International Nuclear Information System (INIS)

    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.)

  10. Desarrollo e implantación en un APS de una metodologia para identificar los fallos de inicio de vida en una Central Nuclear

    OpenAIRE

    González Celades, María

    2015-01-01

    El proyecto Desarrollo e implantación en un Análisis Probabilista de Seguridad de una metodología para identificar los fallos de inicio de vida en una central nuclear tiene como objetivo el desarrollo de una herramienta para identificar qué fallos de componente forman parte del período de inicio de infancia, de una central nuclear. Estos fallos deberían ser eliminados de la Base de Datos del Análisis Probabilista de Seguridad (BD APS) ya que no son representativos de la operación real de la c...

  11. Fossil nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Maurette, M.

    1976-01-01

    The discussion of fossil nuclear reactors (the Oklo phenomenon) covers the earth science background, neutron-induced isotopes and reactor operating conditions, radiation-damage studies, and reactor modeling. In conclusion possible future studies are suggested and the significance of the data obtained in past studies is summarized. (JSR)

  12. Research Nuclear Reactors

    International Nuclear Information System (INIS)

    Published in English and in French, this large report first proposes an overview of the use and history of research nuclear reactors. It discusses their definition, and presents the various types of research reactors which can be either related to nuclear power (critical mock-ups, material test reactors, safety test reactors, training reactors, prototypes), or to research (basic research, industry, health), or to specific particle physics phenomena (neutron diffraction, isotope production, neutron activation, neutron radiography, semiconductor doping). It reports the history of the French research reactors by distinguishing the first atomic pile (ZOE), and the activities and achievements during the fifties, the sixties and the seventies. It also addresses the development of instrumentation for research reactors (neutron, thermal, mechanical and fission gas release measurements). The other parts of the report concern the validation of neutronics calculations for different reactors (the EOLE water critical mock-up, the MASURCA air critical mock-up dedicated to fast neutron reactor study, the MINERVE water critical mock-up, the CALIBAN pulsed research reactor), the testing of materials under irradiation (OSIRIS reactor, laboratories associated with research reactors, the Jules Horowitz reactor and its experimental programs and related devices, irradiation of materials with ion beams), the investigation of accident situations (on the CABRI, Phebus, Silene and Jules Horowitz reactors). The last part proposes a worldwide overview of research reactors

  13. Ground acceleration in a nuclear power plant; Aceleracion del suelo en una central nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Pena G, P.; Balcazar, M.; Vega R, E., E-mail: pablo.pena@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    A methodology that adopts the recommendations of international organizations for determining the ground acceleration at a nuclear power plant is outlined. Systematic presented here emphasizes the type of geological, geophysical and geotechnical studies in different areas of influence, culminating in assessments of Design Basis earthquake and the earthquake Operating Base. The methodology indicates that in regional areas where the site of the nuclear power plant is located, failures are identified in geological structures, and seismic histories of the region are documented. In the area of detail geophysical tools to generate effects to determine subsurface propagation velocities and spectra of the induced seismic waves are used. The mechanical analysis of drill cores allows estimating the efforts that generate and earthquake postulate. Studies show that the magnitude of the Fukushima earthquake, did not affect the integrity of nuclear power plants due to the rocky settlement found. (Author)

  14. Research nuclear reactors

    International Nuclear Information System (INIS)

    Since the divergence of the first nuclear reactor in 1942, about 600 research or test reactors have been built throughout the world. Today 255 research reactors are operating in 57 countries and about 70% are over 25 years old. Whereas there are very few reactor types for power plants because of rationalization and standardisation, there is a great diversity of research reactors. We can divide them into 2 groups: heavy water cooled reactors and light water moderated reactors. Heavy water cooled reactors are dedicated to the production of high flux of thermal neutrons which are extracted from the core by means of neutronic channels. Light water moderated reactors involved pool reactors and slightly pressurized closed reactors, they are polyvalent but their main purposes are material testing, technological irradiations, radionuclide production and neutron radiography. At the moment 8 research reactors are being built in Canada, Germany, Iran, Japan, Kazakhstan, Morocco, Russia and Slovakia and 8 others are planned in 7 countries (France, Indonesia, Nigeria, Russia, Slovakia, Thailand and Tunisia. Different research reactors are described: Phebus, Masurca, Phenix and Petten HFR. The general principles of nuclear safety applied to test reactors are presented. (A.C.)

  15. Análisis de Ruido en Reactores PWR

    OpenAIRE

    Bermejo Piñar, Juan Antonio

    2015-01-01

    Esta tesis se ha llevado a cabo persiguiendo dos objetivos principales: uno de ellos es el desarrollo y la aplicación de modelos para el mantenimiento predictivo de sensores en centrales nucleares, y el otro es profundizar en el entendimiento de los fenómenos que tienen influencia en el ruido de la señal de los detectores de neutrones de los reactores de agua a presión con ayuda de herramientas de simulación 3D. Para el desarrollo de los trabajos se ha contado con medidas de ruido de reactore...

  16. Nuclear power plants making a comeback in Japan; El retorno de la centrales nucleares en Japon

    Energy Technology Data Exchange (ETDEWEB)

    Torralbo, J. R.

    2016-08-01

    We reproduce in this magazine the interesting article published by the president of the SNE in issue 46 of Cuadernos de Energia in October 2015, which describes the events that have taken place since the March 11, 2011 earthquake in Japan, the largest in its history, and the subsequent tsunami, which affected the Fukushima power plant, as well as the measures implemented since then and how some of this country nuclear power plants are being started up again. (Author)

  17. Knowledge management in nuclear power plants; Gestion del conocimiento en las centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Cal, C. de la; Barasoain, F.; Buedo, J. L.

    2013-03-01

    This article aims to show the importance of knowledge management from different perspectives. In this first part part of the article, the overall approach that performs CNAT of knowledge management is described. In the second part, a specific aspect of knowledge management in ANAV, tacit knowledge transfer is showed. finally, the third part discusses the strategies and actions that are followed in CNCO for knowledge management. All this aims to show an overview of knowledge management held in the Spanish Nuclear Power Plants. (Author)

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

    International Nuclear Information System (INIS)

    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)

  19. Nuclear reactor internals arrangement

    International Nuclear Information System (INIS)

    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

  20. 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)

    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)

  1. Reactors. Nuclear propulsion ships

    International Nuclear Information System (INIS)

    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.)

  2. Nuclear reactor repairing device

    International Nuclear Information System (INIS)

    Purpose: To enable free repairing of an arbitrary position in an LMFBR reactor. Constitution: A laser light emitted from a laser oscillator installed out of a nuclear reactor is guided into a portion to be repaired in the reactor by using a reflecting mirror, thereby welding or cutting it. The guidance of the laser out of the reactor into the reactor is performed by an extension tube depending into a through hole of a rotary plug, and the guidance of the laser light into a portion to be repaired is performed by the transmitting and condensing action of the reflecting mirror. (Kamimura, M.)

  3. Nuclear reactor physics course for reactor operators

    International Nuclear Information System (INIS)

    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

  4. Special lecture on nuclear reactor

    International Nuclear Information System (INIS)

    This book gives a special lecture on nuclear reactor, which is divided into two parts. The first part has explanation on nuclear design of nuclear reactor and analysis of core with theories of integral transports, diffusion Nodal, transports Nodal and Monte Carlo skill parallel computer and nuclear calculation and speciality of transmutation reactor. The second part deals with speciality of nuclear reactor and control with nonlinear stabilization of nuclear reactor, nonlinear control of nuclear reactor, neural network and control of nuclear reactor, control theory of observer and analysis method of Adomian.

  5. Reactor core of nuclear reactor

    International Nuclear Information System (INIS)

    In a BWR type nuclear reactor, the number of first fuel assemblies (uranium) loaded in a reactor core is smaller than that of second fuel assemblies (mixed oxide), the average burnup degree upon take-out of the first fuel assemblies is reduced to less than that of the second fuel assemblies, and the number of the kinds of the fuel rods constituting the first fuel assemblies is made smaller than that of the fuel rods constituting the second fuel assemblies. As a result, the variety of the plutonium enrichment degree is reduced to make the distribution of the axial enrichment degree uniform, thereby enabling to simplify the distribution of the enrichment degree. Then the number of molding fabrication steps for MOX fuel assemblies can be reduced, thereby enabling to reduce the cost for molding and fabrication. (N.H.)

  6. 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

  7. The nuclear soliton reactor

    International Nuclear Information System (INIS)

    The basic reactor physics of a completely novel nuclear fission reactor design - the soliton-reactor - is presented on the basis of a simple model. In such a reactor, the neutrons in the critical region convert either fertile material in the adjacent layers into fissile material or reduce the poisoning of fissile material in such a manner that successively new critical regions emerge. The result is an autocatalytically driven burn-up wave which propagates throughout the reactor. Thereby, the relevant characteristic spatial distributions (neutron flux, specific power density and the associated particle densities) are solitons - wave phenomena resulting from non-linear partial differential equations which do not change their shape during propagation. A qualitativley new kind of harnessing nuclear fission energy may become possible with fuel residence times comparable with the useful lifetime of the reactor system. In the long run, fast breeder systems which exploit the natural uranium and thorium resources, without any reprocessing capacity are imaginable. (orig.)

  8. Evaluation of cable ageing in Nuclear Power Plants; Evaluacion del envejecimiento de cables en centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Lopez Vergara, T. [Empresarios Agrupados, A. I. E. Madrid (Spain); Alonso Chicote, J. [TECNATOM, S. A. (Spain); Burnay, S. [AEA Technology (UK)

    2000-07-01

    The majority of power, control and instrumentation cables in nuclear power plants use polymers as their basic material for insulation and jacket. In many cases, these cables form part of safety-related circuits and should therefore be capable of operating correctly under both normal and accident conditions. Since polymeric materials are degraded by the long term action of the radiation and thermal environments found in the plant, it is important to be able to establish the cable condition during the plant lifetime. Nowadays there are a number of different methods to evaluate the remaining lifetime of cables. In the case of new plants, or new cables in old plants, accelerated ageing tests and predictive models can be used to establish the behaviour of the cable materials under operating conditions. There are verified techniques and considerable experience in the definition of predictive models. This type of approach is best carried out during the commissioning stage or in the early stages of operation. In older plants, particularly where there is a wide range of cable types in use, it is more appropriate to use condition monitoring methods to establish the state of degradation of cables in-plant. Over the last 10 years there have been considerable developments in methods for condition monitoring of cables and a tool-box of practical techniques are now available. There is no single technique which is suitable for all cable materials but the range of methods covers nearly all of the types currently in use, at present, the most established methods are the indented, thermal analysis (OIT, OITP and TGA) and dielectric loss measurements, All of these are either non-destructive methods or require only micro-samples of material. (Author) 15 refs.

  9. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    An improved nuclear power reactor fuel element is described which consists of fuel rods, rod guide tubes and an end plate. The system allows direct access to an end of each fuel rod for inspection purposes. (U.K.)

  10. Nuclear reactor simulator

    International Nuclear Information System (INIS)

    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)

  11. 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)

  12. Physics of nuclear reactors

    International Nuclear Information System (INIS)

    This manual covers all the aspects of the science of neutron transport in nuclear reactors and can be used with great advantage by students, engineers or even reactor experts. It is composed of 18 chapters: 1) basis of nuclear physics, 2) the interactions of neutrons with matter, 3) the interactions of electromagnetic radiations and charged-particles with matter, 4) neutron slowing-down, 5) resonant absorption, 6) Doppler effect, 7) neutron thermalization, 8) Boltzmann equation, 9) calculation methods in neutron transport theory, 10) neutron scattering, 11) reactor reactivity, 12) theory of the critical homogenous pile, 13) the neutron reflector, 14) the heterogeneous reactor, 15) the equations of the fuel cycle, 16) neutron counter-reactions, 17) reactor kinetics, and 18) calculation methods in neutron scattering

  13. "Gato encerrado" o "confidencialidad comercial": el problema del secreto en la venta de un reactor nuclear "Gato encerrado" or "confidencialidad comercial": a secret problem regards the sale of a nuclear reactor

    OpenAIRE

    Ana Spivak L'Hoste

    2006-01-01

    En este artículo se propone abordar uno de los aspectos medulares de la disputa pública que surge a partir de la venta de un reactor nuclear diseñado por una empresa argentina a Australia en el año 2000. Se trata del problema del secreto que atraviesa, para quienes están involucrados en la misma y desde dos perspectivas claramente diferentes, la operación comercial. A través del análisis de las categorías locales que hacen a la configuración del problema del secreto, gato encerrado y secreto ...

  14. SNAP Nuclear Space Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Corliss, William R

    1966-01-01

    This booklet describes the principles of nuclear-reactor space power plants and shows how they will contribute to the exploration and use of space. It compares them with chemical fuels, solar cells, and systems using energy from radioisotopes. The SNAP (Systems for Nuclear Auxiliary Power) Program, begun in 1955, is described.

  15. 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)

  16. Management of Spent Nuclear Fuel from Nuclear Power Plant Reactor

    International Nuclear Information System (INIS)

    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)

  17. Nuclear reactor fuel elements

    International Nuclear Information System (INIS)

    A nuclear reactor fuel element comprising a column of vibration compacted fuel which is retained in consolidated condition by a thimble shaped plug. The plug is wedged into gripping engagement with the wall of the sheath by a wedge. The wedge material has a lower coefficient of expansion than the sheath material so that at reactor operating temperature the retainer can relax sufficient to accommodate thermal expansion of the column of fuel. (author)

  18. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    The description is given of a water cooled nuclear reactor comprising a core, cooling water that rises through the core, vertical guide tubes located inside the core and control rods vertically mobile in the guide tubes. In this reactor the cooling water is divided into a first part introduced at the bottom end of the core and rising through it and a second part introduced at the top end of the guide tubes so as to drop in them

  19. 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.

  20. Integral nuclear reactor

    International Nuclear Information System (INIS)

    The invention deals with an inprovement of the design of an integral pressurized water nuclear reactor. A typical embodyment of the invention includes a generally cylindrical pressure vessel that is assembled from three segments which are bolted together at transverse joints to form a pressure tight unit that encloses the steam generator and the reactor. The new construction permits primary to secondary coolant heat exchange and improved control rod drive mecanisms which can be exposed for full service access during reactor core refueling, maintenance and inspection

  1. Generalities about nuclear reactors

    International Nuclear Information System (INIS)

    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)

  2. Coupling Systems of Five CARA Fuel Bundles for Atucha I

    International Nuclear Information System (INIS)

    This paper describe the mechanical design of two options for the coupling systems of five CARA fuel bundles, to be used in the Atucha I nuclear power plant. These systems will be hydraulic tested in a low pressure loop to know their hydraulic loss of pressure

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  6. Spectrographic determination of metallic impurities in organic coolants for nuclear reactors; Determinacion espectrografica de impurezas metalicas en refrigerantes organicos para reactores nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Martin Munoz, M.; Alvarez Gonzalez, F.

    1969-07-01

    A spectrochemical method for determining metallic impurities in organic coolants for nuclear reactors is given. The organic matter in solid samples is eliminated by controlled distillation and dry ashing in the presence of magnesium oxide as carrier. Liquid, samples are vacuum distillated. The residue is analyzed by carrier distillation and by total burning techniques. The analytical results are discussed and compared with those obtained destroying the organic matter without carrier and using the copper spark technique. (Author) 12 refs.

  7. ATUCHA I NPP - Emergency drill practice

    International Nuclear Information System (INIS)

    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

  8. Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Cason, D.L.; Hicks, S.C. [eds.

    1992-01-01

    This publication 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 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 the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, 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.

  9. Materials for nuclear reactors

    International Nuclear Information System (INIS)

    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., UO2 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 UO2 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)

  10. Nuclear reactor power monitor

    International Nuclear Information System (INIS)

    The device of the present invention monitors phenomena occurred in a nuclear reactor more accurately than usual case. that is, the device monitors a reactor power by signals sent from a great number of neutron monitors disposed in the reactor. The device has a means for estimating a phenomenon occurred in the reactor based on the relationship of a difference of signals between each of the great number of neutron monitors to the positions of the neutron monitors disposed in the reactor. The estimation of the phenomena is conducted by, for example, conversion of signals sent from the neutron monitors to a code train. Then, a phenomenon is estimated rapidly by matching the code train described above with a code train contained in a data base. Further. signals sent from the neutron monitors are processed statistically to estimate long term and periodical phenomena. As a result, phenomena occurred in the reactor are monitored more accurately than usual case, thereby enabling to improve reactor safety and operationability. (I.S.)

  11. CANDU nuclear reactor technology

    International Nuclear Information System (INIS)

    AECL has over 40 years of experience in the nuclear field. Over the past 20 years, this unique Canadian nuclear technology has made a worldwide presence, In addition to 22 CANDU reactors in Canada, there are also two in India, one in Pakistan, one in Argentina, four in Korea and five in Romania. CANDU advancements are based on evolutionary plant improvements. They consist of system performance improvements, design technology improvements and research and development in support of advanced nuclear power. Given the good performance of CANOU plants, it is important that this CANDU operating experience be incorporated into new and repeat designs

  12. Rol del fallo mecánico en la optimización del mantenimiento en una central nuclear//Role of the mechanical failure during the maintenance optimization in the nuclear power plant

    OpenAIRE

    Antonio Torres - Valle

    2012-01-01

    Entre las más recientes aplicaciones del Análisis Probabilista de Seguridad (1997 – 2003) de la Central Nuclear Embalse en Argentina, está el Programa de Mantenimiento Orientado a la Seguridad (2006– 2009) el cual se ha desarrollado con el empleo de la metodología de Mantenimiento Centrado en la Confiabilidad (RCM en inglés). El objetivo general del artículo es demostrar la alta contribución de los fallos mecánicos en el diseño de las políticas de mantenimiento de varios sistemas de la instal...

  13. 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)

  14. Dispositivo de posicionamiento de muestras biológicas para su irradiación en un canal radial de un reactor nuclear

    OpenAIRE

    Maritza Rodríguez - Gual; Felix Mas – Milián; Airton Deppman; Paulo Rogerio Pinto - Coelho

    2011-01-01

    Por la demanda de un dispositivo experimental para el posicionamiento de las muestras biológicas para su irradiación en un canal radial de un reactor nuclear de investigaciones en funcionamiento, se construyó y se puso en marcha un dispositivo para la colocación y retirada de las muestras en la posición de irradiación de dicho canal. Se efectuaron las valoraciones económicas comparando con otro tipo de dispositivo con las mismas funciones. Este trabajo formó parte de un proyecto internacional...

  15. Nuclear reactor constructions

    International Nuclear Information System (INIS)

    A nuclear reactor construction comprising a reactor core submerged in a pool of liquid metal coolant in a primary vessel which is suspended from the roof structure of a containment vault. Control rods supported from the roof structure are insertable in the core which is carried on a support structure from the wall of the primary vessel. To prevent excessive relaxation of the support structure whereby the control rods would be displaced relative to the core, the support structure incorporates a normally inactive secondary structure designed to become effective in bracing the primary structure against further relaxation beyond a predetermined limit. (author)

  16. Comparison between a finite difference model (PUMA) and a finite element model (DELFIN) for simulation of the reactor of the atomic power plant of Atucha I

    International Nuclear Information System (INIS)

    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

  17. Nuclear Rocket Engine Reactor

    CERN Document Server

    Lanin, Anatoly

    2013-01-01

    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.

  18. Problems in producing nuclear reactor for medical isotopes and the Global Crisis of molybdenum supply; Problemas en la produccion en reactores nucleares de isotopos con fines medicos y la crisis mundial de suministro de molibdeno ({sup 9}9Mo)

    Energy Technology Data Exchange (ETDEWEB)

    Zubiarrain, A.

    2011-07-01

    Nuclear medicine uses drugs that incorporate a radioactive isotope radiopharmaceuticals. Every year are performed, worldwide, 35 million nuclear medicine procedures, of which 80% are done with radiopharmaceuticals containing the isotope, molybdenum-99, produced in nuclear reactors. In recent years, there have been several supply crisis of molybdenum-99, which have hampered diagnostic procedure with technitium-99m. (Author)

  19. Utilization of nuclear research reactors

    International Nuclear Information System (INIS)

    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)

  20. Nuclear power reactor physics

    International Nuclear Information System (INIS)

    The purpose of this book is to explain the physical working conditions of nuclear reactors for the benefit of non-specialized engineers and engineering students. One of the leading ideas of this course is to distinguish between two fundamentally different concepts: - a science which could be called neutrodynamics (as distinct from neutron physics which covers the knowledge of the neutron considered as an elementary particle and the study of its interactions with nuclei); the aim of this science is to study the interaction of the neutron gas with real material media; the introduction will however be restricted to its simplified expression, the theory and equation of diffusion; - a special application: reactor physics, which is introduced when the diffusing and absorbing material medium is also multiplying. For this reason the chapter on fission is used to introduce this section. In practice the section on reactor physics is much longer than that devoted to neutrodynamics and it is developed in what seemed to be the most relevant direction: nuclear power reactors. Every effort was made to meet the following three requirements: to define the physical bases of neutron interaction with different materials, to give a correct mathematical treatment within the limit of necessary simplifying hypotheses clearly explained; to propose, whenever possible, numerical applications in order to fix orders of magnitude

  1. Nuclear reactor constructions

    International Nuclear Information System (INIS)

    An improvement in the construction of liquid metal cooled nuclear reactors of the kind in which the fuel assembly is submerged in a pool of coolant contained by a primary vessel housed in a concrete vault, is described. In this modification the roof of the vault carries heat exchangers immersed in the pool of coolant, the lower ends of which are hydraulically damped against oscillation caused by seismic disturbances. (U.K.)

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

    OpenAIRE

    O. Mazzantini; F. D'Auria; M. Pecchia; Parisi, C

    2011-01-01

    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 ana...

  3. 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)

  4. Towards nuclear fusion reactors

    International Nuclear Information System (INIS)

    In the middle of 21st century, the population on the earth is expected to double, and the energy that mankind consumes to triple. The nuclear fusion which is said the ultimate energy source for mankind is expected to solve this energy problem. As for fusion reactors, fuel materials exist inexhaustibly, distributing evenly, they have high safety in principle, the product of burning is harmless nonradioactive substance that does not require the treatment and disposal, and the attenuation of induced radioactivity due to neutrons is quick and the effect to global environment is little. The basic plan of second stage nuclear fusion research and development was decided in 1975, aiming at attaining the critical plasma condition. JT-60 has attained it in 1987. The project of international thermonuclear fusion experimental reactor (ITER) was started, and the conceptual design was carried out. Under such background, the third stage basic plan was decided in 1992, and its objective is self ignition condition, long time burning and the basis of the reactor engineering technology. The engineering design of the ITER is investigated. (K.I.)

  5. Nuclear reactor building

    Science.gov (United States)

    Gou, Perng-Fei; Townsend, Harold E.; Barbanti, Giancarlo

    1994-01-01

    A reactor building for enclosing a nuclear reactor includes a containment vessel having a wetwell disposed therein. The wetwell includes inner and outer walls, a floor, and a roof defining a wetwell pool and a suppression chamber disposed thereabove. The wetwell and containment vessel define a drywell surrounding the reactor. A plurality of vents are disposed in the wetwell pool in flow communication with the drywell for channeling into the wetwell pool steam released in the drywell from the reactor during a LOCA for example, for condensing the steam. A shell is disposed inside the wetwell and extends into the wetwell pool to define a dry gap devoid of wetwell water and disposed in flow communication with the suppression chamber. In a preferred embodiment, the wetwell roof is in the form of a slab disposed on spaced apart support beams which define therebetween an auxiliary chamber. The dry gap, and additionally the auxiliary chamber, provide increased volume to the suppression chamber for improving pressure margin.

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

    International Nuclear Information System (INIS)

    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.)

  7. Statistical analysis about corrosion in nuclear power plants; Analisis estadistico de la corrosion en centrales nucleares de potencia

    Energy Technology Data Exchange (ETDEWEB)

    Naquid G, C.; Medina F, A.; Zamora R, L. [Instituto Nacional de Investigaciones Nucleares, Gerencia de Ciencia de Materiales, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2000-07-01

    Nowadays, it has been carried out the investigations related with the structure degradation mechanisms, systems or and components in the nuclear power plants, since a lot of the involved processes are the responsible of the reliability of these ones, of the integrity of their components, of the safety aspects and others. This work presents the statistics of the studies related with materials corrosion in its wide variety and specific mechanisms. These exist at world level in the PWR, BWR, and WWER reactors, analysing the AIRS (Advanced Incident Reporting System) during the period between 1993-1998 in the two first plants in during the period between 1982-1995 for the WWER. The factors identification allows characterize them as those which apply, they are what have happen by the presence of some corrosion mechanism. Those which not apply, these are due to incidental by natural factors, mechanical failures and human errors. Finally, the total number of cases analysed, they correspond to the total cases which apply and not apply. (Author)

  8. Measurement in nuclear reactors

    International Nuclear Information System (INIS)

    A nuclear reactor construction has a flux detector comprising a bundle of fibre optics each having a bead incorporating a substance which scintillates on being struck by neutrons or gamma radiations. The other ends of the fibre optics terminate at an image intensifier. The optical fibres may be of glass made from a mixture of silica, alkaline earth metal oxide, cerous oxide and alkali metal oxide. The beads may be incorporated in a disc forming a detector head, which is in a protective guide tube, through which an inert gas may be passed. (author)

  9. Aportación a los Cálculos Neutrónicos y Termohidráulicos en 3D con Códigos de Mejor EStimación. Aplicación a Transitorios en Reactores Nucleares BWR y PWR

    OpenAIRE

    SÁNCHEZ HERNÁNDEZ, ANA MARÍA

    2012-01-01

    El uso de códigos es una herramienta fundamental en Seguridad Nuclear para la simulación de diferentes situaciones en reactores de potencia. En particular, los códigos termohidráulicos de estimación óptima nos permiten simular de forma más realista los fenómenos que suceden en una central nuclear con la representación del circuito primario. A su vez los códigos neutrónicos de dinámica del núcleo, nos permiten una definición y simulación más precisa del núcleo. El uso de los códigos acopl...

  10. en los pivotes centrales

    Directory of Open Access Journals (Sweden)

    Reynaldo Roque Rodés

    2005-01-01

    Full Text Available Se hace una revisión y comentarios sobre las ventajas y limitantes del empleo del LEPA (Low Enery Precision Aplication en los sistemas de riego de pivotes centrales. Estos sistemas o filosofía de manejo del agua para condiciones de escasez o mala calidad del líquido es una alternativa viable para la producción de alimentos. Introducida en la década del 80 en las planicies del sur de Texas, donde la alta evaporación del agua y la necesidad de regar grandes áreas con pivotes centrales obligaba a la búsqueda de una alternativa para incrementar al máximo la eficiencia de aplicación del riego. Aún en fase de estudio e introducción en Cuba para áreas específicas, puede ser una solución de incremento de los rendimientos de los cultivos, empleando menos agua y aguas con calidad limitada

  11. Water Cooled FBNR Nuclear Reactor

    International Nuclear Information System (INIS)

    A new era of nuclear energy is emerging through innovative nuclear reactors that are to satisfy the new philosophies and criteria that are developed by the INPRO program of the International Atomic Energy Agency (IAEA). The IAEA is establishing a new paradigm in relation to nuclear energy. The future reactors should meet the new standards in respect to safety, economy, non-proliferation, nuclear waste, and environmental impact. The Fixed Bed Nuclear Reactor (FBNR) is a small (70 MWe) nuclear reactor that meets all the established requirements. It is an inherently safe and passively cooled reactor that is fool proof against nuclear proliferation. It is simple in design and economic. It can serve as a dual purpose plant to produce simultaneously both electricity and desalinated water thus making it especially suitable to the needs of most of developing countries. FBNR is developed with the support of the IAEA under its program of Small Reactors Without On-Site Refuelling (SRWOSR). The FBNR reactor uses the pressurized water reactor (PWR) technology. It fulfills the objectives of design simplicity, inherent and passive safety, economy, standardization, shop fabrication, easy transportability and high availability. The inherent safety characteristic of the reactor dispenses with the need for containment; however, a simple underground containment is envisaged for the reactor in order to reduce any adverse visual impact. (author)

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

    International Nuclear Information System (INIS)

    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)

  13. Virtual nuclear reactor for education of nuclear reactor physics

    International Nuclear Information System (INIS)

    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)

  14. El pretensado y su vigilancia en las centrales nucleares españolas : Descripción y análisis histórico de los resultados

    OpenAIRE

    Codinach Fitó, Marc

    2008-01-01

    La utilización del pretensado en las centrales nucleares es un tema singular que difiere de la utilización que se le suele dar a un sistema de armadura activa convencional. En esta tesina se hace un estado del conocimiento de esta tecnología situándolo en el contexto de las centrales nucleares españolas. Además se analizan los datos obtenidos de las vigilancias realizadas hasta hoy en los diferentes edificios de contención de las centrales nucleares españolas y se extraen la...

  15. Nuclear reactor with control rods

    International Nuclear Information System (INIS)

    The invention relates to liquid cooled nuclear reactors. In particular, it concerns reactors with mobile control rods in a straight line and guide tubes to guide these control rods through the internal upper components of the reactor vessel and in the aligned fuel assemblies of the core

  16. Nuclear reaction data and nuclear reactors

    International Nuclear Information System (INIS)

    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

  17. Análisis Integrado de Seguridad de un accidente de SGTR en un reactor nuclear tipo PWR

    OpenAIRE

    Jimenez Varas, Gonzalo

    2012-01-01

    El accidente de rotura de tubos de un generador de vapor (Steam Generator Tube Rupture, SGTR) en los reactores de agua a presión es uno de los transitorios más exigentes desde el punto de vista de operación. Los transitorios de SGTR son especiales, ya que podría dar lugar a emisiones radiológicas al exterior sin necesidad de daño en el núcleo previo o sin que falle la contención, ya que los SG pueden constituir una vía directa desde el reactor al medio ambiente en este transitorio. En los aná...

  18. Optimal control of nuclear reactors

    International Nuclear Information System (INIS)

    The modern control theory is applied to the design of control systems for experimental nuclear reactors that do not belong to power reactors, the component forms of optimal control systems for nuclear reactors are demonstrated. The adoption of output quadratic integral criterion and incomplete state feedback technique can make these systems both efficient and economical. Moreover, approximate handling methods are given so as to simplify the calculations in design. In addition, the adoptable reference values of parameters are given in the illustration

  19. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    The description is given of a nuclear reactor fuel assembly comprising fuel elements arranged in a supporting frame composed of two end pieces, one at the top and the other at the bottom, on which are secured the ends of a number of vertical tubes, each end piece comprising a plane bottom on which two series of holes are made for holding the tubes and for the passage of the coolant. According to the invention, the bottom of each end piece is fixed to an internal plate fitted with the same series of holes for holding the tubes and for the fluid to pass through. These holes are of oblong section and are fitted with fixing elements cooperating with corresponding elements for securing these tubes by transversal movement of the inside plate

  20. Nuclear reactor inspection device

    International Nuclear Information System (INIS)

    A typical embodiment of the invention combines a novel cellular end fitting for a nuclear reactor fuel assembly with a new design for a fuel rod end cap and a radiation sensing device probe to provide a means for swiftly and accurately distinguishing sound fuel rods from those rods that have developed leaks. For example, a somewhat thinner than usual fuel rod end cap is accessible through the open cellular structure of the end fitting to permit a hollow metal probe to contact the fuel rod end cap. This direct contact excludes most of the water, metal and other shielding materials from the volume between the interior of the fuel rod and the radiation detector, thereby improving the quality of the fuel rod examination. A bridge and trolley structure for accurately positioning the probe also is described

  1. Nuclear reactor spacer assembly

    International Nuclear Information System (INIS)

    A fuel assembly for a nuclear reactor is disclosed wherein the fuel element receiving and supporting grid is comprised of a first metal, the guide tubes which pass through the grid assembly are comprised of a second metal and the grid is supported on the guide tubes by means of expanded sleeves located intermediate the grid and guide tubes. The fuel assembly is fabricated by inserting the sleeves, of initial outer diameter commensurate with the guide tube outer diameters, through the holes in the grid assembly provided for the guide tubes and thereafter expanding the sleeves radially outwardly along their entire length such that the guide tubes can subsequently be passed through the sleeves. The step of radial expansion, as a result of windows provided in the sleeves having dimensions commensurate with the geometry of the grid, mechanically captures the grid and simultaneously preloads the sleeve against the grid whereby relative motion between the grid and guide tube will be precluded

  2. Nuclear reactor measurement system

    International Nuclear Information System (INIS)

    An instrument to detect the temperature and flow-rate of the liquid metal current of a coolant fluid sample from adjacent sub-assemblies of a liquid metal-cooled nuclear reactor is described. It includes three thermocouple hot junctions mounted in series, each intended for exposure to a sample-current from a single sub-assembly, electromagnetic coils being mounted around an induction core which detects variations in the liquid metal flow-rate by deformation of the lines of flux. The instrument may also include a thermocouple to detect the mean temperature of the sample-current of coolant fluid from several sources, the result being that the temperature of the coolant fluid current in a sub-assembly may be inferred from the three temperature readings associated with this sub-assembly

  3. Internalization of externalities in the generation costs of electric power centrals of carbon, combined cycle and nuclear; Internalizacion de externalidades en los costos de generacion de centrales electricas de carbon, ciclo combinado y nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Gomez R, M.C. [Universidad Anahuac del Norte (Mexico); Palacios H, J.; Ramirez S, R.; Alonso V, G. [ININ, Carretera Mexico-Toluca Km. 36.5 Ocoyoacac 52750 Edo. de Mexico (Mexico)]. e-mail: fgrivera@avantel.net

    2007-07-01

    The technologies of electric power generation that use fossil fuels, they incorporate in the Even Total Cost of Generation (CTNG) only the direct costs of generation (investment, fuel costs, operation costs and maintenance). nevertheless, the nuclear energy incorporates besides the direct costs, the externalities that causes to the human health and the environment. In this work the CTNG is calculated that incorporates the externalities, of a thermoelectric power station of coal, a plant of combined cycle and of four reactors of Generation III (ABWR, ACR, AP1000 and EPR). The obtained results show that the nuclear power station has smaller CTNG that the technologies that use fossil fuels. It is important to stand out that they are only considering the externalities of the stage of electricity generation, for what the mining phase and transport of the fuel toward the central are not considered in the present document. (Author)

  4. Aplicación de la metodología ISA al análisis de la secuencia de pérdida del sistema de agua de refrigeración de componentes en una central nuclear con reactor de agua a presión

    OpenAIRE

    Ibáñez Hurtado, Luisa

    2015-01-01

    La metodología Integrated Safety Analysis (ISA), desarrollada en el área de Modelación y Simulación (MOSI) del Consejo de Seguridad Nuclear (CSN), es un método de Análisis Integrado de Seguridad que está siendo evaluado y analizado mediante diversas aplicaciones impulsadas por el CSN; el análisis integrado de seguridad, combina las técnicas evolucionadas de los análisis de seguridad al uso: deterministas y probabilistas. Se considera adecuado para sustentar la Regulación Informada por el Ries...

  5. Nuclear energy options for Central Asia

    International Nuclear Information System (INIS)

    Full text: The five countries of Central Asia have a strong basis for the development of commercial nuclear energy. Several test reactors have operated within the region, including the Ak tau BN350 - a very advanced fast breeder reactor combined with a large water desalination plant. The Central Asian countries have a large cadre of well-trained nuclear scientists and engineers who could operate and maintain nuclear power plants and expanded nuclear fuel infrastructure as they evolve. The Central Asia region experiences significant demand for base-load energy in major population centers and industrial development areas. A well-developed electricity transmission grid could transmit nuclear-generated electricity from the power stations to the load centers. Finally, given the large land area and the relatively small population (in relation to the size of the region) there exist many remote and stable sites where nuclear generation centers can be sited and connected to the transmission grid. A good example is the Semipalatinsk Nuclear Test Site (STS) whose vast area could easily contain several nuclear power plants, which would be cooled by the water flow of the Irtish River. The Kazakhstan authorities have already identified several potential nuclear power plant sites within the national transmission system, the STS being one such prospective site. The large-scale availability of uranium in the region affords the uranium exporting countries - particularly Kazakhstan and Uzbekistan - significant leverage with international nuclear reactor vendors in establishing the terms for nuclear plant imports into the region. Such leverage could further be increased if multiple reactor orders are submitted, for instance by two or more countries ordering similar types of plants to be installed at various sites in their territories. The added value of the uranium exports from Central Asia does not have to be measured only in terms of supporting the development of fuel cycle

  6. Sodium-cooled nuclear reactors

    International Nuclear Information System (INIS)

    This book first explains the choice of sodium-cooled reactors by outlining the reasons of the choice of fast neutron reactors (fast neutrons instead of thermal neutrons, recycling opportunity for plutonium, full use of natural uranium, nuclear waste optimization, flexibility of fast neutron reactors in nuclear material management, fast neutron reactors as complements of water-cooled reactors), and by outlining the reasons for the choice of sodium as heat-transfer material. Physical, chemical, and neutron properties of sodium are presented. The second part of the book first presents the main design principles for sodium-cooled fast neutron reactors and their core. The third part proposes an historical overview and an assessment of previously operated sodium-cooled fast neutron reactors (French reactors from Rapsodie to Superphenix, other reactors in the world), and an assessment of the main incidents which occurred in these reactors. It also reports the experience and lessons learned from the dismantling of various sodium-cooled fast breeder reactors in the world. The next chapter addresses safety issues (technical and safety aspects related to the use of sodium) and environmental issues (dosimetry, gaseous and liquid releases, solid wastes, and cooling water). Then, various technological aspects of these reactors are addressed: the energy conversion system, main components, sodium chemistry, sodium-related technology, advances in in-service inspection, materials used in reactors and their behaviour, and fuel system. The next chapter addresses the fuel cycle in these reactors: its integrated specific character, report of the French experience in fast neutron reactor fuel processing, description of the transmutation of minor actinides in these reactors. The last chapter proposes an overview of reactors currently projected or under construction in the world, presents the Astrid project, and gives an assessment of the economy of these reactors. A glossary and an index

  7. Fast reactors and nuclear nonproliferation

    International Nuclear Information System (INIS)

    Problems are discussed with regard to nuclear fuel cycle resistance in fast reactors to nuclear proliferation risk due to the potential for use in military programs of the knowledge, technologies and materials gained from peaceful nuclear power applications. Advantages are addressed for fast reactors in the creation of a more reliable mode of nonproliferation in the closed nuclear fuel cycle in comparison with the existing fully open and partially closed fuel cycles of thermal reactors. Advantages and shortcomings are also discussed from the point of view of nonproliferation from the start with fast reactors using plutonium of thermal reactor spent fuel and enriched uranium fuel to the gradual transition using their own plutonium as fuel. (author)

  8. "Gato encerrado" o "confidencialidad comercial": el problema del secreto en la venta de un reactor nuclear "Gato encerrado" or "confidencialidad comercial": a secret problem regards the sale of a nuclear reactor

    Directory of Open Access Journals (Sweden)

    Ana Spivak L' Hoste

    2006-12-01

    Full Text Available En este artículo se propone abordar uno de los aspectos medulares de la disputa pública que surge a partir de la venta de un reactor nuclear diseñado por una empresa argentina a Australia en el año 2000. Se trata del problema del secreto que atraviesa, para quienes están involucrados en la misma y desde dos perspectivas claramente diferentes, la operación comercial. A través del análisis de las categorías locales que hacen a la configuración del problema del secreto, gato encerrado y secreto comercial, se pretende explorar algunas de las dimensiones sociales, históricas y simbólicas que se articulan en las narrativas que dan vida a esta disputa. Asimismo, se pretende dar cuenta, por ese medio, de qué manera el proceso de construcción de significados es también un proceso de construcción de colectivos sociales (Hecht 1996. Finalmente, se concluirá con una breve reflexión sobre la relevancia analítica y conceptual de la etnografía en el campo de los estudios sociales de la ciencia y la tecnología y, más específicamente, en el ámbito de las disputas públicas que, como ésta, involucran el desarrollo, la producción y el conocimiento científico y tecnológico.This paper deals with one of the many aspects of a science and technology dispute regarding the sale of a nuclear reactor designed by an Argentinean company, INVAP, to Australia in 2000. The analysis is mainly centered on one of the most relevant axis of the conflict: the problem of the secret that passes through a commercial operation. The objective of the article is, on the one hand, to explore some of the social, historical and symbolic dimensions related to the narratives that shaped the dispute through an analysis of the local categories that articulate the secret's problem ("gato encerrado" and "secreto comercial" and to show through this analysis how the process of constructing meanings is also a process of constructing social collectives (Hecht 1996. On the

  9. Fourth Regional Meeting: Nuclear Energy in Central Europe, Proceedings

    International Nuclear Information System (INIS)

    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.

  10. 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)

  11. Nuclear reactor building

    International Nuclear Information System (INIS)

    Purpose: To prevent seismic vibrations of external buildings from transmitting to the side walls of a reactor container in a tank type FBR reactor building. Constitution: The reactor building is structured such that the base mat for a reactor container chamber and a reactor container is separated from the base mat for the walls of building, and gas-tight material such as silicon rubber is filled in the gap therebetween. With such a constitution, even if the crane-supporting wall vibrates violently upon occurrence of earthqualkes, the seismic vibrations do not transmit toward the reactor container chamber. (Horiuchi, T.)

  12. 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)

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

    International Nuclear Information System (INIS)

    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 RELAP5-3DC/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)

  14. The BEPU (Best Estimate Plus Uncertainty) challenge in current licensing of nuclear reactors

    International Nuclear Information System (INIS)

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

  15. Liquid cooled nuclear reactors

    International Nuclear Information System (INIS)

    A construction is described for a liquid metal cooled fast reactor, in which the core is supported in a pool of liquid coolant, wherein a catchment tray is provided for any debris falling from the core. The tray comprises a complex of open top collecting vessels with central support struts, the vessels being spaced apart and arranged in layers in a lattice pitch. The lattice pitches of the vessels in each layer are off-set to the lattice pitches of the vessels in the other layers, so that upper vessels partially overlap lower vessels, and the support struts extend through interspaces defined by the vessels in off-set pitch to a common supporting sub-structure. The complex of vessels offers a complete catchment area for falling debris, whilst being pervious to liquid coolant circulating upwardly by convection. The collecting vessels preferably comprise conical dishes and are arranged in triangular lattice pitch in each layer, and the complex of vessels comprises three layers. Alternatively the collecting vessels may be rectilinear and arranged on a square lattice. The catchment tray may comprise two or more such complexes in stacked array. (U.K.)

  16. Teaching About Nature's Nuclear Reactors

    CERN Document Server

    Herndon, J M

    2005-01-01

    Naturally occurring nuclear reactors existed in uranium deposits on Earth long before Enrico Fermi built the first man-made nuclear reactor beneath Staggs Field in 1942. In the story of their discovery, there are important lessons to be learned about scientific inquiry and scientific discovery. Now, there is evidence to suggest that the Earth's magnetic field and Jupiter's atmospheric turbulence are driven by planetary-scale nuclear reactors. The subject of planetocentric nuclear fission reactors can be a jumping off point for stimulating classroom discussions about the nature and implications of planetary energy sources and about the geomagnetic field. But more importantly, the subject can help to bring into focus the importance of discussing, debating, and challenging current thinking in a variety of areas.

  17. Safety device for nuclear reactor

    International Nuclear Information System (INIS)

    This invention relates to a safety device for a nuclear reactor, particularly a liquid metal (generally sodium) cooled fast reactor. This safety device includes an absorbing element with a support head connected by a disconnectable connector formed by the armature of an electromagnet at the end of an axially mobile vertical control rod. This connection is so designed that in the event of it becoming disconnected, the absorbing element gravity slides in a passage through the reactor core into an open container

  18. Environmental management systems implemented in the Spanish nuclear power plants; Sistemas de gestion ambiental implantados en las centrales nucleares espanolas

    Energy Technology Data Exchange (ETDEWEB)

    Redondo, R.; Fernandez Guisado, M. B.; Hortiguela, R.; Bustamante, L. F.; Esparza, J. L.; Villareal, M.; Yague, F.

    2013-09-01

    The companies that own the Spanish Nuclear Power Plants, aware of social concern and in the context of a growing demanding environmental legislation, have a permanent commitment to the electricity production based on the principles of a maximum respect for the environment, safety, quality, professionalism and continuous improvement. In order to minimize the environmental impact of their plants they have implemented and Environmental Management System based on the ISO 14001 Standard. They minimize the environmental impact by identifying the significant environmental aspects and defining the corresponding objectives. This article describes the referred environmental management systems and their environmental objectives, as applied and defined by the Spanish Nuclear Power Plants. (Author)

  19. Control rod drive of nuclear reactor

    International Nuclear Information System (INIS)

    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)

  20. Modularization in construction processes New Nuclear Power Plants; Modularizacion en procesos de construccion de Nuevas Centrales Nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Martinez, I.; Cobos, A.; Herrera Ropero, D.

    2012-07-01

    The aim of this work is that it has the capacity and expertise to analyze the suitability of modular technology design and construction compared to conventional nuclear plants. It will define the criteria for selecting the areas of modularity and the impact on design and its interfaces with engineering, supply, including logistics and construction.

  1. Nuclear Reactor RA Safety Report, Vol. 11, Reactor operation

    International Nuclear Information System (INIS)

    This volume includes the following chapters describing: Organisation of reactor operation (including operational safety, fuel management, and regulatory rules for RA reactor operation); Control and maintenance of reactor components (reactor core, nuclear fuel, heavy water and cover gas systems, mechanical structures, electric power supply system, reactor instrumentation); Quality assurance and Training of the reactor personnel

  2. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    The Center for Nuclear Engineering has shown expertise in the field of nuclear and energy systems ad correlated areas. Due to the experience obtained over decades in research and technological development at Brazilian Nuclear Program personnel has been trained and started to actively participate in the design of the main system that will compose the Brazilian Multipurpose Reactor (RMB) which will make Brazil self-sufficient in the 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. The Nuclear Fuel Center is responsible for the production of the nuclear fuel necessary for the continuous operation of the IEA-R1 research reactor. Development of new fuel technologies is also a permanent concern

  3. 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

  4. Determination of the energy spectrum of the neutrons in the central thimble of the reactor core TRIGA Mark III; Determinacion del espectro de energia de los neutrones en el dedal central del nucleo del reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Parra M, M. A.

    2014-07-01

    This thesis presents the neutron spectrum measurements inside the core of the TRIGA Mark III reactor at 1 MW power in steady-state, with the bridge placed in the center of the swimming pool, using several metallic threshold foils. The activation detectors are inserted in the Central Thimble of the reactor core, all the foils are irradiated in the same position and irradiation conditions (one by one). The threshold detectors are made of different materials such as: Au{sup 197}, Ni{sup 58}, In{sup 115}, Mg{sup 24}, Al{sup 27}, Fe{sup 58}, Co{sup 59} and Cu{sup 63}, they were selected to cover the full range the energies (10{sup -10} to 20 MeV) of the neutron spectrum in the reactor core. After the irradiation, the activation detectors were measured by means of spectrometry gamma, using a high resolution counting system with a hyper pure Germanium crystal, in order to obtain the saturation activity per target nuclide. The saturation activity is one of the main input data together with the initial spectrum, for the computational code SANDBP (hungarian version of the code SAND-II), which through an iterative adjustment, gives the calculated spectrum. The different saturation activities are necessary for the unfolding method, used by the computational code SANDBP. This research work is very important, since the knowledge of the energetic and spatial distribution of the neutron flux in the irradiation facilities, allows to characterize properly the irradiation facilities, just like, to estimate with a good precision various physics parameters of the reactor such as: neutron fluxes (thermal, intermediate and fast), neutronic dose, neutron activation analysis (NAA), spectral indices (cadmium ratio), buckling, fuel burnup, safety parameters (reactivity, temperature distribution, peak factors). In addition, the knowledge of the already mentioned parameters can give a best use of reactor, optimizing the irradiations requested by the users for their production process or

  5. The future of nuclear reactors

    International Nuclear Information System (INIS)

    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

  6. Studies on nuclear reactor design

    International Nuclear Information System (INIS)

    this thesis presents two studies for safety aspects in nuclear reactor design. the fission process that occurs in the reactor core is the most important process for the harmful effect of produced radiation especially neutrons with different energies and gamma radiations for their strong penetrability . so studying the criticality of the fissile materials in the reactor is one of the most important safety aspects for the reactor design, the attenuation of the neutrons and gammas using suitable shielding materials with suitable thicknesses is the second study that is discussed in this thesis

  7. Gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    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

  8. Technique of nuclear reactors controls

    International Nuclear Information System (INIS)

    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)

  9. Low-power nuclear reactors

    International Nuclear Information System (INIS)

    A brief development history of low-power nuclear reactors is presented in this paper. Nowadays, some countries have plans to build a series of small nuclear power plants (also floating ones) for use in remote regions. Present constructions of such NPP are presented in this paper. (author)

  10. Análisis de la respuesta estructural del edificio de contención de un reactor nuclear PWR frente a una secuencia de Station Blackout

    OpenAIRE

    Martínez Casanovas, Josep Maria

    2013-01-01

    Después del Accidente que sucedió el 11 de marzo de 2011 en la central nuclear de Daichii-Fukushima, han sido muchos los debates reabiertos acerca de la seguridad que ofrecen los reactores nucleares en operación. Debido a las incertidumbres que este accidente ha suscitado, surge ésta tesina, en la que se pretende conocer y comprender técnicamente que sucedería si un accidente similar al de Fukushima ocurriese en uno de los reactores de nuestro país. Con el principal objetivo de encontrar res...

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

    International Nuclear Information System (INIS)

    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. TRACRT 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. TRACRT 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, TRACRT 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)

  12. Nuclear reactors and fuel cycle

    International Nuclear Information System (INIS)

    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 100th nuclear fuel element produced. Besides routine production, development of new technologies is also a permanent concern at CCN. In 2005, U3O8 were replaced by U3Si2-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 to

  13. 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

  14. nuclear reactor design calculations

    International Nuclear Information System (INIS)

    In this work , the sensitivity of different reactor calculation methods, and the effect of different assumptions and/or approximation are evaluated . A new concept named error map is developed to determine the relative importance of different factors affecting the accuracy of calculations. To achieve this goal a generalized, multigroup, multi dimension code UAR-DEPLETION is developed to calculate the spatial distribution of neutron flux, effective multiplication factor and the spatial composition of a reactor core for a period of time and for specified reactor operating conditions. The code also investigates the fuel management strategies and policies for the entire fuel cycle to meet the constraints of material and operating limitations

  15. Nuclear reactor internal structures

    International Nuclear Information System (INIS)

    The upper internal structures of the reactor are connected to the closing head so as to be readily removed with the latter and a skirt connected to the lower portion of said upper structures so as to surround the latter, extends under the control rods when they are removed from the reactor core. Through such an arrangement the skirt protects the control rods and supports the vessel closing-head and the core upper structures, whenever the head is severed from the vessel and put beside the latter in order to discharge the reactor

  16. Third Regional Meeting: Nuclear Energy in Central Europe, Proceedings

    International Nuclear Information System (INIS)

    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

  17. Questions to the reactors power upgrade of the Nuclear Power Plant of Laguna Verde; Cuestionamientos al aumento de potencia de los reactores de la Central Nuclear de Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Salas M, B., E-mail: salasmarb@yahoo.com.mx [UNAM, Facultad de Ciencias, Departamento de Fisica, Circuito exterior s/n, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2014-08-15

    The two reactors of the Nuclear Power Plant of Laguna Verde (NPP-L V) were subjected to power upgrade labors with the purpose of achieving 20% upgrade on the original power; these labors concluded in August 24, 2010 for the Reactor 1 and in January 16, 2011 for the Reactor 2, however in January of 2014, the NNP-L V has not received by part of the Comision Nacional de Seguridad Nuclear y Salvaguardias (CNSNS) the new Operation License to be able to work with the new power, because it does not fulfill all the necessary requirements of safety. In this work is presented and analyzed the information obtained in this respect, with data provided by the Instituto Federal de Acceso a la Informacion Publica y Proteccion de Datos (IFAI) and the Comision Federal de Electricidad (CFE) in Mexico, as well as the opinion of some workers of the NPP-L V. The Governing Board of the CFE announcement that will give special continuation to the behavior on the operation and reliability of the NPP-L V, because the frequency of not announced interruptions was increased 7 times more in the last three years. (Author)

  18. GE's advanced nuclear reactor designs

    International Nuclear Information System (INIS)

    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

  19. Nuclear reactor fuel assembly

    International Nuclear Information System (INIS)

    A fuel assembly construction for liquid metal cooled fast breeder reactors is described in which the sub-assemblies carry a smaller proportion of parasitic material than do conventional sub-assemblies. (U.K.)

  20. Nuclear reactor (1960)

    International Nuclear Information System (INIS)

    The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author)

  1. BWR type nuclear reactor

    International Nuclear Information System (INIS)

    Purpose: To simplify the structure of an emergency core cooling system while suppressing the flow out of coolants upon rapture accidents in a coolant recycling device of BWR type reactors. Constitution: Recirculation pumps are located at a position higher than the reactor core in a pressure vessel, and the lower plenum is bisected vertically by a partition plate. Further, a gas-liquid separator is surrounded with a wall and the water level at the outer side of the wall is made higher than the water level in the inside of the wall. In this structure, coolants are introduced from the upper chamber in the lower plenum into the reactor core, and the steams generated in the reactor core are separated in the gas-liquid separator, whereby the separated liquid is introduced as coolants by way of the inner chamber into the lower chamber of the lower plenum and further sent by way of the outer chamber into the reactor core. Consequently, idle rotation of the recycling pumps due to the flow-in of saturated water is prevented and loss of coolants in the reactor core can also be prevented upon raptures in the pipeway and the driving section of the pump connected to the pressure vessel and in the bottom of the pressure vessel. (Horiuchi, T.)

  2. Temporary storage in dry of the spent nuclear fuel in the Nuclear Power Plant of Laguna Verde; Almacenamiento temporal en seco del combustible nuclear gastado en la Central Nuclear Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez M, N.; Vargas A, A., E-mail: natividad.hernandez@cfe.gob.mx [Comision Federal de Electricidad, Gerencia de Centrales Nucleoelectricas, Carretera Veracruz-Medellin Km. 7.5, 94270 Dos Bocas, Veracruz (Mexico)

    2013-10-15

    To guarantee the continuity in the operation of the two nuclear reactors of the nuclear power plant of Laguna Verde (NPP-L V) is an activity of high priority of the Comision Federal de Electricidad (CFE) in Mexico. At the present time, the CFE is working in the storage project in dry of the spent fuel with the purpose of to liberate space of the pools and to have the enlarged capacity of storage of the spent fuel that is discharged of the reactors. This work presents the storage option in dry of the spent fuel, considering that the original capacity of the spent fuel pools of the NPP-L V was of 1242 spaces each one and that in 1991, through a modification of the original design, the storage capacity was increased to 3177 spaces by pool. At present, the cells occupied by unit are of 2165 (68%) for the Unit-I and 1839 (58%) for the Unit-2, however, in 2017 and 2022 the capacity to discharge the complete core will be limited by what is required of a retirement option of spent fuel assemblies to liberate spaces. (author)

  3. Fixed bed nuclear reactor concept

    International Nuclear Information System (INIS)

    Full text: The fixed bed nuclear reactor (FBNR) is essentially a pressurized light water reactor (PWR) having spherical fuel elements constituting a suspended reactor core at its lowest bed porosity. The core is movable thus under any adverse condition, the fuel elements can leave the reactor core naturally through the force of gravity and fall into the passively cooled fuel chamber or leave the reactor all together entering the spent fuel pool. It is a small and modular reactor being simple in design. Its spent fuel is in such a convenient form and size that may be utilized directly as the source for irradiation and applications in agriculture and industry. This feature results in a positive impact on waste management and environmental protection. The principle features of the proposed reactor are that the concept is polyvalent, simple in design, may operate either as fixed or fluidized bed, have the core suspended contributing to inherent safety, passive cooling features of the reactor. The reactor is modular and has integrated primary system utilizing either water, supercritical steam or helium gas as its coolant. Some of the advantages of the proposed reactor are being modular, low environmental impact, exclusion of severe accidents, short construction period, flexible adaptation to demand, excellent load following characteristics, and competitive economics. The characteristics of the Fluidized Bed Nuclear Reactor (FBNR) concept may be analyzed under the light of the requirements set for the IV generation nuclear reactors. It is shown that FBNR meet the goals of (1) Providing sustainable energy generation that meets clean air objectives and promotes long-term availability of systems and effective fuel utilization for worldwide energy production, (2) Minimize and manage their nuclear waste and notably reduce the long term stewardship burden in the future, thereby improving protection for the public health and the environment, (3) Excel in safety and reliability

  4. Research reactors in Argentina and their neutrons applications

    International Nuclear Information System (INIS)

    Argentina has been successfully developing nuclear energy applications for many years. In particular neutron applications have been developed and used in its research reactors and accelerators. The Argentinean Atomic Energy Commission (CNEA) has six Research Reactors, three of them are Critical Assemblies. The main objectives of these Research Reactors in CNEA's strategic plan are: - Personnel education and training for building, start-up, operation and maintenance of Atucha II; - Support for NPP life extension (Embalse, Atucha I); - Neutronic parameters measurements as support of CAREM design project; - 90% production of national radioisotopes utilization for medical purposes; - Neutrongraphy and prompt gamma capture techniques analysis; - Different techniques of BNCT applied to humans and animals; - under and post graduate students education; - training for national and international operators; - Public diffusion about the benefits of nuclear applications; - neutron activation analysis for aero spatial and nuclear industry; - irradiation of national developed and constructed silicide fuel for qualification to export; - RR and NPP instrumentation qualification. (author)

  5. Innovative designs of nuclear reactors

    International Nuclear Information System (INIS)

    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 ∇e2 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

  6. Neutrino physics with nuclear reactors

    International Nuclear Information System (INIS)

    This is a lecture given at the Gif Summer School held in 1992 in Montpellier. It contains three chapters. These are devoted to neutrino oscillations, to the nuclear reactors used as neutrino sources, and to the experiments performed with neutrinos from nuclear reactors, respectively. The first chapter offers a theoretical frame, the second discusses the investigation capabilities of nuclear reactors as neutrino sources while the last one describes the experimental aspects. These aspects are related to the neutrino flux measurement and the flavor oscillation, the search for neutrino oscillation, the neutrino scattering on electrons, the neutrino decay, the coherent neutrino scattering on nuclei and the electron neutrino-electron antineutrino oscillations implied by the Majorana nature of neutrinos. In concluding the author points to the possible ways of refining these extremely subtle experiments, which will be approached in the near future. 117 refs., 9 figs., 11 tabs

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

    International Nuclear Information System (INIS)

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

  8. AREVA's nuclear reactors portfolio

    International Nuclear Information System (INIS)

    A reasonable assumption for the estimated new build market for the next 25 years is over 340 GWe net. The number of prospect countries is growing almost each day. To address this new build market, AREVA is developing a comprehensive portfolio of reactors intended to meet a wide range of power requirements and of technology choices. The EPR reactor is the flagship of the fleet. Intended for large power requirements, the four first EPRs are being built in Finland, France and China. Other countries and customers are in view, citing just two examples: the Usa where the U.S. EPR has been selected as the technology of choice by several U.S utilities; and the United Kingdom where the Generic Design Acceptance process of the EPR design submitted by AREVA and EDF is well under way, and where there is a strong will to have a plant on line in 2017. For medium power ranges, the AREVA portfolio includes a boiling water reactor and a pressurized water reactor which both offer all of the advantages of an advanced plant design, with excellent safety performance and competitive power generation cost: -) KERENA (1250+ MWe), developed in collaboration with several European utilities, and in particular with Eon; -) ATMEA 1 (1100+ MWe), a 3-loop evolutionary PWR which is being developed by AREVA and Mitsubishi. AREVA is also preparing the future and is deeply involved into Gen IV concepts. It has developed the ANTARES modular HTR reactor (pre-conceptual design completed) and is building upon its vast Sodium Fast Reactor experience to take part into the development of the next prototype. (author)

  9. Nuclear reactor downcomer flow deflector

    Science.gov (United States)

    Gilmore, Charles B.; Altman, David A.; Singleton, Norman R.

    2011-02-15

    A nuclear reactor having a coolant flow deflector secured to a reactor core barrel in line with a coolant inlet nozzle. The flow deflector redirects incoming coolant down an annulus between the core barrel and the reactor vessel. The deflector has a main body with a front side facing the fluid inlet nozzle and a rear side facing the core barrel. The rear side of the main body has at least one protrusion secured to the core barrel so that a gap exists between the rear side of the main body adjacent the protrusion and the core barrel. Preferably, the protrusion is a relief that circumscribes the rear side of the main body.

  10. 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)

  11. Nuclear Power Reactor Wastes. A Regulatory Overview of the Current Situation in Argentina

    International Nuclear Information System (INIS)

    Since the middle of the 1980's up to the year 2005 the Argentinean nuclear power program was stagnated. Entering in 21st century the situation of the nuclear program had an inflection point as a consequence of the rise of the oil prices, the deep reduction of indigenous resources of natural gas and the increase of electricity demand that resulted in the growing of economical activity since 2001. At present the construction of Atucha II, the 3rd nuclear power reactor, has been restarted and negotiations for a 4th nuclear power reactor are in course. The only authorized site for final disposal of low level solid radioactive waste (AGE) suspended its operation at the end of the '90s and a new schedule for decision and commissioning of radioactive waste facilities were established by National Commission of Atomic Energy (CNEA). Nowadays, in the country, the radioactive waste (RW) from NPPs are stored in the same site where they are generated and it is not planned to have in operation a disposal facility before the second half of the next decade. So, the main technical regulatory concern during this period will be to assure that the storage of the radioactive waste is safe, the characterization of RW performed by Operators is appropriate and a reasonable recording system is implemented. (authors)

  12. Nuclear power reactor wastes. A regulatory overview of the current situation in Argentina

    International Nuclear Information System (INIS)

    Since the middle of the 1980's up to the year 2005 the Argentinean nuclear power program was stagnated. Entering in 21st century the situation of the nuclear program had an inflection point as a consequence of the rise of the oil prices, the deep reduction of indigenous resources of natural gas and the increase of electricity demand that resulted in the growing of economical activity since 2001. At present the construction of Atucha II, the 3rd nuclear power reactor, has been restarted and negotiations for a 4th nuclear power reactor are in course. The only authorized site for final disposal of low level solid radioactive waste (AGE) suspended its operation at the end of the '90s and a new schedule for decision and commissioning of radioactive waste facilities were established by National Commission of Atomic Energy (CNEA). Nowadays, in the country, the radioactive waste (RW) from NPPs are stored in the same site where they are generated and it is not planned to have in operation a disposal facility before the second half of the next decade. So, the main technical regulatory concern during this period will be to assure that the storage of the radioactive waste is safe, the characterization of RW performed by Operators is appropriate and a reasonable recording system is implemented. (author)

  13. Determination of the neutrons energy spectrum in the central thimble of the reactor core TRIGA Mark III; Determinacion del espectro de energia de los neutrones en el dedal central del nucleo del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Parra M, M. A.; Luis L, M. A. [Universidad Autonoma Metropolitana, Unidad Azcapotzalco, Division de Ciencias Basicas, Av. San Pablo No. 180, Col. Reynosa Tamaulipas, 02200 Mexico D. F. (Mexico); Raya A, R.; Cruz G, H. S., E-mail: roberto.raya@inin.gob.mx [ININ, Departamento del Reactor, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    This work presents the measurement of the neutrons spectrum in energies in the central thimble of the reactor TRIGA Mark III to a power of 1 MW in stationary state, with the core in the center of the pool. To achieve this objective, several thin sheets were irradiated (one at the time) in the same position of the core. The activation probes were selected in such a way that covered the energy range (1 x 10{sup -10} to 20 MeV) of the neutrons spectrum in the reactor core, for this purpose thin sheets were used of {sup 197}Au, {sup 58}Ni, {sup 115}In, {sup 24}Mg, {sup 27}Al, {sup 58}Fe, {sup 59}Co and {sup 63}Cu. After the irradiation, the high energy gamma emissions of the activated thin sheets were measured by means of gamma spectrometry, in a counting system of high resolution, with a Hyper pure Germanium detector, obtaining this way the activity induced in the thin sheets whose magnitude is proportional to the intensity of the neutrons flow, this activity together to a theoretical initial spectrum are the main entrance data of the computational code SANDBP (Hungarian version of the code Sand-II) that uses the unfolding method for the calculation of the spectrum. (Author)

  14. Nuclear reactor container

    International Nuclear Information System (INIS)

    In a container of a BWR type reactor, spray water is stored in a pedestal cavity. A perforated hole is formed on the side wall of the pedestal, and a stirrer is disposed in the pedestal cavity to stir the stored spray water. During reactor operation, the door on the side wall of the pedestal is closed to prevent discharge of fission products to the dry well when a severe accident should occur. During periodical inspection for the plant, the door is opened to enable an operator to access to the inside of the pedestal. When a molten reactor core should drop to the pedestal cavity, fission products generated from the failed reactor core left in a pressure vessel pass through the spray water in the pedestal cavity. Then, most of the fission products are held in the spray water by a scrubbing effect when they pass through the spray water. In addition, the stored spray water is stirred by the stirrer to enhance the scrubbing effect thereby enabling to further decrease the amount of the fission products discharged to the dry well. (N.H.)

  15. Nuclear reactor constructions

    International Nuclear Information System (INIS)

    A method of constructing a radiation shielding plug for use in the roof of the coolant containment vault of liquid metal cooled fast breeder reactors is described. The construction allows relative movement of that part of service cables and pipes which are carried by the fixed roof and that part which is carried by the rotatable plug. (U.K.)

  16. Tritium recovery as waste sub product in the Fluorine 18 production in a nuclear reactor; Recuperacion de tritio como subproducto de desecho en la produccion de F-18 en un reactor nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Flores R, H.; Palma G, F.A.; Ramirez, F.M

    1990-09-15

    The tritium is a radioisotope that can be used to carry out basic as applied research. The current researches on the labelling of the organic molecules as well as its application in diagnostic, radiotherapy and hydrology among others confirm the before said. Due to their utility, they have been carried out studies to recover it of radioactive or nuclear waste as well as, to concentrate it of the natural water, the one which due to the nuclear tests in the last decades has gotten rich in tritium. In this work previous studies to recover the tritium coming from the process that was used to produce F-18 following the reaction {sup 6} Li (n, {alpha}) {sup 3} H, {sup 16} O (t, n) {sup 18} F in made up of lithium oxygenated, in the TRIGA Mark III Nuclear Reactor of the Nuclear Center of Mexico. The method consists on purifying by ion exchange the waste solutions where F-18 took place, to distill them and to concentrate them for an electrochemical method. It was already adapts a system reported to concentrate big volumes (approximately 250 ml) in such a way that could be used for small volumes. It was recovered 30% of the considered initial quantity of tritium. A modification to the proposed methodology will allow to recover the waste tritium in a percentage greater to 80%. (Author)

  17. 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

  18. Anàlisis de contenció en un accident termohidràulic d'una central nuclear

    OpenAIRE

    Zamora Povea, Imanol

    2013-01-01

    Aquest projecte és el resultat de la cooperació entre el món universitari i l’empresa privada, doncs ha requerit de la cooperació de la Secció d'Enginyeria Nuclear de la Universitat Politècnica de Catalunya i l’àrea Nuclear Services de l’enginyeria IDOM S.A. El treball desenvolupat ha tingut dues vessants diferenciades. La primera línia de treball ha sigut l’elaboració d’un model de la contenció de la central nuclear Vandellòs II mitjançant el codi GOTHIC v7.2a. Així mateix, s’ha tractat d...

  19. Nuclear reactors and disarmament

    International Nuclear Information System (INIS)

    From a brief analysis of the perspectives of nuclear weapons arsenals reduction, a rational use of the energetic potential of the ogives and the authentic destruction of its warlike power is proposed. The fissionable material conversion contained in the nuclear fuel ogives for peaceful uses should be part of the disarmament agreements. This paper pretends to give an approximate idea on the resources re assignation implicancies. (Author)

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

    International Nuclear Information System (INIS)

    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)

  1. Nuclear reactor container

    International Nuclear Information System (INIS)

    Upon reactor accident, hydrogen and oxygen are generated by water-zirconium reaction and radiolysis of water, which are accumulated in the reactor. If the concentration of hydrogen and oxygen exceeds a burning limit, there is a possibility of hydrogen burning to cause a danger of deteriorating the integrity of the reactor container and the equipments therein. The limit for the occurrence of the detonation is determined by a relationship between the scale of a detonation cell and the size of the container, and if the scale is greater than the container, the detonation does not occur. The scale of the cell is determined by a gas combustion rate and, if the combustion reaction is suppressed, detonation does not occur even in a large container. Then, an appropriate diluent is added to increase heat capacity of a gas mixture to thereby suppress the temperature elevation of the gas. Incombustible gases having a great heat capacity are preferred for the diluent, and CO2 is used. As the concentration of the CO2 gas to be added is increased, the detonation cell is made greater. Thus, occurrence of detonation due to combustion of the accumulated hydrogen can be prevented. (N.H.)

  2. Nuclear safety. Concerns about the nuclear power reactors in Cuba

    International Nuclear Information System (INIS)

    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

  3. Light-water nuclear reactors

    International Nuclear Information System (INIS)

    This work gives basic information on light-water reactors which is advanced enough for the reader to become familiar with the essential objectives and aspects of their design, their operation and their insertion in the industrial, economic and human environment. In view of the capital role of electric energy in the modern economy a significant place is given to electron-nuclear power stations, particularly those of the type adopted for the French programme. The work includes sixteen chapters. The first chapter relates the history and presents the various applications of light water reactors. The second refers to the general elementary knowledge of reactor physics. The third chapter deals with the high power light-water nuclear power station and thereby introduces the ensuing chapters which, up to and including chapter 13, are devoted to the components and the various aspects of the operation of power stations, in particular safety and the relationship with the environment. Chapter 14 provides information on the reactors adapted to applications other than the generation of electricity on an industrial scale. Chapter 15 shows the extent of the industrial effort devoted to light-water reactors and chapter 16 indicates the paths along which the present work is preparing the future of these reactors. The various chapters have been written to allow for separate consultation. An index of the main technical terms and a bibliography complete the work

  4. Nuclear reactor PBMR and cogeneration

    International Nuclear Information System (INIS)

    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)

  5. Treatment of operational experience of nuclear power plants in WANO; Tratamiento de la experiencia operativa de las centrales nucleares en WANO

    Energy Technology Data Exchange (ETDEWEB)

    Ibanez, M.

    2013-09-01

    The article describes the activities associated to the Operating Experience Programme of the World Association of Nuclear Operators. The programme manages the event reports submitted by the nuclear power plants to the WANO database for the preparation by the Operating Experience Central Team of some documents like the significant Operating Experience Reports and Significant Event Reports that help the stations to avoid similar events. (Author)

  6. Health requirements for nuclear reactor operators

    International Nuclear Information System (INIS)

    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.)

  7. Gaseous fuel nuclear reactor research

    Science.gov (United States)

    Schwenk, F. C.; Thom, K.

    1975-01-01

    Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

  8. Instrumentation for nuclear reactor control

    International Nuclear Information System (INIS)

    This lecture is concerned with engineers and technicians not specialized in nuclear reactor control. The different methods of measurement used are briefly reviewed: current or pulse measurement, and Campbell system; the electronic networks are described and a part is devoted to the cables connecting detectors and electronic assemblies

  9. 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.

  10. Nuclear reactor effluent monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Minns, J.L.; Essig, T.H. [Nuclear Regulatory Commission, Washington, DC (United States)

    1993-12-31

    Radiological environmental monitoring and effluent monitoring at nuclear power plants is important both for normal operations, as well as in the event of an accident. During normal operations, environmental monitoring verifies the effectiveness of in-plant measures for controlling the release of radioactive materials in the plant. Following an accident, it would be an additional mechanism for estimating doses to members of the general public. This paper identifies the U.S. Nuclear Regulatory Commission (NRC) regulatory basis for requiring radiological environmental and effluent monitoring, licensee conditions for effluent and environmental monitoring, NRC independent oversight activities, and NRC`s program results.

  11. Water cooled nuclear reactor

    International Nuclear Information System (INIS)

    In the reactor operating with supercritical pressure and temperature part of the water flowing through the moderator tubes is deflected at the outlet and mixed with a residual partial flow of the coolant fed into the core as well as passed along the fuel rods in opposite direction. By special guiding of the flow downward through the guide tubes of the control rods insertion of the control rods is simplified because of reduced frictional forces. By this means it is also achieved to design less critical the control rod cooling with respect to flow rate control and operating behavior in case of a scram. (orig.)

  12. Nuclear reactor plant

    International Nuclear Information System (INIS)

    The plant consists mainly of a steam-raising unit and a steam turbine with high pressure and low pressure stages. There is at least one superheater or intermediate superheater between the steam-raising unit and the low pressure stage. In order to improve the plant efficiency, a high temperature reactor is provided as a source of heat for the superheater or intermediate superheater, which supplies the superheater heat with an energy efficiency of over 60%. This increases the overall net efficiency from 33% to over 36%. (orig.)

  13. Decommissioning of Salaspils nuclear reactor

    International Nuclear Information System (INIS)

    In May 1995, the Latvian Government decided to shut down the Research Reactor Salaspils (SRR) and to dispense with nuclear energy in future. The reactor has been out of operation since July 1998. A conceptual study for the decommissioning of SRR has been carried out by Noell-KRC-Energie- und Umwelttechnik GmbH from 1998-1999. he Latvian Government decided on 26 October 1999 to start the direct dismantling to 'green field' in 2001. The results of decommissioning and dismantling performed in 1999-2001 are presented and discussed. The main efforts were devoted to collecting and conditioning 'historical' radioactive waste from different storages outside and inside the reactor hall. All radioactive material more than 20 tons were conditioned in concrete containers for disposal in the radioactive waste depository 'Radons' in the Baldone site. Personal protective and radiation measurement equipment was upgraded significantly. All non-radioactive equipment and material outside the reactor buildings were free-released and dismantled for reuse or conventional disposal. Weakly contaminated material from the reactor hall was collected and removed for free-release measurements. The technology of dismantling of the reactor's systems, i.e. second cooling circuit, zero power reactors and equipment, is discussed in the paper. (author)

  14. Rol del fallo mecánico en la optimización del mantenimiento en una central nuclear//Role of the mechanical failure during the maintenance optimization in the nuclear power plant

    Directory of Open Access Journals (Sweden)

    Antonio Torres-Valle

    2012-05-01

    Full Text Available Entre las más recientes aplicaciones del Análisis Probabilista de Seguridad (1997 – 2003 de la Central Nuclear Embalse en Argentina, está el Programa de Mantenimiento Orientado a la Seguridad (2006– 2009 el cual se ha desarrollado con el empleo de la metodología de Mantenimiento Centrado en la Confiabilidad (RCM en inglés. El objetivo general del artículo es demostrar la alta contribución de los fallos mecánicos en el diseño de las políticas de mantenimiento de varios sistemas de la instalación a través del empleo de la metodología RCM. La composición, estructura y políticas de explotación de los sistemas tecnológicos de muchas instalaciones con riesgo asociado, similares a las de los sistemas analizados en este estudio, permite inferir que los resultados que se obtendrán serán equivalentes de aplicarse la metodología RCM en dichas instalaciones. Palabras claves: mantenimiento centrado en la confiabilidad, mantenimiento predictivo, mantenimiento preventivo, fallo mecánico, seguridad, confiabilidad, riesgo.______________________________________________________________________________ Abstract One of the most recent applications of Probabilistic Safety Analysis (1997 – 2003 to Embalse Nuclear Power Plant in Argentina, is the Safety Oriented Maintenance Program (2006 – 2009 developed with employment of the Reliability Centered Maintenance (RCM methodology. The general objective of the paper is to demonstrate the high contribution of the mechanical failures in the maintenance program design through the RCM methodology. The composition, structure and operation strategies of the technological systems of many risk associated facilities, similar to the analysed systems included in this study, allow deduce that the results will equivalent in case of application of RCM methodology in such facilities. Key words: reliability centered maintenance (RCM, predictive maintenance, preventive maintenance, mechanical failure, safety

  15. Application of COMSOL in the solution of the neutron diffusion equations for fast nuclear reactors in stationary state; Aplicacion de COMSOL en la solucion de las ecuaciones de difusion de neutrones para reactores nucleares rapidos en estado estacionario

    Energy Technology Data Exchange (ETDEWEB)

    Silva A, L.; Del Valle G, E., E-mail: evalle@ipn.mx [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN s/n, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico)

    2012-10-15

    This work shows an application of the program COMSOL Multi physics Ver. 4.2a in the solution of the neutron diffusion equations for several energy groups in nuclear reactors whose core is formed by assemblies of hexagonal transversal cut as is the cas of fast reactors. A reference problem of 4 energy groups is described of which takes the cross sections which are processed by means of a program that prepares the definition of the constants utilized in COMSOL for the generic partial differential equations that this uses. The considered solution domain is the sixth part of the core which is applied frontier conditions of reflection and incoming flux zero. The discretization mesh is elaborated in automatic way by COMSOL and the solution method is one of finite elements of Lagrange grade two. The reference problem is known as the Knk with and without control rod which led to propose the calculation of the effective multiplication factor in function of the control rod fraction from a value 0 (completely inserted control rod) until the value 1 (completely extracted control rod). Besides this the reactivity was determined as well as the change of this in function of control rod fraction. The neutrons scalar flux for each energy group with and without control rod is proportioned. The reported results show a behavior similar to the one reported in other works but using the discreet ordinates S{sub 2} approximation. (Author)

  16. Nuclear fuel supply view in Argentina

    International Nuclear Information System (INIS)

    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)

  17. NUCLEAR REACTOR FUEL ELEMENT

    Science.gov (United States)

    Wheelock, C.W.; Baumeister, E.B.

    1961-09-01

    A reactor fuel element utilizing fissionable fuel materials in plate form is described. This fuel element consists of bundles of fuel-bearing plates. The bundles are stacked inside of a tube which forms the shell of the fuel element. The plates each have longitudinal fins running parallel to the direction of coolant flow, and interspersed among and parallel to the fins are ribs which position the plates relative to each other and to the fuel element shell. The plate bundles are held together by thin bands or wires. The ex tended surface increases the heat transfer capabilities of a fuel element by a factor of 3 or more over those of a simple flat plate.

  18. Economic analysis of nuclear reactors

    International Nuclear Information System (INIS)

    The report presents several methods for estimating the power costs of nuclear reactors. When based on a consistent set of economic assumptions, total power costs may be useful in comparing reactor alternatives. The principal items contributing to the total power costs of a nuclear power plant are: (1) capital costs, (2) fuel cycle costs, (3) operation and maintenance costs, and (4) income taxes and fixed charges. There is a large variation in capital costs and fuel expenses among different reactor types. For example, the standard once-through LWR has relatively low capital costs; however, the fuel costs may be very high if U3O8 is expensive. In contrast, the FBR has relatively high capital costs but low fuel expenses. Thus, the distribution of expenses varies significantly between these two reactors. In order to compare power costs, expenses and revenues associated with each reactor may be spread over the lifetime of the plant. A single annual cost, often called a levelized cost, may be obtained by the methods described. Levelized power costs may then be used as a basis for economic comparisons. The paper discusses each of the power cost components. An exact expression for total levelized power costs is derived. Approximate techniques of estimating power costs will be presented

  19. Nuclear reactor fuelling machine

    International Nuclear Information System (INIS)

    The refuelling machine described comprises a rotatable support structure having a guide tube attached to it by a parellel linkage mechanism, whereby the guide tube can be displaced sideways from the support structure. A gripper unit is housed within the guide tube for gripping the end of a fuel assembly or other reactor component and has means for maintenance in the engaging condition during travel of the unit along the guide tube, except for a small portion of the travel at one end of the guide tube, where the inner surface of the guide tube is shaped so as to maintain the gripper unit in a disengaging condition. The gripper unit has a rotatable head, means for moving it linearly within the guide tube so that a component carried by the unit can be housed in the guide tube, and means for rotating the head of the unit through 1800 relative to its body, to effect rotation of a component carried by the unit. The means for rotating the head of the gripper unit comprises ring and pinion gearing, operable through a series of rotatable shafts interconnected by universal couplings. The reason for provision for 1800 rotation is that due to the variation in the neutron flux across the reactor core the side of a fuel assembly towards the outside of the core will be subjected to a lower neutron flux and therefore will grow less than the side of the fuel assembly towards the inside of the core. This can lead to bowing and possible jamming of the fuel assemblies. Full constructional details are given. See also BP 1112384. (U.K.)

  20. Life time of nuclear power plants and new types of reactors; La duree de vie des centrales nucleaires et les nouveaux types de reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2003-05-01

    This report, realized by the Evaluation Parliamentary Office of scientific and technological choices, aims to answer simple but fundamental questions for the french electric power production. What are the phenomena which may limit the exploitation time of nuclear power plants? How can we fight against the aging, at which cost and with which safety? The first chapter presents the management of the nuclear power plants life time, an essential element of the park optimization but not a sufficient element. The second chapter details the EPR and the other reactors for 2015 as a bond between the today and tomorrow parks. The last chapter deals with the necessity of efforts in the research and development to succeed in 2035 and presents other reactors in project. (A.L.B.)

  1. Liquid-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Hydrogen can be added to nuclear reactors with a liquid hydrogen-containing coolant on the suction side of a high pressure pump in the purification system. According to the invention this is performed by means of a liquid jet condenser which uses the coolant as liquid and which is preferably charged from the pressure side of the high pressure pump and conveys the liquid to a mixer connected in series with the high pressure pump. The invention is to be used especially in pressurized water reactors. (orig./PW)

  2. Liquid-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Hydrogen can be added to nuclear reactors with a liquid hydrogen-containing coolant on the suction side of a high pressure pump in the purification system. According to the invention this is performed by means of a liquid jet condenser which uses the coolant as liquid and which is preferably charged from the pressure side of the high pressure pump and conveys the liquid to a mixer connected in series with the high pressure pump. The invention is to be used especially in pressurized water reactors. (orig.)

  3. Atucha-2 PHWR Monte Carlo MCNP5 and KENO-VI models development and application

    International Nuclear Information System (INIS)

    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 KENOVI 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. (authors)

  4. Determination of nitrogen in wheat flour through Activation analysis using Fast neutron flux of a Thermal nuclear reactor; Determinacion de nitrogeno en harina de trigo mediante analisis por activacion empleando el flujo de neutrones rapidos de un reactor nuclear termico

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez G, T

    1976-07-01

    In this work is done a technical study for determining Nitrogen (protein) and other elements in wheat flour Activation analysis, with Fast neutrons from a Thermal nuclear reactor. Initially it is given an introduction about the basic principles of the methods of analysis. Equipment used in Activation analysis and a brief description of the neutron source (Thermal nuclear reactor). The realized experiments for determining the flux form in the irradiation site, the half life of N-13 and the interferences due to the sample composition are included too. Finally, the obtained results by Activation and the Kjeldahl method are tabulated. (Author)

  5. Recycling device of nuclear reactor

    International Nuclear Information System (INIS)

    In a recycling device of a nuclear reactor, a coolant recycling system is disposed by using an outer loop, while a branched connection pipe is connected to a feed water jet pump driving system. Further, the coolant recycling system is constituted with a remaining-heat removing system having a heat exchanger. The connection pipe branched from the downstream of the heat exchanger is connected to the suction side of the jet pump driving pump. Even when feedwater is not returned or returned only insufficiently from a condensate/feedwater system, such as in a case of reactor start up, since sufficient jet pump driving water can be ensured, reactor power can be controlled by controlling the reactor core flow rate by the driving water, to improve the operationability. Further, the burden on control rods can be decreased to improve reliability compared with the case of controlling the power only by the operation of the control rods. Further, since the recycling flow rate of coolants in the reactor core can be ensured sufficiently, occurrence of temperature difference between the upper and the lower portions of a pressure vessel can be prevented effectively, to improve reactor integrity. (N.H.)

  6. Nuclear reactor safety

    International Nuclear Information System (INIS)

    Dr. Buhl feels that nuclear-energy issues are too complex to be understood as single topics, and can only be understood in relationship to broader issues. In fact, goals and risks associated with all energy options must be seen as interrelated with other broad issues, and it should be understood that there are presently no clearcut criteria to ensure that the best decisions are made. The technical community is responsible for helping the public to understand the basic incompatibility of hard and soft technologies and that there is no risk-free energy source. Four principles are outlined for assessing the risks of various energy technologies: (1) take a holistic view; (2) compare the risk with the unit energy output; (3) compare the risk with those of everyday activities; and (4) identify unusual risks associated with a particular option. Dr. Buhl refers to the study conducted by Dr. Inhaber of Canada who used this approach and concluded that nuclear power and natural gas have the lowest overall risk

  7. Nuclear reactor container

    International Nuclear Information System (INIS)

    A gas containing vessel has a water pool which is in communication with a dry well containing a reactor pressure vessel by way of a communication pipe is disposed. A capacity of a gas phase portion of the gas containing chamber, a capacity of the dry well, a water depth of a bent tube communicating the dry well with a pressure suppression pool of a pressure suppression chamber and a water depth of the communication pipe are determined so as to satisfy specific conditions. Since the water depth of the communication pipe is less than the water depth of the bent tube, incondensible gases and steams in the dry well flow into the water pool of the gas containing chamber at the initial stage of loss of coolant accident. Subsequently, steams in the dry well flow into the pressure suppression pool of the pressure suppression chamber by way of the bent tube. Accordingly, since the incondensible gases in the dry well do not flow into the pressure suppression chamber, pool swelling phenomenon in the pressure suppression chamber is not caused even if the water depth of the bent tube which leads to the pressure suppression chamber is great. Further, pressure increase due to transfer of the incondensible gases is decreased. (I.N.)

  8. Análisis de secuencias de accidente de pérdida de refrigerante en centrales nucleares de agua a presión mediante la metodología ISA

    OpenAIRE

    González Cadelo, Juan

    2015-01-01

    El accidente de pérdida de refrigerante (LOCA) en un reactor nuclear es uno de los accidentes Base de Diseño más preocupantes y estudiados desde el origen del uso de la tecnología de fisión en la industria productora de energía. El LOCA ocupa, desde el punto de vista de los análisis de seguridad, un lugar de vanguardia tanto en el análisis determinista (DSA) como probabilista (PSA), cuya diferenciada perspectiva ha ido evolucionando notablemente en lo que al crédito a la actuación de las salv...

  9. Water cooled FBNR nuclear reactor

    International Nuclear Information System (INIS)

    Full text: The world with its increasing population and the desire for a more equitable and higher standard of living, is in the search for energy that is abundant and does not contribute to the problem of global warming. The answer to this is a new paradigm in nuclear energy; i.e., through the innovative nuclear reactors that meet the IAEA's INPRO philosophies and criteria that will guarantee the generation of safe and clean energy. The emerging countries to nuclear energy that are not in hurry for energy and look into the future are looking into the participation in the development of such innovative nuclear reactors. They can start developing the non-nuclear components of such reactors in parallel with creating the nuclear infra-structures according to the guidelines of the IAEA suggested in its milestones document. In this way, they can benefit from numerous advantages that the development of a high technology can bring to their countries be it scientific, technological, economic or political. A solution to the present world economic crisis is investing in such projects that contribute to the real economy rather than speculative economy. This will help both local and world economy. One such innovative nuclear reactor is the FBNR that is being developed with the support of the IAEA in its program of Small Reactors Without On-site Refuelling. It is a small (70 MWe) reactor with simple design based on the proven PWR technology (www.sefidvash.net/fbnr). The simplicity in design and the world wide existence of water reactor technology, makes it a near term project compared to other future reactors. Small reactors are most adequate for both the developing and developed countries. They require low capital investment, and can be deployed gradually as energy demand calls for. The generation of energy at the local of consumption avoids high cost of energy transmission. The paradigm of economy of scale does not apply to the FBNR as it is a small reactor by its nature. The

  10. Three dimensional diffusion calculations of nuclear reactors

    International Nuclear Information System (INIS)

    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.)

  11. Applied monitoring methods for the control of fuel elements and reactor internals in Argentine nuclear power plants poolside facilities

    International Nuclear Information System (INIS)

    Two pressurized heavy water reactors are being operated in Argentina: Atucha 1, a Siemens KWU designed and built pressure vessel 345 MWe reactor, started in 1974, and Embalse, a CANDU-600 MWe, started in 1983. Post-irradiation tests have been performed in accordance with the needs of fuel development and control programs for internal parts, such as the coolant channels (CC) where the fuel elements (FE) dwelled. The bays built in both NPP offer a good place to perform a large variety of pool-side inspection and testing activities. During the last 20 years more than 2000 CANDU-type FE, more than 1000 Atucha FE and above 100 Atucha CC, have been inspected in these bays. The usual inspection activities consist of: visual inspection, sipping tests, disassembling of FE, metrology, gamma scanning and eddy current testing. In this paper all of these techniques are briefly described and the obtained experience is shown. (author)

  12. Production of radionuclides in nuclear reactor

    International Nuclear Information System (INIS)

    Given is a short review on the production of radionuclides which was performed in the Vinca Institute of Nuclear Sciences by using the nuclear reactor RA. Regarding the considerations of the possible re-starting of this reactor its use for the production of medical radionuclides should be taken into account. Listed are some of the important medical radionuclides routinely produced in nuclear reactors in the world and discussed the conditions for their obtaining in the reactor RA. (author)

  13. Actinide transmutation in nuclear reactors

    International Nuclear Information System (INIS)

    This report has also been published as a PhD thesis. It discusses the reduction of the transuranics part of nuclear waste. Requirements and criteria for efficient burning of transuranics are developed. It is found that a large reduction of transuranics produced per unit of energy is possible when the losses in reprocessing are small and when special transuranics burner reactors are used at the end of the nuclear era to reduce the transuranics inventory. Two special burner reactors have been studied in this thesis. In chapter 3, the Advanced Liquid Metal Reactor is discussed. A method has been developed to optimize the burning capability while complying to constraints imposed on the design for safety, reliability, and economics. An oxide fueled and metallic fueled ALMR have been compared for safety and transuranics burning. Concluded is that the burning capability is the same, but that the higher thermal conductivity of the metallic fuel has a positive effect on safety. In search for a more effective waste transmuter, a modified Molten Salt Reactor was designed for this study. The continuous refueling capability and the molten salt fuel make a safe design possible without uranium as fuel. A four times faster reduction of the transuranics is possible with this reactor type. The amount of transuranics can be halved every 10 years. The most important conclusion of this work is that it is of utmost importance in the study of waste transmutation that a high burning is obtained with a safe design. In future work, safety should be the highest priority in the design process of burner reactors. (orig.)

  14. Reactors for nuclear electric propulsion

    International Nuclear Information System (INIS)

    Propulsion is the key to space exploitation and power is the key to propulsion. This paper examines the role of nuclear fission reactors as the primary power source for high specific impulse electric propulsion systems for space missions of the 1980s and 1990s. Particular mission applications include transfer to and a reusable orbital transfer vehicle from low-Earth orbit to geosynchronous orbit, outer planet exploration and reconnaissance missions, and as a versatile space tug supporting lunar resource development. Nuclear electric propulsion is examined as an indispensable component in space activities of the next two decades

  15. Nuclear reactors in remote earth

    International Nuclear Information System (INIS)

    Same basic geological principles along with other facts, have allowed us to establish the existence in the remote past (Between 2.5 and 4 x 10''9 years ago) of the uranium deposits and/or uranium mineralized volumes, which be-have as nuclear reactors. A simplified neutronic diffusion model have allowed us to describe the main characteristics of such systems. The obtained results indicate that this phenomenon was a rather frequent fact. (Author) 7 refs

  16. Gas-cooled nuclear reactor

    International Nuclear Information System (INIS)

    The gas temperature of a hot gas loop in gas-cooled nuclear reactor plants shall be able to be modified without influencing the gas temperature of the other loops. If necessary, it should be possible to stop the loop. This is possible by means of a mixer which is places below the heat absorbing component in the hot channel and which is connected to a cold gas line. (orig.)

  17. Advance: research project on aging electrical wiring in nuclear power plants; Advance: proyecto de investigacion de envejecimiento en cableado electrico en centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Cano, J. C.; Ruiz, S.

    2013-07-01

    As Nuclear Power Plants get older it is more important to know the real condition of low voltage, instrumentation, power and control cables. Additionally, as new plants are being built, the election of cables and the use of in-situ monitoring techniques to get reliable aging indicators, can be very useful during the plant life. The goal of this Project is to adapt, optimize and asses Condition Monitoring techniques for Nuclear Power Plants cables. These techniques, together with the appropriate acceptance criteria, would allow specialists to know the state of the cable over its entire length and estimate its residual life. In the Project, accelerated ageing is used in cables installed in European NPPs in order to evaluate different techniques to detect local and global ageing. Results are compared with accepted tests to validate its use for the estimation of cables residual life. This paper describes the main stages of the Project and some results. (Author)

  18. Demand of natural uranium to satisfy the requirements of nuclear fuel of new nuclear power plants in Mexico; Demanda de uranio natural para satisfacer los requerimientos de combustible nuclear de nuevas centrales nucleares en Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez S, J. R.; Rios, M. del C.; Alonso, G.; Palacios H, J. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: jrrs@nuclear.inin.mx

    2008-07-01

    Due to the expectation of that in Mexico new plants of nuclear energy could be installed, turns out from the interest to evaluate the uranium requirements to operate those plants and to also evaluate if the existing reserves in the country could be sufficient to satisfy that demand. Three different scenes from nuclear power plant expansion for the country are postulated here that are desirable for the diversification of generation technologies. The first scene considers a growth in the generation by nuclear means of two reactors of type ABWR that could enter operation by years 2015 and 2020, in the second considers the installation of four reactors but as of 2015 and new every 5 years, in the scene of high growth considers the installation of 6 reactors of the same type that in the other scenes, settling one every three years as of 2015. The results indicate that the uranium reserves could be sufficient to only maintain in operation to one of the reactors proposed by the time of their useful life. (Author)

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

    International Nuclear Information System (INIS)

    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

  20. Modelización y validación experimental del transporte de tritio en metal líquido y otros materiales para reactores de fusión nuclear magnética e inercial

    OpenAIRE

    Moral Fernández, Nuria

    2015-01-01

    La presente tesis se centra en el estudio de los fenómenos de transporte de los isótopos de hidrógeno, y más concretamente del tritio, en materiales de interés para los reactores de fusión nuclear. Los futuros reactores de fusión nuclear necesitarán una Planta de Tritio, con una envoltura regeneradora (breeding blanket) y unos sistemas auxiliares claves para su diseño. Por lo tanto su desarrollo y cualificación son cruciales para demostrar que los reactores de fusión son una opción viable com...

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

    International Nuclear Information System (INIS)

    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

  2. Research nuclear reactor operation management

    International Nuclear Information System (INIS)

    Some aspects of reactor operation management are highlighted. The main mission of the operational staff at a testing reactor is to operate it safely and efficiently, to ensure proper conditions for different research programs implying the use of the reactor. For reaching this aim, there were settled down operating plans for every objective, and procedure and working instructions for staff training were established, both for the start-up and for the safe operation of the reactor. Damages during operation or special situations which can arise, at stop, start-up, maintenance procedures were thoroughly considered. While the technical skill is considered to be the most important quality of the staff, the organising capacity is a must in the operation of any nuclear facility. Staff training aims at gaining both theoretical and practical experience based on standards about staff quality at each work level. 'Plow' sheet has to be carefully done, setting clear the decision responsibility for each person so that everyone's own technical level to be coupled to the problems which implies his responsibility. Possible events which may arise in operation, e.g., criticality, irradiation, contamination, and which do not arise in other fields, have to be carefully studied. One stresses that the management based on technical and scientific arguments have to cover through technical, economical and nuclear safety requirements a series of interlinked subprograms. Every such subprograms is subject to some peculiar demands by the help of which the entire activity field is coordinated. Hence for any subprogram there are established the objectives to be achieved, the applicable regulations, well-defined responsibilities, training of the personnel involved, the material and documentation basis required and activity planning. The following up of positive or negative responses generated by experiments and the information synthesis close the management scope. Important management aspects

  3. Mixed wastes treatment in Atucha I

    International Nuclear Information System (INIS)

    Full text: During decontamination works of the fueling machine of Atucha I nuclear power plant (AINPP), a liquid waste with special characteristics was generated, which needed the development of a treatment method. The waste consisted of an emulsion designed for the cleaning of mechanical components and was formed by an organic solvent dispersed in water with aid of an emulsifier additive. After several cleaning operations, the emulsion contained an important quantity of lubricants and radioactive dirt. The treatment had the objective of recycling a toxic waste such as the organic solvent and reducing the volume of the residual mass. Laboratory tests were made tending to the emulsion separation in their components. Ionic force and ionic mobility were modified for join the emulsion micelles and produce their coalescence. Different salts and working temperatures were tried and it was stated that the combination of 1% of Na2SO4 added and 40 degree C temperature were the optimum taking into account the available equipment in AINPP and cost considerations. The process was carried out in batch mode and 3 residual streams were obtained, an aqueous one which was sent to Residual Water System of AINPP, an organic liquid consisting of decontaminated hydrocarbons, useful for other cleaning tasks and finally a solid one, sited in the in-between interface of the other two liquids, consisting of insoluble soaps used as lubricant thickness, containing the principal proportion of radioactivity. As a result of this process we have achieved a volume reduction higher than 90%, the recycling of the organic solvent and concentration of radioactivity in a solid greasy mass with low water solubility. (author)

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

    International Nuclear Information System (INIS)

    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)

  5. Research means to back the development of nuclear reactors; Les moyens de recherche en support a l'evolution des reacteurs nucleaires

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2005-07-01

    After 50 year long feedback experience on nuclear reactor operations it is legitimate to wonder whether experimental facilities used to support nuclear power programs are still necessary. The various participants of this conference said yes for mainly 4 reasons: -) to validate the extension of the service life of a reactor without putting at risk its high safety standard, -) to give the reactor more flexibility to cope with the power demand, -) to confront the results given by computerized simulations with experimental data, and -) to qualify the nuclear systems of tomorrow. (A.C.)

  6. Negative sequence relay applied to generator 1 of the Laguna Verde nuclear power plant; Aplicacion de un relevador de secuencia negativa en el generador 1 de la central de Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Diaz de la Serna P, Enrique [Instituto de Investigaciones Electricas, Cuernavaca (Mexico)

    1988-12-31

    The rotor of a synchronous generator can be dangerously heated in a short time by stator current unbalance, therefore it must be protected with a specific relay. This article discusses the protection and the adjustments selected for Unit 1 of the Comision Federal de Electricidad Laguna Verde Nuclear Nuclear Power Station. [Espanol] El rotor de un generador sincrono puede calentarse peligrosamente en un tiempo corto debido a desbalance de corrientes en el estator, por lo que debe protegerse con un relevador especifico. En este articulo se describen la proteccion y los ajustes seleccionados para la unidad 1, de la central nucleoelectrica Laguna Verde de la Comision Federal de Electricidad.

  7. Nuclear reactors Monitoring using neutrinos detectors1

    International Nuclear Information System (INIS)

    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 and how it could be used for nuclear safeguards policy.

  8. Exporting apocalypse: CANDU reactors and nuclear proliferation

    International Nuclear Information System (INIS)

    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

  9. Nuclear reactors appointment book Uruguay at the moment

    International Nuclear Information System (INIS)

    This essay have included among its chapters Energy and development, fuels, Nuclear Energy, types of Nuclear Reactors, main objections against Power Nuclear Reactors, other Reactors proposals, legal framework and Nuclear safety in Uruguay

  10. Seismic Design of Nuclear Reactor

    International Nuclear Information System (INIS)

    In case the requirement of design is against natural phenomena, it is important to grasp the detailed characteristics of the natural phenomena for the proper design, and as the grasp is more strict and accurate, the design of high adaptability or durability to the requirement can be done. The aseismatic design of nuclear reactors is similar to it, and the decision of the magnitude of supposed earthquakes is important. The aseismatic design of nuclear power stations in Japan has been carried out in conformity with the national guideline for examining the aseismatic design. The aseismatic design of nuclear reactors is carried out in the order of the survey of geological features, ground and earthquakes, the determination of the input magnitude and characteristics of earthquakes, the formation of simulated earthquake waves, the analysis of the response of buildings and structures to earthquakes, and structural analysis. The decision of input earthquakes is done by the detailed historical earthquake data based on local features and the survey of geological features and ground. The determination of earthquake input, the analysis of earthquake response and structural analysis, and the other features of the aseismatic design are explained. (K.I.)

  11. Comments on nuclear reactor safety in Ontario

    International Nuclear Information System (INIS)

    The Chalk River Technicians and Technologists Union representing 500 technical employees at the Chalk River Nuclear Laboratories of AECL submit comments on nuclear reactor safety to the Ontario Nuclear Safety Review. Issues identified by the Review Commissioner are addressed from the perspective of both a labour organization and experience in the nuclear R and D field. In general, Local 1568 believes Ontario's CANDU nuclear reactors are not only safe but also essential to the continued economic prosperity of the province

  12. 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)

  13. Artificial intelligence in nuclear reactor operation

    International Nuclear Information System (INIS)

    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)

  14. Plenum separator system for pool-type nuclear reactors

    International Nuclear Information System (INIS)

    A pool-type liquid-metal-heat-transfer nuclear reactor is described which consists of: a vertically disposed generally cylindrical reactor vessel having a closed bottom portion, and a closure head atop the reactor vessel and closing the reactor vessel; the reactor vessel enclosing the major components of the nuclear reactor which include a reactor core supported at a centrally disposed lower portion of the reactor vessel; a bottom-supported gas-plenum-forming hollow cylindrical member closed at its upper end, the hollow cylindrical member sealed to and supported by the reactor vessel; a hollow cylindrically conformed neutron shield member spaced from and radially surrounding the reactor core; separate liquid-metal plena confining liquid metal during normal reactor operation and comprising a hot upper plenum, a cold lower plenum and intermediate temperature plena; the liquid metal intakes of the liquid metal pumps positioned in the cold lower plenum with the cooler liquid metal therein being pumped upwardly through the reactor core to be heated and exit therefrom in a turbulent fashion; and the liquid metal intakes of the heat exchangers positioned within the hot upper plenum and the liquid metal discharges of the heat exchangers positioned within the cold lower plenum to discharge cooled liquid metal into the cold lower plenum

  15. Molten salts in nuclear reactors

    International Nuclear Information System (INIS)

    Collection of references dealing with the physicochemical studies of fused salts, in particular the alkali and alkali earth halides. Numerous binary, ternary and quaternary systems of these halides with those of uranium and thorium are examined, and the physical properties, density, viscosity, vapour pressure etc... going from the halides to the mixtures are also considered. References relating to the corrosion of materials by these salts are included and the treatment of the salts with a view to recuperation after irradiation in a nuclear reactor is discussed. (author)

  16. Internal structure for nuclear reactor

    International Nuclear Information System (INIS)

    The description is given of an internal structure for a nuclear reactor of the type having inside a presssure vessel a core composed of a number of fuel assemblies and a number of mobile control components. It includes a bottom grid integral with the vessel on which are secured the fuel assemblies, an intermediate grid located above the assemblies and also integral with the vessel, an upper grid located at the upper part of the vessel and integral with it and multiple vertical maintenance devices, extending between the upper and intermediate grids

  17. Autonomous Control of Space Nuclear Reactors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Nuclear reactors to support future robotic and manned missions impose new and innovative technological requirements for their control and protection...

  18. The application problems of nuclear reactors

    International Nuclear Information System (INIS)

    Latvia is surrounded by closely located nuclear reactors. In a distance of 1000 km from Latvia there are 62 working and 12 suspended high power nuclear reactors. Near the borders of Latvia, in the 3 km range four countries are exploiting 12 nuclear reactors, whose reliability and safe operation is always arousing profound interest in our community. On estimating the prospects of Latvian energetics we can conclude that at the beginning of the next century it will be extremely complicated task to supply our country with electricity and heat without nuclear reactors. This is due to lack of the domestic energy resources and to the necessity of reducing harmful pollutions of TECs. (authors)

  19. Cleaning chemical and mechanical of heat exchangers in french nuclear plants; Limpieza mecanica y quimica de intercambiadores de calor en centrales nucleares francesas

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz, J. t.; Guerra, P.; Carreres, C.

    2013-03-01

    This project was carried out under the frame of the approval of LAINSA as a supplier of EDF in France. The inspection performed on systems called the moisture separator reheaters (GSS) of CPO series reactor of EDF nuclear power plants has shown evidence of significant clogging due to deposits of magnetite inside the tubes of tube bundle. The pressure drop between inlet and outlet of the heating was close to maximum design criterion. This effect could result in equipment damage and loss of plant productivity. The aim of the work was the design, development, approval and implementation of a procedure for un blocking the tubes of the GSS respecting the integrity of materials and ensuring the harmlessness of cleaning procedures. The procedure used was to completely remove magnetite deposits in order to recover a passage diameter and a surface finish equivalent to the origin, thus avoiding the replacement of the GSS and obtaining a considerable reduction of costs. The achieve these objectives we have developed a procedure that is basically a mechanical pre-cleaning of all tubes of the GSS in order to unblock tem, followed by a chemical cleaning where magnetite is dissolved and crawled out of the tube bundle. The main results were: -Corrosion less than 10 microns. 100-110 Kg of magnetite removed by heat exchanger. -Final pressure drop similar to that of new equipment. -Waste water: 70 m{sup 3} per exchanger, which were managed by an authorized waste management company. This procedure has been applied successfully in 14 GSS type heat exchangers in Fessenheim and Bugey nuclear power plants in France between 2009 and 2011. This project demonstrates that the long experience of LAINSA in the Spanish nuclear industry along with the knowledge and experience in chemical cleaning of SOLARCA, have served to successfully work demanding and mature markets such as the French nuclear market, solving the problem of deposits of magnetite with an effective and safe method for the treated

  20. Experimental possibilities of IGR reactor for the researches on the nuclear reactor safety

    International Nuclear Information System (INIS)

    The IGR reactor (National nuclear centre of the Republic of Kazakstan, Kurchatov) with high technical and neutron-physical properties has wide experimental possibilities for the dynamic studies. On this reactor possible curried out two general types of regimes. First regime is a 'flare', non-regular neutron impulse of power of bell form. In this regime the maximum flux density of thermal neutrons. Second regime is a 'impulse', regulated on the given regime (law) power impulse. Profile of power change in this regime has sections of linear ascent and fall, sections of stationary power. IGR reactor has pneumatic hydraulic stand, provided accumulation in the ramps of high pressure. Experimental volume of the reactor are composites central and lateral channels, which are passed through active zone of height equal to 1400 mm. The above mentioned possibilities of IGR reactor are provided unique conditions for studies in the field of nuclear reactor safety

  1. Helium desorption in EFDA iron materials for use in nuclear fusion reactors; Desorcion de helio en materiales de fierro EFDA para su aplicacion en los reactores de fusion nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Salazar R, A. R.; Pinedo V, J. L. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Sanchez, F. J.; Ibarra, A.; Vila, R., E-mail: arsr2707@hotmail.com [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas, Av. Complutense No. 40, 28040 Madrid (Spain)

    2015-09-15

    In this paper the implantation with monoenergetic ions (He{sup +}) was realized with an energy of 5 KeV in iron samples (99.9999 %) EFDA (European Fusion Development Agreement) using a collimated beam, after this a Thermal Desorption Spectrometry of Helium (THeDS) was made using a leak meter that detects amounts of helium of up to 10{sup -}- {sup 12} mbar l/s. Doses with which the implantation was carried out were 2 x 10{sup 15} He{sup +} /cm{sup 2}, 1 x 10{sup 16} He{sup +} /cm{sup 2}, 2 x 10{sup 16} He{sup +} /cm{sup 2}, 1 x 10{sup 17} He{sup +} /cm{sup 2} during times of 90 s, 450 s, 900 s and 4500 s, respectively. Also, using the SRIM program was calculated the depth at which the helium ions penetrate the sample of pure ion, finding that the maximum distance is 0.025μm in the sample. For this study, 11 samples of Fe EFDA were prepared to find defects that are caused after implantation of helium in order to provide valuable information to the manufacture of materials for future fusion reactors. However understand the effects of helium in the micro structural evolution and mechanical properties of structural materials are some of the most difficult questions to answer in materials research for nuclear fusion. When analyzing the spectra of THeDS was found that five different groups of desorption peaks existed, which are attributed to defects of He caused in the material, these defects are He{sub n} V (2≤n≤6), He{sub n} V{sub m}, He V for the groups I, II and IV respectively. These results are due to the comparison of the peaks presented in the desorption spectrum of He, with those of other authors who have made theoretical calculations. Is important to note that the thermal desorption spectrum of helium was different depending on the dose with which the implantation of He{sup +} was performed. (Author)

  2. Nuclear reactor built, being built, or planned

    International Nuclear Information System (INIS)

    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

  3. Liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Reference is made to liquid metal cooled nuclear reactors of the 'pool' type. In such reactors the core, the heat exchangers, and the coolant circulating pumps are submerged in a pool of liquid metal. In operation of the reactor it is necessary to be able to locate and identify components submerged in the pool, and before moving rotating shields in the roof of the pool-containing vault it is necessary to ensure that all the normally suspended absorber rods have been inserted in the core and released from their suspensions. Television cameras are unsuitable for use in the opaque liquid metal but ultrasound in the megahertz range has been used to give a television screen kind of display. There is some difficulty, however, in transmitting ultrasound signals from a transducer into the pool of coolant because the transducer must be protected from the high temperature environment of the coolant. This difficulty has been partially overcome, however, by transmitting the signals by way of a wave guide extending from the transducer into the coolant pool. Such a wave guide may comprise a column of liquid metal within a dip tube. The column of liquid coolant is uninterrupted by a supporting diaphragm. Such a system is here described. (U.K.)

  4. Decommissioning of nuclear reactor systems

    International Nuclear Information System (INIS)

    The decision-making process involving the decommissioning of the British graphite-moderated, gas-cooled Magnox power stations is complex. There are timing, engineering, waste disposal, cost and lost generation capacity factors and the ultimate uptake of radiation dose to consider and, bearing on all of these, the overall decision of when and how to proceed with decommissioning may be heavily weighed by political and public tolerance dimensions. These factors and dimensions are briefly reviewed with reference to the ageing Magnox nuclear power stations, of which Berkeley and Hunterston A are now closed down and undergoing the first stages of decommissioning and Trawsfynydd, although still considered as available capacity, has had both reactors closed down since February 1991 and is awaiting substantiation and acceptance of a revised reactor pressure vessel safety case. Although the other first-generation Magnox power station at Hinkley Point, Bradwell, Dungeness and Sizewell are operational, it is most doubtful that these stations will be able to eke out a generating function for much longer. It is concluded that the British nuclear industry has adopted a policy of deferred decommissioning, that is delaying the process of complete dismantlement of the radioactive components and assemblies for at least one hundred years following close-down of the plant. (Author)

  5. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  7. Notions and methodologies for uncertainty analysis in simulations of transitory events of a nuclear central; Nociones y metodologias para analisis de incertidumbre en simulaciones de eventos transitorios de una central nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Alva N, J. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Ingenieria Nuclear, Avenida IPN S/N Colonia Lindavista, 07738 Mexico D.F. (Mexico); Ortiz V, J.; Amador G, R.; Delfin L, A. [ININ, Carretera Mexico-Toluca SN La Marquesa, 52750 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: neriaesfm@gmail.com

    2007-07-01

    The present work has as objective to gather the basic notions related with the uncertainty analysis and some of the methodologies to be applied in the studies of transitory events analysis of a nuclear power station, in particular of those thermal hydraulics phenomena. The concepts and methodologies mentioned in this work are the result of an exhaustive bibliographical investigation of the topic in the nuclear area. The methodologies of uncertainties analysis have been developed by diverse institutions and they are broadly used at world level for their application in the results of the computer codes of the class of better estimation in the thermal hydraulics analysis and safety of plants and nuclear reactors. The main sources of uncertainty, types of uncertainty and aspects related with the models of better estimation and better estimation method are also presented. (Author)

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

    International Nuclear Information System (INIS)

    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 1s4s. 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

  9. Licensed reactor nuclear safety criteria applicable to DOE reactors

    International Nuclear Information System (INIS)

    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

  10. Proliferation Resistant Nuclear Reactor Fuel

    International Nuclear Information System (INIS)

    Global appetite for fission power is projected to grow dramatically this century, and for good reason. Despite considerable research to identify new sources of energy, fission remains the most plentiful and practical alternative to fossil fuels. The environmental challenges of fossil fuel have made the fission power option increasingly attractive, particularly as we are forced to rely on reserves in ecologically fragile or politically unstable corners of the globe. Caught between a globally eroding fossil fuel reserve as well as the uncertainty and considerable costs in the development of fusion power, most of the world will most likely come to rely on fission power for at least the remainder of the 21st century. Despite inevitable growth, fission power faces enduring challenges in sustainability and security. One of fission power's greatest hurdles to universal acceptance is the risk of potential misuse for nefarious purposes of fissionable byproducts in spent fuel, such as plutonium. With this issue in mind, we have discussed intrinsic concepts in this report that are motivated by the premise that the utility, desirability, and applicability of nuclear materials can be reduced. In a general sense, the intrinsic solutions aim to reduce or eliminate the quantity of existing weapons usable material; avoid production of new weapons-usable material through enrichment, breeding, extraction; or employ engineering solutions to make the fuel cycle less useful or more difficult for producing weapons-usable material. By their nature, these schemes require modifications to existing fuel cycles. As such, the concomitants of these modifications require engagement from the nuclear reactor and fuel-design community to fully assess their effects. Unfortunately, active pursuit of any scheme that could further complicate the spread of domestic nuclear power will probably be understandably unpopular. Nevertheless, the nonproliferation and counterterrorism issues are paramount, and

  11. Flow visualization techniques in nuclear reactors, (1)

    International Nuclear Information System (INIS)

    Heat energy generated in nuclear reactors is transferred by coolants to utilizing systems such as electric power and process industries etc. Therefore, heat removal characteristics of nuclear reactors depends on flow conditions of coolants in a reactor core and in cooling systems. In order to make clear flow patterns of these coolants, the flow visualization method is often applied prior to actual measurements of pressure, velocity and so on. This paper describes basic techniques for flow visualization especially in nuclear reactor, and gives applied examples of this technique. (author)

  12. Problems of nuclear reactor safety. Vol. 1

    International Nuclear Information System (INIS)

    Proceedings of the 9. Topical Meeting 'Problems of nuclear reactor safety' are presented. Papers include results of studies and developments associated with methods of calculation and complex computerized simulation for stationary and transient processes in nuclear power plants. Main problems of reactor safety are discussed as well as rector accidents on operating NPP's are analyzed

  13. Nuclear reactor simulator for a teaching laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kosilov, A.N. (Moscow Engineering Physics Inst. (USSR))

    A nuclear reactor simulator is described which has been developed by the Department of Automatics and Telemechanics of the Moscow Engineering Physics Institute to provide students with an insight into, and familiarity with the characteristics of a nuclear reactor and its systems through evaluation, manipulation and experimentation.

  14. Nuclear reactor fuel rod spacer

    International Nuclear Information System (INIS)

    A spacer for positioning at least the four corner fuel rods in a tubular flow channel of a nuclear reactor is disclosed. The spacer comprises a support member having four side bands interconnected by four corner bands to form a unitary structure. Each of the side bands has a L-shaped lobe adjacent to each of its ends with one leg of each lobe extending to the adjacent end of its side band. Each of the corner bands is narrower than the side bands and is offset so as to be spaced from the lobe. One leg of each lobe is positioned to engage the tubular flow channel to maintain proper spacing between the flow channel and the adjacent corner fuel rod and to improve the thermal-hydraulic performance of the spacer

  15. Nuclear reactor internals alignment configuration

    Science.gov (United States)

    Gilmore, Charles B.; Singleton, Norman R.

    2009-11-10

    An alignment system that employs jacking block assemblies and alignment posts around the periphery of the top plate of a nuclear reactor lower internals core shroud to align an upper core plate with the lower internals and the core shroud with the core barrel. The distal ends of the alignment posts are chamfered and are closely received within notches machined in the upper core plate at spaced locations around the outer circumference of the upper core plate. The jacking block assemblies are used to center the core shroud in the core barrel and the alignment posts assure the proper orientation of the upper core plate. The alignment posts may alternately be formed in the upper core plate and the notches may be formed in top plate.

  16. Emergency cooling system for nuclear reactors

    International Nuclear Information System (INIS)

    Upon loss of coolant in a nuclear reactor as when a coolant supply or return line breaks, or both lines break, borated liquid coolant from an emergency source is supplied in an amount to absorb heat being generated in the reactor even after the control rods have been inserted. The liquid coolant flows from pressurized storage vessels outside the reactor to an internal manifold from which it is distributed to unused control rod guide thimbles in the reactor fuel assemblies. (author)

  17. Nuclear reactors: Notifiable events in 2002

    International Nuclear Information System (INIS)

    Notifiable events in nuclear power plants in the Federal Republic of Germany are reported to the regulatory authorities under the Atomic Energy Act in accordance with standardized national reporting criteria, and are recorded centrally. The binding legal provisions covering these reports can be found in the Nuclear Safety Commissioner and Reporting Ordinance (AtSMV). On an international level, events are classified in the International Nuclear Event Scale (INES) comprising eight levels. The four quarterly reports covering 2002 include 167 notifiable events for nuclear power plants in operation and in the decommissioning stage. Of these events, 157 are in reporting category N (normal), while ten are in reporting category E (urgent). No events have been reported in category S (immediate). 154 events are INES level 0, 13 events are INES level 1. 13 category-N events were reported for research reactors. All of them are INES level 0. There were no releases of radioactive material above the licensed levels for ex-vent air and liquid effluents. (orig.)

  18. Noise thermometry in nuclear reactors

    International Nuclear Information System (INIS)

    Since in nuclear reactors the measuring sensor cannot be easily replaced, the value of the sensor resistance, as well as the selection of transmission lines with respect to good transmission characteristics of the whole arrangement and minimizing the correlative error terms, must already be optimized when designing a noise thermometer arrangement. The TRARAU computer program was developed for this purpose enabling the influences of the lines to be computed by taking into consideration all the effects occurring through the lines, such as transmission errors and correlative error terms. In order to check the accuracy of the TRARAU computer program a series of laboratory measurements were implemented enabling both the pure transmission behaviour of the line arrangement with respect to the measuring signal to be detected, as well as the overall line error. In all cases this resulted in a very good agreement of the measured values with the computed values. The transmission behaviour of noise thermometer arrangements occuring in practice were studied with the example of two reactor experiments. In both cases it was possible to demonstrate successfully the potential of the computer program TRARAU. As the parametric studies have shown, optimum matching over unlimited band widths is not feasible in principle. By reducing the upper band limit, however, the line error can practically always be kept sufficiently small. With good matching larger band widths can also be used. (orig./HP)

  19. Nuclear Archeology for CANDU Power Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Broadhead, Bryan L [ORNL

    2011-01-01

    The goal of this work is the development of so-called 'nuclear archeology' techniques to predict the irradiation history of both fuel-related and non-fuel-related materials irradiated in the CANDU (CANada Deuterium Uranium) family of nuclear reactors. In this application to CANDU-type reactors, two different scenarios for the collection of the appropriate data for use in these procedures will be assumed: the first scenario is the removal of the pressure tubes, calandria tubes, or fuel cladding and destructive analysis of the activation products contained in these structural materials; the second scenario is the nondestructive analysis (NDA) of the same hardware items via high-resolution gamma ray scans. There are obvious advantages and disadvantages for each approach; however, the NDA approach is the central focus of this work because of its simplicity and lack of invasiveness. The use of these techniques along with a previously developed inverse capability is expected to allow for the prediction of average flux levels and irradiation time, and the total fluence for samples where the values of selected isotopes can be measured.

  20. Nuclear reactor safety research in Kazakhstan

    International Nuclear Information System (INIS)

    Full text : The paper summarizes activities being implemented by the National Nuclear Center of the Republic of Kazakhstan in support of safe operation of nuclear reactors; shows its crucial efforts and further road map in this line. As is known, the world community considers nuclear reactor safety as one of the urgent research areas. Kazakhstan has been pursuing studies in support of nuclear energy safety since early 80s. The findings allow to coordinate available computational methods and design new ones while validating new NPP Projects and making analysis for reactor installations available

  1. Physical Characteristics of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    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

  2. Validation of finite element code DELFIN by means of the zero power experiences at the nuclear power plant of Atucha I

    International Nuclear Information System (INIS)

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

  3. Daddy, What's a Nuclear Reactor?

    International Nuclear Information System (INIS)

    No matter what we think of the nuclear industry, it is part of mankind's heritage. The decommissioning process is slowly making facilities associated with this industry disappear and not enough is being done to preserve the information for future generations. This paper provides some food for thought and provides a possible way forward. Industrial archaeology is an ever expanding branch of archaeology that is dedicated to preserving, interpreting and documenting our industrial past and heritage. Normally it begins with analyzing an old building or ruins and trying to determine what was done, how it was done and what changes might have occurred during its operation. We have a unique opportunity to document all of these issues and provide them before the nuclear facility disappears. Entombment is an acceptable decommissioning strategy; however we would have to change our concept of entombment. It is proposed that a number of nuclear facilities be entombed or preserved for future generations to appreciate. This would include a number of different types of facilities such as different types of nuclear power and research reactors, a reprocessing plant, part of an enrichment plant and a fuel manufacturing plant. One of the main issues that would require resolution would be that of maintaining information of the location of the buried facility and the information about its operation and structure, and passing this information on to future generations. This can be done, but a system would have to be established prior to burial of the facility so that no information would be lost. In general, our current set of requirements and laws may need to be re-examined and modified to take into account these new situations. As an alternative, and to compliment the above proposal, it is recommended that a study and documentation of the nuclear industry be considered as part of twentieth century industrial archaeology. This study should not only include the power and fuel cycle

  4. Nuclear reactor philosophy and criteria

    International Nuclear Information System (INIS)

    Nuclear power plant safety criteria and principles developed in Canada are directed towards minimizing the chance of failure of the fuel and preventing or reducing to an acceptably low level the escape of fission products should fuel failure occur. Safety criteria and practices are set forth in the Reactor Siting Guide, which is based upon the concept of defence in depth. The Guide specifies that design and construction shall follow the best applicable code, standard or practice; the total of all serious process system failures shall not exceed one in three years; special safety systems are to be physically and functionally separate from process systems and each other; and safety systems shall be testable, with unavailability less than 10-3. Doses to the most exposed member of the public due to normal operation, serious process failures, and dual failures are specified. Licensees are also required to consider the effects of extreme conditions due to airplane crashes, explosions, turbine disintegration, pipe burst, and natural disasters. Safety requirements are changing as nuclear power plant designs evolve and in response to social and economic pressures

  5. Requirements of coolants in nuclear reactors

    International Nuclear Information System (INIS)

    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)

  6. Generation III reactors - the nuclear renaissance

    International Nuclear Information System (INIS)

    The European Pressurized Reactor - GEN III+, the PWR type reactor, remains the world's first and currently being built power reactor everywhere. ATMEA1, a new 1,100 MWe pressurized water reactor combines state-of-the art- technology from AREVA and Mitsubishi Heavy Industries to meet the challenges of the nuclear renaissance. Thus, the next evolutionary design of Generation III reactors will be deployed over many decades and will represent a large part of the worldwide fleet throughout the 21st century. Generation III reactors will equip the future NPPs ensuring improved safety and reliability, with passive safety systems and a very low probability for core melt. The Generation III Reactors as 'The Nuclear Renaissance' is presented in the paper. (author)

  7. Proliferation Resistant Nuclear Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Gray, L W; Moody, K J; Bradley, K S; Lorenzana, H E

    2011-02-18

    Global appetite for fission power is projected to grow dramatically this century, and for good reason. Despite considerable research to identify new sources of energy, fission remains the most plentiful and practical alternative to fossil fuels. The environmental challenges of fossil fuel have made the fission power option increasingly attractive, particularly as we are forced to rely on reserves in ecologically fragile or politically unstable corners of the globe. Caught between a globally eroding fossil fuel reserve as well as the uncertainty and considerable costs in the development of fusion power, most of the world will most likely come to rely on fission power for at least the remainder of the 21st century. Despite inevitable growth, fission power faces enduring challenges in sustainability and security. One of fission power's greatest hurdles to universal acceptance is the risk of potential misuse for nefarious purposes of fissionable byproducts in spent fuel, such as plutonium. With this issue in mind, we have discussed intrinsic concepts in this report that are motivated by the premise that the utility, desirability, and applicability of nuclear materials can be reduced. In a general sense, the intrinsic solutions aim to reduce or eliminate the quantity of existing weapons usable material; avoid production of new weapons-usable material through enrichment, breeding, extraction; or employ engineering solutions to make the fuel cycle less useful or more difficult for producing weapons-usable material. By their nature, these schemes require modifications to existing fuel cycles. As such, the concomitants of these modifications require engagement from the nuclear reactor and fuel-design community to fully assess their effects. Unfortunately, active pursuit of any scheme that could further complicate the spread of domestic nuclear power will probably be understandably unpopular. Nevertheless, the nonproliferation and counterterrorism issues are paramount

  8. Nuclear reactor technology: the next 50 years

    International Nuclear Information System (INIS)

    In light of the growing awareness of the environmental externalities of fossil fuel combustion, alternatives for electric power generation such as solar, wind and nuclear energy are becoming more desirable. In developed countries, large power markets are currently served by a centralized energy system through well inter-connected electricity grids. However, as shares of variable renewable energy sources (mainly wind and solar power) are increasing in the future; larger fluctuation in power generation can be expected which lead to higher risk of grid instabilities. Less-capital intensive small and medium sized nuclear reactors (SMR) are emerging as an important element of alternative power generation system to fossil fuel, with a unique additional role of balancing the power generation fluctuation caused by the solar and wind power generation. In regions not served by large electricity grids, including many parts of the developing countries with increasing demand for energy at rates above world's average, power generation using locally available energy sources including renewable energy is the practical means of providing basic energy needed for social and economic development. The integration of locally supportable SMR and local renewable energy system in a hybrid fashion can reduce the relative scale but not eliminate the fluctuation in power generation caused by the irregular availability of solar and wind energy. Without the use of commercial electricity trading that is only available in regions served by large inter-connected electricity grids, further minimization of power generation fluctuation can be done by the installation of local energy (electricity and/or heat) applications and/or energy storage device. The operation of these applications and energy storage can be done in synchronization with the availability of excess power throughout the fluctuation of the overall power generation in the region. Under these conditions, SMRs utilization as part of

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

    International Nuclear Information System (INIS)

    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

  10. Technology, Safety and Costs of Decommissioning Nuclear Reactors At Multiple-Reactor Stations

    Energy Technology Data Exchange (ETDEWEB)

    Wittenbrock, N. G.

    1982-01-01

    Safety and cost information is developed for the conceptual decommissioning of large (1175-MWe) pressurized water reactors (PWRs) and large (1155-MWe) boiling water reactors {BWRs) 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 nuclear 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. Five scenarios for decommissioning reactors at a multiple-reactor station are investigated. The number of reactors on a site is assumed to be either four or ten; nuclear waste disposal is varied between immediate offsite disposal, interim onsite storage, and immediate onsite disposal. It is assumed that the decommissioned reactors are not replaced in one scenario but are replaced in the other scenarios. Centralized service facilities are provided in two scenarios but are not provided in the other three. Decommissioning of a PWR or a BWR at a multiple-reactor station probably will be less costly and result in lower radiation doses than decommissioning an identical reactor at a single-reactor station. Regardless of whether the light water reactor being decommissioned is at a single- or multiple-reactor station: • the estimated occupational radiation dose for decommissioning an LWR is lowest for SAFSTOR and highest for DECON • the estimated

  11. 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

  12. Proceedings of the International Conference Nuclear Energy in Central Europe 2001

    International Nuclear Information System (INIS)

    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

  13. Nuclear reactor steam depressurization valve

    International Nuclear Information System (INIS)

    This patent describes improvement in a nuclear reactor plant, an improved steam depressurization valve positioned intermediate along a steam discharge pipe for controlling the venting of steam pressure from the reactor through the pipe. The improvement comprises: a housing including a domed cover forming a chamber and having a partition plate dividing the chamber into a fluid pressure activation compartment and a steam flow control compartment, the valve housing being provided with an inlet connection and an outlet connection in the steam flow control compartment, and a fluid duct in communication with a source of fluid pressure for operating the valve; a valve set mounted within the fluid flow control compartment comprising a cylindrical section surrounding the inlet connection with one end adjoining the connection and having a radially projecting flange at the other end with a contoured extended valve sealing flange provided with an annular valve sealing member, and a valve cylinder traversing the partition plate and reciprocally movable within an opening in the partition plate with one terminal and extending into the fluid pressure activation compartment and the other terminal end extending into the steam flow control compartment coaxially aligned with the valve seat surrounding the inlet connection, the valve cylinder being surrounded by two bellow fluid seals and provided with guides to inhibit lateral movement, an end of the valve cylinder extending into the fluid flow control compartment having a radially projecting flange substantially conterminous with the valve seat flange and having a contoured surface facing and complimentary to the contoured valve seating surface whereby the two contoured valve surfaces can meet in matching relationship, thus providing a pressure actuated reciprocatable valve member for making closing contact with the valve seat and withdrawing therefrom for opening fluid flow through the valve

  14. Atucha-2 obliquely inserted control rods RELAP5-3D/NESTLE model

    International Nuclear Information System (INIS)

    Atucha-2 is a Siemens-designed PHWR reactor in phase of commissioning in the Republic of Argentina. Its geometrical complexity and peculiarity (e.g., oblique control rods, positive void coefficient) required a developed and validated complex three dimensional (3D) neutron kinetics (NK) model. In the framework of the agreement between NASA and University of Pisa a detailed NESTLE (three-dimensional neutron kinetics code) model of the Atucha-2 NPP was developed. This document summarizes the procedures for the implementation of the oblique control rods into the RELAP5/NESTLE model: a particular arrangement of RELAP5/NESTLE control rods insertion mode for such kind of oblique control rods and an implementation into the homogenized two group cross sections of ad-hoc calculated correction factors (these parameters were obtained by previously executed Monte Carlo calculations) was developed. Some applications, among the scenarios selected to perform safety analysis of the Atucha-2 NPP (CNA-II), are also reported: preliminary Scram Rod Worth, analysis of a Control Rod Ejection Accidents and a CR faulty withdrawal. (author)

  15. 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.

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

    International Nuclear Information System (INIS)

    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.

  17. Proliferation resistance features in nuclear reactor designs

    International Nuclear Information System (INIS)

    Full text: The presentation gives an overview of the fundamental principles of non-proliferation of nuclear materials and technologies in the process of designing the nuclear reactors. The nuclear power engineering includes the activities involving the risk of proliferation of nuclear weapons (such as separation of uranium isotopes (enrichment), long-term storage of irradiated fuel, reprocessing of irradiated fuel by means of separation of plutonium and/or uranium wherefrom, storage of separated fissile materials). Proliferation resistance can be defined as the characteristic of a given nuclear power system which would prevent change-over or unauthorized production and use of the nuclear materials or technologies intended to possession of nuclear weapons or other nuclear explosives. The basic principles of non-proliferation as formulated in the frame of IAEA-sponsored international project INPRO have been analyzed for their relevance in designing the innovative nuclear power systems based on lead-cooled fast reactors. (author)

  18. Proliferation resistance features in nuclear reactor designs

    International Nuclear Information System (INIS)

    The paper presents a review of the main principles for technologies and materials protection from unauthorized proliferation and application to be considered in nuclear reactors designing. Nuclear power features certain operations sensitive to nuclear weapons proliferation (such as separation of uranium isotopes (enrichment), long storage of spent fuel, processing of spent fuel, plutonium and/or uranium recovery from spent fuel, storage of recovered fissile materials). Proliferation resistance is defined as a nuclear energy system characteristic that impedes the diversion or undeclared production of nuclear material, or misuse of technology with the purpose of acquiring nuclear weapons or other nuclear explosive devices. The basic principles of non-proliferation established in the INPRO international project sponsored by IAEA have been discussed as implemented for designing of the innovative nuclear energy systems based on fast lead-cooled nuclear reactors

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

    International Nuclear Information System (INIS)

    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

  20. Thermodynamic study of residual heat from a high temperature nuclear reactor to analyze its viability in cogeneration processes; Estudio termodinamico del calor residual de un reactor nuclear de alta temperatura para analizar su viabilidad en procesos de cogeneracion

    Energy Technology Data Exchange (ETDEWEB)

    Santillan R, A.; Valle H, J.; Escalante, J. A., E-mail: santillanaura@gmail.com [Universidad Politecnica Metropolitana de Hidalgo, Boulevard acceso a Tolcayuca 1009, Ex-Hacienda San Javier, 43860 Tolcayuca, Hidalgo (Mexico)

    2015-09-15

    In this paper the thermodynamic study of a nuclear power plant of high temperature at gas turbine (GTHTR300) is presented for estimating the exploitable waste heat in a process of desalination of seawater. One of the most studied and viable sustainable energy for the production of electricity, without the emission of greenhouse gases, is the nuclear energy. The fourth generation nuclear power plants have greater advantages than those currently installed plants; these advantages have to do with security, increased efficiencies and feasibility to be coupled to electrical cogeneration processes. In this paper the thermodynamic study of a nuclear power plant type GTHTR300 is realized, which is selected by greater efficiencies and have optimal conditions for use in electrical cogeneration processes due to high operating temperatures, which are between 700 and 950 degrees Celsius. The aim of the study is to determine the heat losses and the work done at each stage of the system, determining where they are the greatest losses and analyzing in that processes can be taken advantage. Based on the study was appointed that most of the energy losses are in form of heat in the coolers and usually this is emitted into the atmosphere without being used. From the results a process of desalination of seawater as electrical cogeneration process is proposed. This paper contains a brief description of the operation of the nuclear power plant, focusing on operation conditions and thermodynamic characteristics for the implementation of electrical cogeneration process, a thermodynamic analysis based on mass and energy balance was developed. The results allow quantifying the losses of thermal energy and determining the optimal section for coupling of the reactor with the desalination process, seeking to have a great overall efficiency. (Author)

  1. Calculation models for a nuclear reactor

    International Nuclear Information System (INIS)

    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)

  2. 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).

  3. Nuclear power in eastern and central Europe. Background paper

    International Nuclear Information System (INIS)

    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

  4. Nuclear data needs for fusion reactors

    International Nuclear Information System (INIS)

    The nuclear design of fusion reactor components (e.g., first wall, blanket, shield, magnet, limiter, divertor, etc.) requires an accurate prediction of the radiation field, the radiation damage parameters, and the activation analysis. The fusion nucleonics for these tasks are reviewed with special attention to point out nuclear data needs and deficiencies which effect the design process. The main areas included in this review are tritium breeding analyses, nuclear heating calculations, radiation damage in reactor components, shield designs, and results of uncertainty analyses as applied to fusion reactor studies. Design choices and reactor parameters that impact the neutronics performance of the blanket are discussed with emphasis on the tritium breeding ratio. Nuclear data required for kerma factors, shielding analysis, and radiation damage are discussed. Improvements in the evaluated data libraries are described to overcome the existing problems

  5. Autonomous Control of Space Nuclear Reactors Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Nuclear reactors to support future lunar and Mars robotic and manned missions impose new and innovative technological requirements for their control and protection...

  6. A swivelling transfer device for nuclear reactors

    International Nuclear Information System (INIS)

    The invention relates to a swivelling transfer device for fuel-assemblies. According to the invention, the device comprises, within a protective enclosure, a swivelling system comprising two sets of rails rotatable about an axis and so arranged that the lower and thereof penetrates into the extensions of the extremities of ramps dipped into the reactor and into a storage enclosure. This can apply to the transfer of nuclear reactor fuel assemblies, in particular for reactors of the molten sodium fast neutron type

  7. Nuclear data requirements for fusion reactor shielding

    International Nuclear Information System (INIS)

    The nuclear data requirements for experimental, demonstration and commercial fusion reactors are reviewed. Particular emphasis is given to the shield as well as major reactor components of concern to the nuclear performance. The nuclear data requirements are defined as a result of analyzing four key areas. These are the most likely candidate materials, energy range, types of needed nuclear data, and the required accuracy in the data. Deducing the latter from the target goals for the accuracy in prediction is also discussed. A specific proposal of measurements is recommended. Priorities for acquisition of data are also assigned. (author)

  8. Nuclear data requirements for fusion reactor nucleonics

    International Nuclear Information System (INIS)

    Nuclear data requirements for fusion reactor nucleonics are reviewed and the present status of data are assessed. The discussion is divided into broad categories dealing with data for Fusion Materials Irradiation Test Facility (FMIT), D-T Fusion Reactors, Alternate Fuel Cycles and the Evaluated Data Files that are available or would be available in the near future

  9. Nuclear reactor vessels with sealable rotatable covers

    International Nuclear Information System (INIS)

    Liquid metal cooled nuclear reactor installations have a rotating cover with an annular blade or skirt which clips into an annular trough of mercury and forms a gas seal. The design is such that abnormal pressure occurring in the reactor vessel is balanced by an increase in level of the mercury. Also applicable to irradiated fuel storage. (U.K.)

  10. Reactor Physics and the Nuclear Fuel Cycle

    Directory of Open Access Journals (Sweden)

    Md Minhaj Ahmed

    2013-11-01

    Full Text Available Questions regarding the feasibility of fusion power are examined, taking into account fuel cycles and breeding reactions, energy balance and reactor conditions, approaches to fusion, magnetic confinement, magneto hydro dynamic instabilities, micro instabilities, and the main technological problems which have to be solved. Basic processes and balances in fusion reactors are considered along with some aspects of the neutronics in fusion reactors, the physics of neutral beam heating, plasma heating by relativistic electrons, radiofrequency heating of fusion plasmas, adiabatic compression and ignition of fusion reactors, dynamics and control of fusion reactors, and aspects of thermal efficiency and waste heat. Attention is also given to fission-fusion hybrid systems, inertial-confinement fusion systems, the radiological aspects of fusion reactors, design considerations of fusion reactors, and a comparative study of the approaches to fusion power. The nuclear fuel cycle, also called nuclear fuel chain, is the progression of nuclear fuel through a series of differing stages. It consists of steps in the front end, which are the preparation of the fuel, steps in the service period in which the fuel is used during reactor operation, and steps in the back end, which are necessary to safely manage, contain, and either reprocess or dispose of spent nuclear fuel. If spent fuel is not reprocessed, the fuel cycle is referred to as an open fuel cycle (or a once-through fuel cycle; if the spent fuel is reprocessed, it is referred to as a closed fuel cycle..

  11. Problems of nuclear reactor safety. Vol. 2

    International Nuclear Information System (INIS)

    Theses of proceedings of the 9 Topical Meeting on problems of nuclear power plant safety are presented. Reports include results of neutron-physical experiments carried out for reactor safety justification. Concepts of advanced reactors with improved safety are considered. Results of researches on fuel cycles are given too

  12. Nuclear reactor fuel elements charging tool

    International Nuclear Information System (INIS)

    To assist the loading of nuclear reactor fuel elements in a reactor core, positioning blocks with a pyramidal upper face charged to guide the fuel element leg are placed on the lower core plate. A carrier equipped with means of controlled displacement permits movement of the blocks over the lower core plate

  13. Large Scale Weather Control Using Nuclear Reactors

    CERN Document Server

    Singh-Modgil, M

    2002-01-01

    It is pointed out that controlled release of thermal energy from fission type nuclear reactors can be used to alter weather patterns over significantly large geographical regions. (1) Nuclear heat creates a low pressure region, which can be used to draw moist air from oceans, onto deserts. (2) Creation of low pressure zones over oceans using Nuclear heat can lead to Controlled Cyclone Creation (CCC).(3) Nuclear heat can also be used to melt glaciers and control water flow in rivers.

  14. Optical techniques for nuclear reactor inspection

    International Nuclear Information System (INIS)

    Optical inspection techniques available and relevant to the various stages of the life cycle of a nuclear reactor are briefly reviewed. Experience in the three main types of nuclear reactor of interest to the CEGB, Magnox, AGR and PWR, is discussed. Conventional optical systems and stereoscopic viewing systems are described together with specialized and novel techniques, mainly Marchwood Engineering Laboratory's developments, which have proved valuable in tackling a variety of inspection problems. (U.K.)

  15. RA Research nuclear reactor - Annual report 1987

    International Nuclear Information System (INIS)

    Annual report concerning the project 'RA research nuclear reactor' for 1987, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities

  16. Nuclear reactor vessel fuel thermal insulating barrier

    Science.gov (United States)

    Keegan, C. Patrick; Scobel, James H.; Wright, Richard F.

    2013-03-19

    The reactor vessel of a nuclear reactor installation which is suspended from the cold leg nozzles in a reactor cavity is provided with a lower thermal insulating barrier spaced from the reactor vessel that has a hemispherical lower section that increases in volume from the center line of the reactor to the outer extent of the diameter of the thermal insulating barrier and smoothly transitions up the side walls of the vessel. The space between the thermal insulating harrier and the reactor vessel forms a chamber which can be flooded with cooling water through passive valving to directly cool the reactor vessel in the event of a severe accident. The passive inlet valve for the cooling water includes a buoyant door that is normally maintained sealed under its own weight and floats open when the cavity is Hooded. Passively opening steam vents are also provided.

  17. Materials for generation-IV nuclear reactors

    International Nuclear Information System (INIS)

    Materials science and materials development are key issues for the implementation of innovative reactor systems such as those defined in the framework of the Generation IV. Six systems have been selected for Generation IV consideration: gas-cooled fast reactor, lead-cooled fast reactor, molten salt-cooled reactor, sodium-cooled fast reactor, supercritical water-cooled reactor, and very high temperature reactor. The structural materials need to resist much higher temperatures, higher neutron doses and extremely corrosive environment, which are beyond the experience of the current nuclear power plants. For this reason, the first consideration in the development of Generation-IV concepts is selection and deployment of materials that operate successfully in the aggressive operating environments expected in the Gen-IV concepts. This paper summarizes the Gen-IV operating environments and describes the various candidate materials under consideration for use in different structural applications. (author)

  18. Fixed bed suspended core nuclear reactor concept

    International Nuclear Information System (INIS)

    The fixed bed nuclear reactor (FBNR) is essentially a pressurized light water reactor having spherical fuel elements constituting a suspended reactor core at its lowest bed porosity. The principle features of the proposed reactor are that the concept is polyvalent, simple in design, may operate either as fixed or fluidized bed, have the core suspended contributing to inherent safety, passive cooling features of the reactor. The reactor is modular and has an integrated primary system utilizing either water, supercritical steam or helium gas as its coolant. Some of the advantages of the proposed reactor are being modular, low environmental impact, exclusion of severe accidents, short construction period, flexible adaptation to demand, excellent load following characteristics, and competitive economics. (orig.)

  19. Research nuclear reactor RA - Annual Report 1989

    International Nuclear Information System (INIS)

    Annual report concerning the project 'RA research nuclear reactor' for 1989, financed by the Serbian ministry of science is divided into two parts. First part is concerned with RA reactor operation and maintenance, which is the task of the Division for reactor engineering of the Institute for multidisciplinary studies and RA reactor engineering. Second part deals with radiation protection activities at the RA reactor which is the responsibility of the Institute for radiation protection. Scientific council of the Institute for multidisciplinary studies and RA reactor engineering has stated that this report describes adequately the activity and tasks fulfilled at the RA reactor in 1989. The scope and the quality of the work done were considered successful both concerning the maintenance and reconstruction, as well as radiation protection activities

  20. Nuclear data needs for fast reactors

    International Nuclear Information System (INIS)

    The nuclear data, i.e., the numerical information about every nuclide - especially those representing the probabilities of various nuclear interactions and of radioactivity - of interest in a nuclear fission reactor are among the most essential inputs to be known a priori, to the best possible accuracy, for the design of nuclear reactor. The nuclides of interest cover not just (1) the fuel nuclides, the containers, the coolant, the moderator (if any), etc., that are initially inserted, but also (2) the actinides, the fission products, etc. that would be produced from the moment the reactor goes into operation and (3) the decay products that are produced even while the reactor is shutdown. The nuclide-list is known to cover a few hundreds. The neutron-nuclear interaction cross-section data, required for a few tens of reactions, very sensitively depend on the nuclide species and the neutron energy. Hence the data requirement significantly varies between thermal and fast reactors. The present talk is intended to touch upon the kinds and forms of nuclear data needed in the design and analysis of fast reactors. The recent variants available in the databases and some inter-comparison results will also be presented. (author)

  1. Fast-acting nuclear reactor control device

    International Nuclear Information System (INIS)

    A fast-acting nuclear reactor control device is described for controlling a safety control rod within the core of a nuclear reactor, the reactor controlled by a reactor control system, the device comprising: a safety control rod drive shaft and an electromagnetic clutch co-axial with the drive shaft operatively connected to the safety control rod for driving and positioning the safety control rod within or without the reactor core during reactor operation, the safety rod being oriented in a substantially vertical position to allow the rod to fall into the reactor core under the influence of gravity during shutdown of the reactor; the safety control rod drive shaft further operatively connected to a hydraulic pump such that operation of the drive shaft simultaneously drives and positions the safety control rod and operates the hydraulic pump such that a hydraulic fluid is forced into an accumulator, filling the accumulator with oil for the storage and supply of primary potential energy for safety control rod insertion such that the release of potential energy in the accumulator causes hydraulic fluid to flow through the hydraulic pump, converting the hydraulic pump to a hydraulic motor having speed and power capable of full length insertion and high speed driving of the safety control rod into the reactor core; a solenoid valve interposed between the hydraulic pump and the accumulator, said solenoid valve being a normally open valve, actuated to close when the safety control rod is out of the reactor during reactor operation; and further wherein said solenoid opens in response to a signal from the reactor control system calling for shutdown of the reactor and rapid insertion of the safety control rod into the reactor core, such that the opening of the solenoid releases the potential energy in the accumulator to place the safety control rod in a safe shutdown position

  2. 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 ...

  3. Computerized reactor monitor and control for nuclear reactors

    International Nuclear Information System (INIS)

    The analysis of a computerized process control system developed by Transelektro-KFKI-Videoton (Hangary) for a twenty-year-old research reactor in Budapest and or a new one in Tajura (Libya) is given. The paper describes the computer hardware (R-10) and the implemented software (PROCESS-24K) as well as their applications at nuclear reactors. The computer program provides for man-machine communication, data acquisition and processing, trend and alarm analysis, the control of the reactor power, reactor physical calculations and additional operational functions. The reliability and the possible further development of the computerized systems which are suitable for application at reactors of different design are also discussed. (Sz.J.)

  4. Hafnium control rod for nuclear reactors

    International Nuclear Information System (INIS)

    This patent describes an improved control device for nuclear fission reactor having a core fissionable fuel is an assembly composed of fuel units grouped into spaced apart bundles which are immersed in liquid coolant in operating service, and wherein the control device is provided with means for reciprocal movement into and out from the core of fuel intermediate the spaced apart bundles of the assembly. The control device having a frame including an upper and lower support member connected by an elongated central spine support with the upper and lower support members each of transverse cruciform configuration having four radially extending arms projecting outward from the central spine support and a sheath of bladelike configuration extending from each of the arms of the upper support to each of the arms of the lower support and each of the sheaths longitudinally adjoining the spine support. Each sheath containing therein a neutron absorbing component consisting essentially of a plurality of parallel flattened hollow tubes of hafnium metal with their axis aligned with the spine support, and each sheath. Each of the plurality of parallel flattened hollow tubes therein in their flattened portions are provided with a multiplicity of openings along their length for entry and the presence of liquid coolant through the length of the flattened hollow tubes of hafnium

  5. Distance to realization of nuclear fusion reactors

    International Nuclear Information System (INIS)

    Recently the research and development of nuclear fusion have progressed conspicuously, and reached the point of attaining the critical condition. In this paper, it is attempted to forecast how long does it take to realize a final nuclear fusion power reactor (a demonstration reactor). The research and development of nuclear fusion have two important meanings. One is it is a promising means for ensuring an energy source for the future in Japan. Another is it has been brought up to the present status as the large scale project research maintaining the creativity and originality without requiring the introduction of technology from foreign countries. Hereafter, it is necessary to bring it up large as the Japanese basic technology. The research and development of nuclear fusion has advanced steadily, producing many physical knowledges and technical development. The principle and present status of nuclear fusion are explained. Now, an experimental fusion reactor is investigated as the next step. A large helical system project was started as 7-year project from 1990. The start of the operation of a prototype nuclear fusion power reactor is assumed in 2026, and that of a demonstration reactor is assumed in 2040. The investment for nuclear fusion and the extending effect are discussed. (K.I.)

  6. Application of the SPA in the design of a hydrogen producer plant coupled to a nuclear reactor; Aplicacion del APS en el diseno de una planta productora de hidrogeno acoplada a un reactor nuclear

    Energy Technology Data Exchange (ETDEWEB)

    Ruiz S, T.; Nelson, P. F.; Francois, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Paseo Cuauhnahuac No. 8532, Col. Progreso, 62550 Jiutepec, Morelos (Mexico); Cruz G, M. J., E-mail: truizsmx@yahoo.com.mx [UNAM, Facultad de Quimica, Ciudad Universitaria, 04510 Mexico D. F. (Mexico)

    2013-10-15

    At the present time, one of the processes that is broadly investigated and that, theoretically demonstrates to be one of the most efficient for the hydrogen production, is the thermal-chemistry cycle Sulfur-Iodine (S-I) coupled to a nuclear reactor of very high temperature (VHTR). Because this chemical process of hydrogen production requires of a great inventory of toxic materials (sulphide compounds, hydriodic acid and iodine), is necessary the design of emergency systems with the purpose of protecting the facilities and the equipment s, the environment, as well as the near population. Given the impact of an accidental liberation of the process materials, as well as the proximity with the nuclear plant, is necessary that these emergency systems are the most reliable possible. This way, the results of the consequences analysis are utilized for the optimal localization of the gas sensors that activate the emergency systems, and the flows of the substances that are used for the leakage control. For all this, the use of the Safety Probabilistic Analysis methodology, as well as some standards of the nuclear industry, can be applied to the chemical installation to determine the fault sequences that can take to final states of not controlled leakage. This way, the use of methodologies of Event Tree Analysis and Fault Trees show in their results the components that but contribute in fault of such systems. In this work, is presented the evaluation of the joined models of event and fault trees and like with the obtained results, some proposals to increase the safety of the facilities are exposed. Also, the results of the evaluations of these proposals, and their impact of the probability of the not controlled fault sequences in a plant that is still in design stage are showed. (Author)

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

    International Nuclear Information System (INIS)

    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

  8. International ENEA/ISMES/ENS specialist meeting on 'On-site experimental verification of the seismic behaviour of nuclear reactor structures and components'. Proceedings

    International Nuclear Information System (INIS)

    The seismic verification of nuclear plants is a subject of increasing interest in all the industrial countries, with respect to both the safety aspects and the impact of the seismic event on the design and the costs of a nuclear reactor. This topic is especially of great interest for a country like Italy, whose territory is unfortunately characterized by non - negligible seismicity: we remember, not too many years ago, the catastrophic earthquakes of Frioul and Irpinia, that caused thousands of dead people. The meeting aimed at establishing the state-of-the-art on on-site testing of nuclear reactors structures and components, with particular attention to experiences and research programmes concerning: methodologies of on-site tests and interpretation of the experimental data; seismic monitoring systems, recorded data, their use and interpretation; calibration and validation of numerical analyses. Six technical sessions were held, during which 23 high papers were presented and discussed, and six panel discussions were held (the importance of discussion was emphasized in the meeting). The technical contributions consisted of: an introduction paper, summarizing the seismic studies performed in Italy for PEC reactor and explaining the reasons why on-site tests had been performed on this reactor; 6 invited lectures, one for each of the countries that are more deeply involved in seismic analysis, providing the state-of-the-art on the topics of interest for the meeting; 16 contributed papers dealing with more specific technical items, related to the various countries and international organizations

  9. Engineering and maintenance applied to safety-related valves in nuclear power plants; Ingenieria y mantenimiento aplicado a valvulas relacionadas con la seguridad en centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Verdu, M. F.; Perez-Aranda, J.

    2014-04-01

    Nuclear Division in Iberdrola engineering and Construction has a team with extensive experience on engineering and services works related to valves. Also, this team is linked to UNESA as Technical support and Reference Center. Iberdrola engineering and construction experience in nuclear power plants valves, gives effective response to engineering and maintenance works that can be demanded in a nuclear power plant and it requires a high degree of qualification and knowledge both in Operation and Outages. (Author)

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

    International Nuclear Information System (INIS)

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

  11. Nuclear data usage for research reactors

    International Nuclear Information System (INIS)

    In the department of research reactor, many neutronics calculations have been performed to construct, to operate and to modify research reactors of JAERI with several kinds of nuclear data libraries. This paper presents latest two neutronic analyses on research reactors. First one is design work of a low enriched uranium (LEU) fuel for JRR-4 (Japan Research Reactor No.4). The other is design of a uranium silicon dispersion type (silicide) fuel of JRR-3M (Japan Research Reactor No.3 Modified). Before starting the design work, to estimate the accuracy of computer code and calculation method, experimental data are calculated with several nuclear data libraries. From both cases of calculations, it is confirmed that JENDL-3.2 gives about 1 %Δk/k higher excess reactivity than JENDL-3.1. (author)

  12. Small reactors and the 'second nuclear era'

    International Nuclear Information System (INIS)

    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)

  13. Nuclear Power from Fission Reactors. An Introduction.

    Science.gov (United States)

    Department of Energy, Washington, DC. Technical Information Center.

    The purpose of this booklet is to provide a basic understanding of nuclear fission energy and different fission reaction concepts. Topics discussed are: energy use and production, current uses of fuels, oil and gas consumption, alternative energy sources, fossil fuel plants, nuclear plants, boiling water and pressurized water reactors, the light…

  14. Analysis of the evolution of the collective dose in nuclear power plants in Spain; Analisis de la evolucion de la dosis colectiva en las centrales nucleares de Espana

    Energy Technology Data Exchange (ETDEWEB)

    Ponjuan Reyes, G.; Ruibia Rodiz, M. A. de la; Rosales Calvo, M.; Labarta Mancho, T.; Calavia Gimenez, I.

    2011-07-01

    This article presents an analysis of the evolution of occupational collective dose of the Spanish nuclear power plants during the period 2000 - 2008 within the international context, by the Nuclear Safety Council (CSN) in order to have information contrasted to assessing the extent of application of the ALARA criteria in the Spanish plants and identify areas of priority attention.

  15. 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...

  16. Safety review, assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    In 1998, the NNSA organized to complete the nuclear safety review on the test loop in-reactor operation of the High-flux Engineering Experimental Reactor (HFEER) and the re-operation of the China Pulsed Reactor and the Uranium-water Criticality Facility. The NNSA conducted the nuclear safety review on the CP application of the China Experimental Fast Reactor (CEFR) and the siting of China Advanced Research Reactor (CARR), and carried out the construction supervision on HTR-10, and dealt with the event about the technological tube breakage of HWRR and other events

  17. Cold nuclear fusion reactor and nuclear fusion rocket

    OpenAIRE

    Huang Zhenqiang

    2013-01-01

    "Nuclear restraint inertial guidance directly hit the cold nuclear fusion reactor and ion speed dc transformer" [1], referred to as "cold fusion reactor" invention patents, Chinese Patent Application No. CN: 200910129632.7 [2]. The invention is characterized in that: at room temperature under vacuum conditions, specific combinations of the installation space of the electromagnetic field, based on light nuclei intrinsic magnetic moment and the electric field, the first two strings of the nucle...

  18. News on the natural nuclear reactor

    International Nuclear Information System (INIS)

    Data characterizing conditions of occurrence and the status of a natural nuclear reactor the remnants of which are discovered in the ore open pit of the Oklo deposit (Gabon) are presented. Transport of alkali earth elements (Rb, Sr, Cs and Ba) as well as Pd, Ag, Cd and Te isotopes near the reactor was investigated. Reactor criticality arose, probably, during or soon after U deposition. The reactor has ceased after 500000 years of operation; the energy of about 15 GW x year was generated. Approximately 80 t of uranium (12 tons of sup(235)U) were utilized during reactor operation. Approximately 10 tons of fission products and 4 tons of sup(239)Pu were formed. Reactor operation was periodical, multiply repeated. Water migrating over sandstone pores was not only a moderator but a self-regulator as well

  19. Desalination of seawater with nuclear reactors

    International Nuclear Information System (INIS)

    About 40 % of the world population is concerned with water scarcity. This article reviews the different techniques of desalination: distillation (MED and MSF), reverse osmosis (RO), and electrodialysis (ED). The use of nuclear energy rests on several arguments: 1) it is economically efficient compared to fossil energy. 2) nuclear reactors provide heat covering a broad range of temperature, which allows the implementation of all the desalination techniques. 3) the heat normally lost at the heat sink could be used for desalination. And 4) nuclear is respectful of the environment. The feedback experience concerning nuclear desalination is estimated to about 100 reactor-years, it is sufficient to allow the understanding of all the physical and technological processes involved. In Japan, 8 PWR-type reactors are coupled to MED, MSF, and RO desalination techniques, the water produced is used locally mainly for feeding steam generators. (A.C.)

  20. Desalination of seawater with nuclear reactors

    International Nuclear Information System (INIS)

    About 40 % of the world population is concerned with water scarcity. This article reviews the different techniques of desalination: distillation (MED and MSF), reverse osmosis (RO), and electrodialysis (ED). The use of nuclear energy rests on several arguments: 1) it is economically efficient compared to fossil energy; 2) nuclear reactors provide heat covering a broad range of temperature, which allows the implementation of all the desalination techniques; 3) the heat normally lost at the heat sink could be used for desalination; and 4) nuclear is respectful of the environment. The feedback experience concerning nuclear desalination is estimated to about 100 reactor-years, it is sufficient to allow the understanding of all the physical and technological processes involved. In Japan, 8 PWR-type reactors are coupled to MED, MSF, and RO desalination techniques, the water produced is used locally mainly for feeding steam generators. (A.C.)

  1. Nuclear data needs for fusion reactors

    International Nuclear Information System (INIS)

    The nuclear design of fusion components (e.g., first wall, blanket, shield, magnet, limiter, divertor, etc.) requires an accurate prediction of the radiation field, the radiation damage parameters, and the activation analysis. The fusion nucleonics for these tasks are reviewed with special attention to point out nuclear data needs and deficiencies which effect the design process. The main areas included in this review are tritium breeding analyses, nuclear heating calculations, radiation damage in reactor components, shield designs, and results of uncertainty analyses as applied to fusion reactor studies. Design choices and reactor parameters that impact the neutronics performance of the blanket are discussed with emphasis on the tritium breeding ratio. Nuclear data required for kerma factors, shielding analysis, and radiation damage are discussed. Improvements in the evaluated data libraries are described to overcome the existing problems. 84 refs., 11 figs., 9 tabs

  2. Monitoring Nuclear Reactors with Antineutrino Detectors: The ANGRA Project

    International Nuclear Information System (INIS)

    Nuclear reactors are intense source of antineutrinos and the thermal power released in the fission process is directly related to the emitted antineutrino flux. As antineutrinos interact very weekly with matter and escape the reactor containment without any significant change in their number, measuring the antineutrino flux can give quasi real time information on the reactor status (on/off) and thermal power. Furthermore as the amount of fissile elements change during the burn up of the nuclear fuel and each fissile isotope has a characteristic antineutrino energy spectrum, measuring the antineutrino energy spectrum escaping the reactor can give also information on the composition of the nuclear fuel. These unique features make antineutrino detectors good candidates to become in the near future a new safeguards tool for monitoring nuclear reactors through counting rates and spectral antineutrino measurements. We will present the status of the ANGRA Neutrinos project, aimed at developing an antineutrino detector for monitoring nuclear reactor activity. The Angra experiment will be deployed at the Brazilian nuclear power plant Angra II. A water Cherenkov detector of one ton target will be placed in a commercial container just outside the reactor containment, about 30 m of the reactor core. The 4 GW thermal power of the Angra II reactor will provide a few thousand antineutrino inverse beta decay interactions per day. The main challenge of the experiment will be to overcome the very high cosmic ray induced background at sea level, consisting mainly of muons, neutrons, gammas and electrons. We have estimated a muon flux of the order of 350 Hz and a neutron flux of about 4 Hz (left after the polyethylene shielding) in the target region. To overcome the background the detector will consist of three subsystems: 1) a muon veto 99% efficient placed in the outer most detector layer; 2) a neutron shield 30cm thick consisting of polyethylene sheets; 3) a central detector

  3. Evaluation of fatigue damage in nuclear power plants: evolution and new tools of analysis; Evaluacion del dano a fatiga en centrales nucleares: evolucion y nuevas herramientas de analisis

    Energy Technology Data Exchange (ETDEWEB)

    Cicero, R.; Corchon, F.

    2011-07-01

    This paper presents new fatigue mechanisms requiring analysis, tools developed for evaluation and the latest trends and studies that are currently working in the nuclear field, and allow proper management referring facilities the said degradation mechanism.

  4. Evaluation of the radiological impact due to the operation of nuclear power stations in Argentina

    International Nuclear Information System (INIS)

    There are at present in Argentina two commercial nuclear power stations in operation, Atucha I and Embalse, generating about 10% of the total electrical energy output. Atucha I NPP, equipped with a pressure vessel reactor, has an output capacity of 345 MW(e), while Embalse NPP is equipped with Candu-type reactor and its output capacity is 670 MW(e). Both plants operate with natural uranium as a fuel, and heavy water as a coolant and as a moderator. During the operation of both nuclear installations, radioactive fission and activation products are produced. These radioactive materials are for the most part retained within the fuel elements. Most of the radionuclides with diffuse into or are formed within the coolant are removed by the gaseous and liquid waste processing systems. Low-level releases which occur during normal operation are controlled and monitored. Radionuclides may reach the environment through either the gaseous or liquid effluents streams

  5. IGALL, key factor in long-term operation of nuclear power plants; IGALL, factor clae en la operacion a largo plazo de centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Calatayud, M. A.; Sainero, J.

    2014-10-01

    The ageing management is a key factor during the Long Term Operation of Nuclear Power Plants and also during the licensed period; IGALL is an international accepted reference guide that shows how the manage it. The knowledge of the reasons to develop this project, how the results have been reached and the public access to them for future use of the Nuclear Power Plants, are the purpose of this article. (Author)

  6. Features of a subcritical nuclear reactor

    International Nuclear Information System (INIS)

    Highlights: • The keff was calculated using six factor formula and MCNP code. • Both methods agree when the reactor is loaded from 800 to 1900 kg. • With the MCNP5 code the neutron spectra and doses were estimated. • The Ambient dose was measured outside the subcritical assembly. - Abstract: A subcritical nuclear reactor is a device where the nuclear-fission chain reaction is initiated and maintained using an external neutron source. It is a valuable educational and research tool where in a safe way many reactor parameters can be measured. Here, we have used the six-factor formula to calculate the effective multiplication factor of a subcritical nuclear reactor Nuclear Chicago model 9000. Using the MCNP5 code, a three-dimensional model of the subcritical reactor was developed to estimate the effective multiplication factor, the neutron spectra, and the total and thermal neutron fluences along the radial and axial axis. The MCNP5 results of the effective multiplication factor were compared with those obtained from the six-factor formula. The effective dose and the Ambient dose equivalent, at three sites outside the reactor, were estimated; the Ambient dose equivalent was also measured and compared with the calculated values

  7. Feedback of reactor operating data to nuclear methods development

    International Nuclear Information System (INIS)

    The problems in obtaining power reactor data for reliable nuclear methods development and the major sources of power reactor data for this purpose are reviewed. Specific examples of the use of power reactor data in nuclear methods development are discussed. The paper concludes with recommendations on the key elements of an effective program to use power reactor data in nuclear methods development

  8. Nuclear insurance in Central and Eastern Europe

    International Nuclear Information System (INIS)

    In the world outside the former Soviet Union, insurance industries in their respective domestic markets have pooled their resources so as to provide a secure and cost-effective conduit for the transaction of insurance business on behalf of the nuclear industry. These are the so-called nuclear pools. This paper explains the four main principles behind nuclear liability insurance and discusses their application to Central Europe and in particular to the problems facing the nuclear industry in Eastern Europe. (author)

  9. Optimally moderated nuclear fission reactor and fuel source therefor

    Science.gov (United States)

    Ougouag, Abderrafi M.; Terry, William K.; Gougar, Hans D.

    2008-07-22

    An improved nuclear fission reactor of the continuous fueling type involves determining an asymptotic equilibrium state for the nuclear fission reactor and providing the reactor with a moderator-to-fuel ratio that is optimally moderated for the asymptotic equilibrium state of the nuclear fission reactor; the fuel-to-moderator ratio allowing the nuclear fission reactor to be substantially continuously operated in an optimally moderated state.

  10. Design of a nuclear reactor cooperative controller

    International Nuclear Information System (INIS)

    This paper describes the development of a fuzzy logic controller software package and explores the feasibility of its use in nuclear reactor operation. The controller complements reactor operator actions, and the operators can override the controller decisions. Techniques of providing learning capability to the controller are also being investigated to improve the reasoning and control skill of the controller. The fuzzy logic controller is implemented in C language and its overall structure is shown. The heart of the systems consists of a fuzzifier, a rule interpreter, and a defuzzifier. The controller is designed as a stand-alone package that can be interfaced to a simulated model of a nuclear reactor. Since no model is an accurate representation of the actual process being modeled, some tuning must be performed to use the controller in an actual reactor. This is accomplished using the learning feature of the controller

  11. La política nuclear espanyola: el caos del reactor nuclear Argos

    OpenAIRE

    Barca Salom, Francesc Xavier

    2000-01-01

    L’11 de juny de 1962 s’inaugurava a l’Escola Tècnica Superior d’Enginyeria Industrial de Barcelona un reactor nuclear experimental, que era batejat amb el nom mític d’Argos. Tota la premsa barcelonina se’n feu ressò i el presentava com el primer reactor construït íntegrament a Espanya per la Junta d’Energia Nuclear. La idea de dotar l’Escola d’un reactor nuclear havia nascut, però, set anys abans, precisament en el mateix moment de la creació de la Càtedra Ferran Tallada d’enginyeria...

  12. Implementation of a new digital system in Atucha-II NPP

    International Nuclear Information System (INIS)

    Atucha-II is a delayed nuclear power plant project in Argentina. The Safety and Operational I and C were designed using modular electronic systems based on analog technologies of 1980 that presently are out of fabrication and no more available in the market. When restarting the project in 2006, it was found that some of the operational I and C modules were not supplied in sufficient quantity to complete the erection and the quantity of spares was not enough to assure the plant operation. To solve these problems it was decided to begin a modernization process of the central I and C in steps. The first step, which is described here and will be performed before the commissioning of the plant, consists in the change of part of the operational cubicles and the Supervisory Computer System to a new programmable digital system. The original design of the central I and C based in Functional Complexes makes the hardware migration easy to implement. Each functional complex is a group of cubicles in which all the I and C functions are located which are related to the monitoring and control of a given process system. By means of the replacing of some functional complexes to digital I and C, electronic modules can be obtained to complete the installation or to be used as spares. Also the supervisory computer will be changed to a new digital HIS (Human System Interface), which will assume all the annunciation and data logging functions of the old computer system and will also allow the monitoring and control functions of the new digital operational I and C. The new HSI system also can be connected by bus to digital electronic cubicles of the Safety I and C, which will replace the present analog ones in a future step. With this basic first step of modernization future partial I and C replacements during plant outages with the plant in operation will be easier. (author)

  13. Water shielding nuclear reactor container

    International Nuclear Information System (INIS)

    The reactor container of the present invention contains a reactor pressure vessel, and has double steel plate walls endurable to elevated inner pressure and keeping airtightness, and shielding water is filled inside from a water injection port. It is endurable to a great inner pressure satisfactorily and keep airtightness by the two spaced relatively thin steel plates. It exhibits radiation shielding effect by filling water substantially the same as that of a conventional reactor container made of iron reinforced concretes. Then, it is no more necessary to use concretes for the construction of the reactor container, which shortens the term of the construction, and saves the construction cost. In addition, a cooling effect for the reactor container is provided. Syphons are disposed contiguously to a water injection port and the top end of the syphon is immersed in an equipment temporarily storage pool, and further, pipelines are connected to the double steel plate walls or the syphons for supplying shielding water to enhance the cooling effect. (N.H.)

  14. Nuclear reactor power control device

    International Nuclear Information System (INIS)

    When occurrence of earthquakes is judged in a BWR type reactor, the power is decreased by inserting a portion of control rods, reducing a speed of recycling pumps, stopping recycling pumps, increasing the opening degree of a main steam control valve and opening a main steam relief valve. The reactor scram can be avoided by bypassing neutron flux high signal, settling a filter to neutron flux signals and setting a reactor scram set value by neutron flux signals, for example, to 120%. There is constituted an interlock for performing reactor scram when both of a neutron flux high signal and a signal outputted if a surface heat flux corresponding signal formed by applying calculation to the neutron flux high signal exceeds a set value are valid, to avoid unnecessary reactor scram. As a measuring means, not only an acceleration meter in the power plant, but also acceleration meters at remote places, acceleration meters or displacement meters for various kinds of equipments in the power plant are used, and when signals from them exceed set values, earthquake judgement is conducted. (N.H.)

  15. 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

  16. Nuclear reactor fissile isotopes antineutrino spectra

    OpenAIRE

    Sinev, V.

    2012-01-01

    Positron spectrum from inverse beta decay reaction on proton was measured in 1988-1990 as a result of neutrino exploration experiment. The measured spectrum has the largest statistics and lowest energy threshold between other neutrino experiments made that time at nuclear reactors. On base of the positron spectrum the standard antineutrino spectrum for typical reactor fuel composition was restored. In presented analysis the partial spectra forming this standard spectrum were extracted using s...

  17. The safety of Ontario's nuclear reactors

    International Nuclear Information System (INIS)

    A Select Committee of the Legislature of Ontario was established to examine the affairs of Ontario Hydro, the provincial electrical utility. Extensive public hearings were held on several topics including the safety of nuclear power reactors operating in Ontario. The Committee found that these reactors are acceptably safe. Many of the 24 recommendations in this report deal with the licensing process and public access to information. (O.T.)

  18. Nuclear Research Center IRT reactor dynamics calculation

    International Nuclear Information System (INIS)

    The main features of the code DIRT, for dynamical calculations are described in the paper. With the results obtained by the program, an analysis of the dynamic behaviour of the Research Reactor IRT of the Nuclear Research Center (CIN) is performed. Different transitories were considered such as variation of the system reactivity, coolant inlet temperature variation and also variations of the coolant velocity through the reactor core. 3 refs

  19. Liquid metal cooled nuclear reactor constructions

    International Nuclear Information System (INIS)

    A liquid metal cooled nuclear reactor construction is described comprising a reactor core submerged in a pool of liquid metal coolant contained in a vessel which is housed in a concrete containment vault, the roof structure of the vault having thermal insulation comprising a series of super-imposed spaced plates, with baffles disposed so as to restrict convectional flow of metal vapour through the interspaces of the plates and between the uppermost plate or plates and the vault roof structure. (author)

  20. Advances in seismic safety of operating nuclear power plants: IAEA Project Results Seismic-EBP; Avances en la seguridad sismica de centrales nucleares en operacion: Resultados del proyecto OIEA Seismic-EBP

    Energy Technology Data Exchange (ETDEWEB)

    Jimenez juan, A.; Sanchez Cabanero, J. G.; Moreno gonzalez, A.

    2010-07-01

    Records of strong earthquakes occurred in the proximity of facilities in operation have shown that there may be greater than the expected acceleration, the seismic danger that the light of the actual experience can exceed the design basis, and that the aging installations and modifications over its operating life significantly influence seismic response capacity. All this confirms that the earthquake is the external event with the greatest impact on the safety of nuclear installations and the largest contributor (even more than 70%) to the core damage frequency of central probabilistic safety analyzes.

  1. Modernization of the turbo in the Laguna Verde Nuclear Power Plant; Modernizacion del turbogrupo en la Central Nuclear de Laguna Verde

    Energy Technology Data Exchange (ETDEWEB)

    Liebana, B.; Merino, A.; Cobos, A.; Gonzalez, J. J.

    2010-07-01

    The power increase of the Laguna Verde Nuclear Power Plant is a project for the rehabilitation and modernization of the turbo and associated equipment to get an increase of its power and of its service life. The project scope includes the design, the engineering, the equipment supply, the installation, the testing and the commissioning. This paper describes the first phase of the project.

  2. Advanced nuclear reactor types and technologies

    Energy Technology Data Exchange (ETDEWEB)

    Ignatiev, V. [ed.; Feinberg, O.; Morozov, A. [Russian Research Centre `Kurchatov Institute`, Moscow (Russian Federation); Devell, L. [Studsvik Eco and Safety AB, Nykoeping (Sweden)

    1995-07-01

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary.

  3. Advanced nuclear reactor types and technologies

    International Nuclear Information System (INIS)

    The document is a comprehensive world-wide catalogue of concepts and designs of advanced fission reactor types and fuel cycle technologies. Two parts have been prepared: Part 1 Reactors for Power Production and Part 2 Heating and Other Reactor Applications. Part 3, which will cover advanced waste management technology, reprocessing and disposal for different nuclear fission options is planned for compilation during 1995. The catalogue was prepared according to a special format which briefly presents the project title, technical approach, development status, application of the technology, reactor type, power output, and organization which developed these designs. Part 1 and 2 cover water cooled reactors, liquid metal fast reactors, gas-cooled reactors and molten salt reactors. Subcritical accelerator-driven systems are also considered. Various reactor applications as power production, heat generation, ship propulsion, space power sources and transmutation of such waste are included. Each project is described within a few pages with the main features of an actual design using a table with main technical data and figure as well as references for additional information. Each chapter starts with an introduction which briefly describes main trends and approaches in this field. Explanations of terms and abbreviations are provided in a glossary

  4. Equipments for use in nuclear reactors

    International Nuclear Information System (INIS)

    Purpose: To prevent nickel from leaching into nuclear reactor coolants thereby suppress the generation of cobalt 58 due to activation. Constitution: Equipments for use in nuclear reactor cores such as control rod covering tubes, control rod sheaths, feedwater heater tubes, etc. are constituted with ferrite type stainless steels not containing nickel and containing more than 10 wt% chromium incorporated as the basic element with nitrogen at a level higher than the impurity content. While the ferrite type materials have been used for the pressure vessel, there has been a problem in view of the embrittlement under irradiation when used as the equipments in nuclear reactor cores. However, it has been found that the embrittlement under irradiation can be improved by the incorporation of nitrogen in excess of the impurity level. Accordingly, the material of the present invention can suppress the generation of cobalt 58, provide corrosion resistance and avoid embrittlement under irradiation. (Takahashi, M.)

  5. Nuclear reactor safety in the USA

    International Nuclear Information System (INIS)

    Nuclear reactor safety in the USA has emphasized a defense-in-depth approach to protecting the public from reactor accidents. This approach was severely tested by the Three Mile Island accident and was found to be effective in safeguarding the public health and safety. However, the economic impact of the TMI accident was very large. Consequently, more attention is now being given to plant protection as well as public-health protection in reactor-safety studies. Sophisticated computer simulations at Los Alamos are making major contributions in this area. In terms of public risk, nuclear power plants compare favorably with other large-scale alternatives to electricity generation. Unfortunately, there is a large gulf between the real risks of nuclear power and the present public perception of these risks

  6. Device for nuclear reactor control

    International Nuclear Information System (INIS)

    The device for power height distribution control in channel-type uranium-graphite reactor cores is described. The device is a water filled vertical channel positioned in the reactor core. The device consists of a controlling rod, displacer in a form of a throttle and gas cavity and discharge throttle. The rod is fixed in upper position with an electromagnet. By shifting a displacer and changing flow rate established are the required height and position of a controlling liquid column. In the emergency protection, a drive shifts the displacer under core space or the displacer drops under the action of its own weight at electromagnet clutch doenergyzation whereas the channel is filled by liquid. The application of the suggested device permits to improve economic and operating characteristics of reactors

  7. Fukushima, two years later, modification requirements in nuclear power plants; Fukushima, dos anos despues, requerimientos de modificacion en centrales nucleares de potencia

    Energy Technology Data Exchange (ETDEWEB)

    Sanchez J, J.; Camargo C, R.; Nunez C, A.; Mendoza F, J. E.; Salmeron V, J. A., E-mail: jerson.sanchez@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2013-10-15

    The occurred events in the nuclear power plant of Fukushima Daiichi as consequence of the strong earthquake of 9 grades in the Richter scale and the later tsunami with waves estimated in more than 14 meters high began a series of important questions about the safety of the nuclear power plants in operation and of the new designs. Firstly, have allowed to be questioned on the magnitudes and consequences of the extreme external natural events; that can put in risk the integrity of the safety barriers of a nuclear power plant when being presented in a multiple way. As consequence of the events of the Fukushima Daiichi NPP, the countries with NPPs in operation and /or construction carried out evaluations about their safety operation. They have also realized evaluations about accidents and their impact in the safety, analysis and studies too that have forced to the regulatory bodies to continue a systematic and methodical revision of their procedures and regulations, to identify the possible improvements to the safety in response to the events happened in Japan; everything has taken it to determine the necessity to incorporate additional requirements to the nuclear power plants to mitigate events Beyond the Design Base. Due to Mexico has the nuclear power plant of Laguna Verde, with two units of BWR-5 type with contention Mark III, some the modifications can be applicable to these units to administrate and/or to mitigate the consequences of the possible occurrence of an accident Beyond the Design Base and that could generate a severe accident. In this work an exposition is presented on the modification requirements to confront external natural events Beyond the Design Base, and its application in our country. (Author)

  8. Nuclear reactor fuel element splitter

    International Nuclear Information System (INIS)

    A method and apparatus are disclosed for removing nuclear fuel from a clad fuel element. The fuel element is power driven past laser beams which simultaneously cut the cladding lengthwise into at least two longitudinal pieces. The axially cut lengths of cladding are then separated, causing the nuclear fuel contained therein to drop into a receptacle for later disposition. The cut lengths of cladding comprise nuclear waste which is disposed of in a suitable manner. 6 claims, 10 drawing figures

  9. 78 FR 64028 - Decommissioning of Nuclear Power Reactors

    Science.gov (United States)

    2013-10-25

    ... COMMISSION Decommissioning of Nuclear Power Reactors AGENCY: Nuclear Regulatory Commission. ACTION... regulatory guide (RG) 1.184 ``Decommissioning of Nuclear Power Reactors.'' This guide describes a method NRC... decommissioning process for nuclear power reactors. The revision takes advantage of the 13 years...

  10. Regulatory actions towards dose reduction at Atucha I NPS

    International Nuclear Information System (INIS)

    Atucha I, 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% of cobalt for hard-facing applications). The erosion and corrosion processes on the surfaces of the piping and components of the primary coolant and moderator systems generate a varied type of particles oxides called 'crud'. The crud and 60Co 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 60Co 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 made by ICRP Publication 26. The recommendations included in ICRP Publication 60 made more than difficult the radiological situation at the Atucha I. 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 60Co. Some options to reduce individual and collective doses were analysed by the Authority. To carry out the evaluation of the deposit mechanisms and the actual activity level of 60Co, 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

  11. Regulatory actions towards dose reduction at Atucha 1 NPS

    International Nuclear Information System (INIS)

    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

  12. Analysis of the micro-structural damages by neutronic irradiation of the steel of reactor vessels of the nuclear power plant of Laguna Verde. Characterization of the design steel; Analisis de los danos micro-estructurales por irradiacion neutronica del acero de la vasija de los reactores de la Central Nuclear de Laguna Verde. Caracterizacion del acero de diseno

    Energy Technology Data Exchange (ETDEWEB)

    Moranchel y Rodriguez, M.; Garcia B, A. [IPN, Escuela Superior de Fisica y Matematicas, Departamento de Fisica, Av. Luis Enrique Erro s/n, Unidad Profesional Adolfo Lopez Mateos, Col. Lindavista, 07738 Mexico D. F. (Mexico); Longoria G, L. C., E-mail: mmoranchel@ipn.m [ININ, Direccion de Investigacion Cientifica, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2010-09-15

    The vessel of a nuclear reactor is one of the safety barriers more important in the design, construction and operation of the reactor. If the vessel results affected to the grade of to have fracture and/or cracks it is very probable the conclusion of their useful life in order to guarantee the nuclear safety and the radiological protection of the exposure occupational personnel, of the public and the environment avoiding the exposition to radioactive sources. The materials of the vessel of a nuclear reactor are exposed continually to the neutronic irradiation that generates the same nuclear reactor. The neutrons that impact to the vessel have the sufficient energy to penetrate certain depth in function of the energy of the incident neutron until reaching the repose or to be absorbed by some nucleus. In the course of their penetration, the neutrons interact with the nuclei, atoms, molecules and with the same crystalline nets of the vessel material producing vacuums, interstitial, precipitate and segregations among other defects that can modify the mechanical properties of the steel. The steel A533-B is the material with which is manufactured the vessel of the nuclear reactors of nuclear power plant of Laguna Verde, is an alloy that, among other components, it contains atoms of Ni that if they are segregated by the neutrons impact this would favor to the cracking of the same vessel. This work is part of an investigation to analyze the micro-structural damages of the reactor vessels of the nuclear power plant of Laguna Verde due to the neutronic irradiation which is exposed in a continuous way. We will show the characterization of the design steel of the vessel, what offers a comprehension about their chemical composition, the superficial topography and the crystalline nets of the steel A533-B. It will also allow analyze the existence of precipitates, segregates, the type of crystalline net and the distances inter-plains of the design steel of the vessel. (Author)

  13. Nuclear reactors with auxiliary boiler circuit

    International Nuclear Information System (INIS)

    A gas-cooled nuclear reactor has a main circulatory system for the gaseous coolant incorporating one or more main energy converting units, such as gas turbines, and an auxiliary circulatory system for the gaseous coolant incorporating at least one steam generating boiler arranged to be heated by the coolant after its passage through the reactor core to provide steam for driving an auxiliary steam turbine, such an arrangement providing a simplified start-up procedure also providing emergency duties associated with long term heat removal on reactor shut down

  14. IAEA role in nuclear reactor safety standardization

    International Nuclear Information System (INIS)

    In 1981 the electricity generation by nuclear power plants all over the world reached 8% of total production. It can be expected that at the turn of century up to 25% of electric power will be provided by means of nuclear fuel burning. In connection with NPP total number growth, their attraction to large population centres, widening of the atomic energy application areas, the importance of nuclearreactor safety problems can only increase. The safety measures have usually the structure of sequential barriers: for accident preventing, for protection from accidents, for accident localization. NPP safety is a complex problem having scientific, engineering, juridical, social and political aspects. Since these problems have an international importance, IAEA should actively work on their solving. Practically all the topics of nuclear power development and nuclear reactor s;fety lie within the activity area of the Department of nuclear power and safety, its sections: of nuclear safety, nuclear power, nuclear fuel cycle. In 1974 a decision was made in IAEA about initiation of work on development of an international nuclear safety standards system (NUSS Programme). These activities are divided into five major branches: a government organization for nuclear safety regulations; site selection for NPP; NPP desing; operation, start of operation and decommissioning; quality provision for NPP. The report presents a list of documents, comprising the NUSS Programme. The complection of all the works within the scope of the Programme is planned for 1985. After 1985 the start of development of fast neutron reactor and fuel cycle enterprise safety standards is planned

  15. Central Institute for Nuclear Research (1956 - 1979)

    International Nuclear Information System (INIS)

    The Central Institute for Nuclear Research (ZfK) of the Academy of Sciences of the GDR is presented. This first overall survey covers the development of the ZfK since 1956, the main research activities and results, a description of the departments responsible for the complex implementation of nuclear research, the social services for staff and the activities of different organizations in the largest central institute of the Academy of Sciences of the GDR. (author)

  16. Atucha I: NPP Spent Fuel Dry Storage Conceptual Design

    International Nuclear Information System (INIS)

    The present report shows a Spent Fuel Dry Storage Conceptual Design to emptying the former oldest fuel elements pool and storage them in a Dry Storage System, in order to reach the 32 (Full Power Years) FPY of Atucha I Nuclear Power Plant (CNA I) End Of Life [1]. The project consist mainly in the enlargement of one of the Pool Buildings of the Station (there are two of them) where the (Spent Fuel Elements) SFE will be stored in vertical underground silos. Each silo is composed of two storage units that contains 9 fuel elements each (18 fuel elements in each bin). This design allows a vertical storage of 2016 spent fuel elements (7 rows by 16 columns). The SFE must be transferred from the Pool Building to the Dry Storage Building through a dedicated shield for lifting and transporting the SFE. To move the shield, the actual 60 Ton capacity crane will be used. The operation time to emptying a complete pool will be approximately one year (1998 SFE). Therefore the storage system should be finished by 2013, in order not to penalize the continue operation of the Station. This conceptual design meets the basic principles of Nuclear Safety, protecting workers, public and the overall environment of ionizing radiation and radioactive contamination. This is achieved by transport and storage shielding, operation procedures and comply key conditions like subcriticality of the system, SFE monitoring and SFE heat removal. (author)

  17. Local AREA networks in advanced nuclear reactors

    International Nuclear Information System (INIS)

    The report assesses Local Area Network Communications with a view to their application in advanced nuclear reactor control and protection systems. Attention is focussed on commercially available techniques and systems for achieving the high reliability and availability required. A basis for evaluating network characteristics in terms of broadband or baseband type, medium, topology, node structure and access method is established. The reliability and availability of networks is then discussed. Several commercial networks are briefly assessed and a distinction made between general purpose networks and those suitable for process control. The communications requirements of nuclear reactor control and protection systems are compared with the facilities provided by current technology

  18. Establishment and evolution of Radiological Environmental Monitoring Programmes in nuclear power plants; Establecimiento y evolucion de los programas de vigilancia radiologica ambiental en las centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Marugan, I.; Luque, S.; Martin, J. L.; Rey, C.; Salas, R.; Sterling, A.; Ramos, L. M.

    2013-09-01

    This article presents a brief overview of how the Radiological Environmental Monitoring Programmes carried out around nuclear power plants have evolved associated to different reasons as the legal framework, operational phases of the facilities, development on the detection and measurement of low levels of radiation due to the state of art and best available technologies, changes within sites as well as in their surroundings and accident taken place inside and outside of our borders. (Author)

  19. Methods in nuclear reactors calculations

    International Nuclear Information System (INIS)

    Studies are made of the neutron transport equation corresponding to the the real and virtual reactors, as well as the starting hypotheses. Methods are developed to solve the transport equation in slab geometry, and Pl; Bl; Ml; Sn and discrete ordinates approximations. (Author)

  20. Advanced nuclear reactor systems - an Indian perspective

    International Nuclear Information System (INIS)

    The Indian nuclear power programme envisages use of closed nuclear fuel cycle and thorium utilisation as its mainstay for its sustainable growth. The current levels of deployment of nuclear energy in India need to be multiplied nearly hundred fold to reach levels of electricity generation that would facilitate the country to achieve energy independence as well as a developed status. The Indian thorium based nuclear energy systems are being developed to achieve sustainability in respect of fuel resource along with enhanced safety and reduced waste generation. Advanced Heavy Water Reactor and its variants have been designed to meet these objectives. The Indian High Temperature Reactor programme also envisages use of thorium-based fuel with advanced levels of passive safety features. (author)

  1. What they have in common the engineering from the Spanish nuclear power plants?; Que tienen en comun las ingenierias de las centrales nucleares espanolas

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez Mendez, M.

    2012-11-01

    In recent years, Spain Nuclear Power Plant Engineering have switched their project/task management method to Critical Chain multi-project management, developed by Dr. Goldratt, achieving outstanding results in improving quality and productivity. Multitasking reduction, task and resource synchronizing without the need of exact schedules, implementing a real-time priority information system, relying on the software Concerto, and daily decision making are the basis for the management change that has generated productivity increases of between 20% to 50%, opening new horizons for improvement in other scenarios such as optimizing refueling shutdowns. (Author)

  2. Reference Neutron Radiographs of Nuclear Reactor Fuel

    OpenAIRE

    Domanus, Joseph Czeslaw

    1986-01-01

    Reference neutron radiographs of nuclear reactor fuel were produced by the Euraton Neutron Radiography Working Group and published in 1984 by the Reidel Publishing Company. In this collection a classification is given of the various neutron radiographic findings, that can occur in different parts of pelletized, annular and vibro-conpacted nuclear fuel pins. Those parts of the pins are shown where changes of appearance differ from those for the parts as fabricated. Also radiographs of those as...

  3. Nuclear data for fusion reactor technology

    International Nuclear Information System (INIS)

    The meeting was organized in four sessions and four working groups devoted to the following topics: Requirements of nuclear data for fusion reactor technology (6 papers); Status of experimental and theoretical investigations of microscopic nuclear data (10 papers); Status of existing libraries for fusion neutronic calculations (5 papers); and Status of integral experiments and benchmark tests (6 papers). A separate abstract was prepared for each of these papers

  4. Radiation shield for nuclear reactors

    International Nuclear Information System (INIS)

    A reusable radiation shield for use in a reactor installation comprises a thin-walled, flexible and resilient container, made of plastic or elastomeric material, containing a hydrogenous fluid with boron compounds in solution. The container can be filled and drained in position and the fluid can be recirculated if required. When not in use the container can be folded and stored in a small space. The invention relates to a shield to span the top of the annular space between a reactor vessel and the primary shield. For this purpose a continuous toroidal container or a series of discrete segments is used. Other forms can be employed for different purposes, e.g. mattress- or blanket-like forms can be draped over potential sources of radiation or suspended from a mobile carrier and placed between a worker and a radiation source. (author)

  5. Nuclear reactor alignment plate configuration

    Science.gov (United States)

    Altman, David A; Forsyth, David R; Smith, Richard E; Singleton, Norman R

    2014-01-28

    An alignment plate that is attached to a core barrel of a pressurized water reactor and fits within slots within a top plate of a lower core shroud and upper core plate to maintain lateral alignment of the reactor internals. The alignment plate is connected to the core barrel through two vertically-spaced dowel pins that extend from the outside surface of the core barrel through a reinforcement pad and into corresponding holes in the alignment plate. Additionally, threaded fasteners are inserted around the perimeter of the reinforcement pad and into the alignment plate to further secure the alignment plate to the core barrel. A fillet weld also is deposited around the perimeter of the reinforcement pad. To accomodate thermal growth between the alignment plate and the core barrel, a gap is left above, below and at both sides of one of the dowel pins in the alignment plate holes through with the dowel pins pass.

  6. Desarrollo de una metodología de simulación de secuencias en accidente en centrales nucleares de agua ligera considerando actuaciones del operador

    OpenAIRE

    Expósito Lorenzo, Antonio

    2006-01-01

    This thesis presents the work carried out to develop a simulation system for nuclear power plants, denominated TRETA / COPMA-III integrated simulator, which allows the simulation of the thermalhydraulic processes that take place in this type of facilities, in normal operation and emergency operation, as well as the control room crew actions related with the management of the emergency situations. The simulation of the thermalhydraulic processes is carried out by means of the TR...

  7. Neutronics of nuclear power reactors

    International Nuclear Information System (INIS)

    This review, prepared on the occasion of 25th ETAN Conference describes the research activities in the field of neutronics which started in 1947. A number of researchers in Yugoslav Institutes was engaged in development of neutronics theory and calculation methods related to power reactors since 1960. To illustrate the activities of Yugoslav authors, this review contains the list of the most important relevant papers published in international journals

  8. Actinide transmutation in nuclear reactors

    International Nuclear Information System (INIS)

    An optimization method is developed to maximize the burning capability of the ALMR while complying with all constraints imposed on the design for reliability and safety. This method leads to a maximal transuranics enrichment, which is being limited by constraints on reactivity. The enrichment can be raised by using the neutrons less efficiently by increasing leakage from the fuel. With the developed optimization method, a metallic and an oxide fueled ALMR were optimized. Both reactors perform equally well considering the burning of transuranics. However, metallic fuel has a much higher heat conductivity coefficient, which in general leads to better safety characteristics. In search of a more effective waste transmuter, a modified Molten Salt Reactor was designed. A MSR operates on a liquid fuel salt which makes continuous refueling possible, eliminating the issue of the burnup reactivity loss. Also, a prompt negative reactivity feedback is possible for an overmoderated reactor design, even when the Doppler coefficient is positive, due to the fuel expansion with fuel temperature increase. Furthermore, the molten salt fuel can be reprocessed based on a reduction process which is not sensitive to the short-lived spontaneously fissioning actinides. (orig./HP)

  9. Nuclear fusion reactor material data base

    International Nuclear Information System (INIS)

    The working conditions for the materials to be used for nuclear fusion reactors are many sided, complicated and harsh. The existing experimental results can not be employed directly for reactor design. In such a case, it is insufficient to simply accumulate the experimental data on the specific properties of specific materials, and it is necessary to predict the material behaviour in the reactor system by rearranging those data in accordance with the purpose. When extreme characteristics are frequently pursued, wide insight is necessary regarding from the fundamental theory to the testing of practical equipment. In the development of nuclear fusion reactor materials, it is especially important to satisfy the condition that the design purpose of the system for selecting the optimum materials should be fully understood. A new material engineering approach has become necessary, in which a barrier existing so far between materials and the design is removed. From this viewpoint, the specifications, present status and design and development of material data base presently under development, the use of the data base made for trial, and the interface of material development and nuclear fusion reactor design, are described. In this data base, most of the data handle literature data, and the event data base mainly composed of experimental data is very few, similarly to other fields. Data modification will be necessary to respond the questions of users. (Wakatsuki, Y.)

  10. Nuclear reactors - the inevitable energy option

    International Nuclear Information System (INIS)

    The demand for energy in India is sure to rise year after year. Every possible energy source needs to be utilized to its fullest potential to bridge the gap between the demand and supply of electricity. Even while deciding the energy option, the availability of natural resources for future generation and effect of environment for the energy option chosen are to be taken care of. Out of the non conventional sources of electricity, nuclear electricity has greatest potential. Robust and safe energy option has to be harnessed to its potential. We have to bring down the cost of electricity. Even among nuclear reactors, electricity through Fast Breeder Reactors has greater potential. The Prototype Fast Breeder Reactor is a trend setter for moving into an era of electricity generation in the country. The paper brings details of the safety features, accomplishments of the technical challenges and the efforts on hand to reduce the unit energy cost by Nuclear Reactors. The paper also touches upon advantages, environmental impact of Fast Breeder Reactors for this abundant energy resources. Paper will also give a glimpse on technological challenges in design, construction and the preservation. (author)

  11. Regulatory challenges in the management of aging of structural materials in nuclear power plants; Retos reguladores en la gestion del envejecimiento de los materiales estructurales de centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Castelo, C.; Mendoza, C.; Mas, E.; Conde, J. M.

    2013-07-01

    The article discusses two major pathways by which a regulatory body, and in particular the CSN, may participate in the acquisition of the necessary knowledge on mechanisms of aging of nuclear structural materials: to participate in forums to share operational experience and R and R project, both nationally and internationally. It notes the importance of this participation to carry out its regulatory function based on the knowledge acquired and the unique challenge of transferring that knowledge to rules and guidelines for their application. The article discusses various R and D projects in which the CSN participates directly. It calls for the presence of regulatory bodies in R and D project funded by the EU and the transfer of the results of such projects to codes, standards or guidelines for feasible implementation. (Author)

  12. The siting of UK nuclear reactors

    International Nuclear Information System (INIS)

    Choosing a suitable site for a nuclear power station requires the consideration and balancing of several factors. Some ‘physical’ site characteristics, such as the local climate and the potential for seismic activity, will be generic to all reactors designs, while others, such as the availability of cooling water, the area of land required and geological conditions capable of sustaining the weight of the reactor and other buildings will to an extent be dependent on the particular design of reactor chosen (or alternatively the reactor design chosen may to an extent be dependent on the characteristics of an available site). However, one particularly interesting tension is a human and demographic one. On the one hand it is beneficial to place nuclear stations close to centres of population, to reduce transmission losses and other costs (including to the local environment) of transporting electricity over large distances from generator to consumer. On the other it is advantageous to place nuclear stations some distance away from such population centres in order to minimise the potential human consequences of a major release of radioactive materials in the (extremely unlikely) event of a major nuclear accident, not only in terms of direct exposure but also concerning the management of emergency planning, notably evacuation. This paper considers the emergence of policies aimed at managing this tension in the UK. In the first phase of nuclear development (roughly speaking 1945–1965) there was a highly cautious attitude, with installations being placed in remote rural locations with very low population density. The second phase (1965–1985) saw a more relaxed approach, allowing the development of AGR nuclear power stations (which with concrete pressure vessels were regarded as significantly safer) closer to population centres (in ‘semi-urban’ locations, notably at Hartlepool and Heysham). In the third phase (1985–2005) there was very little new nuclear

  13. Power Nuclear Reactors: technology and innovation for development in future

    International Nuclear Information System (INIS)

    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

  14. Fluidized bed nuclear reactor as a IV generation reactor

    International Nuclear Information System (INIS)

    The object of this paper is to analyze the characteristics of the Fluidized Bed Nuclear Reactor (FBNR) concept under the light of the requirements set for the IV generation nuclear reactors. It is seen that FBNR generally meets the goals of providing sustainable energy generation that meets clean air objectives and promotes long-term availability of systems and effective fuel utilization for worldwide energy production; minimize and manage their nuclear waste and notably reduce the long term stewardship burden in the future, thereby improving protection for the public health and the environment; increase the assurance that it is a very unattractive and least desirable route for diversion or theft of weapons-usable materials; excel in safety and reliability; have a very low likelihood and degree of reactor core damage; eliminate the need for offsite emergency response; have a clear life-cycle cost advantage over other energy sources; have a level of financial risk comparable to other energy projects. The other advantages of the proposed design are being modular, low environmental impact, exclusion of severe accidents, short construction period, flexible adaptation to demand, excellent load following characteristics, and competitive economics. (author)

  15. Nuclear reactor spring strip grid spacer

    International Nuclear Information System (INIS)

    An improved and novel grid spacer for maintaining the fuel rods of a nuclear reactor fuel assembly in substantially parallel array is described. The invention provides for spring strips to maintain the fuel elements in their desired orientation which have more positive alignment than previous types while allowing greater flexibility to counterbalance the effects of differential thermal expansion. (UK)

  16. Inspecting fuel pellets for nuclear reactor

    International Nuclear Information System (INIS)

    An improved method of controlling the inspection, sorting and classifying of nuclear reactor fuel pellets, including a mechanical handling system and a computer controlled data processing system, is described. Having investigated the diameter, length, surface flaws and weights of the pellets, they are sorted accordingly and the relevant data are stored. (U.K.)

  17. Current Abstracts Nuclear Reactors and Technology

    Energy Technology Data Exchange (ETDEWEB)

    Bales, J.D.; Hicks, S.C. [eds.

    1993-01-01

    This publication 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 during the past month. Also included are US information obtained through acquisition programs or interagency agreements and international 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 the Energy Science and Technology Database and Nuclear Science Abstracts (NSA) database. Current information, added daily to the Energy Science and Technology Database, 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.

  18. Mounting impellors of nuclear reactor pumps

    International Nuclear Information System (INIS)

    A coolant pump for a liquid metal cooled nuclear reactor construction is described which has its impellor hub mounted on the drive shaft with an annular clearance. A drive coupling is effected by a plurality of cylindrical keys disposed parallel to the longitudinal axes of the hub and shaft and equally spaced in the clearance between them. (author)

  19. ETRR-2 nuclear reactor: Facility specification

    International Nuclear Information System (INIS)

    The report provides technical details on the ETRR-2 nuclear reactor core and immediate structure for analysis purposes. The goal of the report is to provide sufficient geometric and material data to build a computational neutronic model of the facility. (author)

  20. Nuclear reactor fuel assembly spacer grids

    International Nuclear Information System (INIS)

    Designs of nuclear reactor fuel assembly spacer grids for supporting and spacing fuel elements are described which do not utilize resilient grid plate protrusions in the peripheral band but retain the advantages inherent in the combination resilient and rigid protrusion cells. (U.K.)

  1. Refueling machine for a nuclear reactor

    International Nuclear Information System (INIS)

    An improved refuelling machine for inserting and removing fuel assemblies from a nuclear reactor is described which has been designed to increase the reliability of such machines. The system incorporates features which enable the refuelling operation to be performed more efficiently and economically. (U.K.)

  2. Nuclear heating in fusion reactors

    International Nuclear Information System (INIS)

    Nuclear heating rates are estimated with multiplying neutron and gamma-ray fluxes by nuclear heating constants of KERMA factors. The gamma-ray KERMA factors can be given exactly by the value of gamma-ray energy deposition of (incident gamma-ray energy) - (outgoing gamma-ray energy). On the other hand there are two methods to calculating the neutron KERMA factors. One is a direct method and the other an energy balance method. In the direct method the neutron KERMA factors are given by kinematics. The uncertainties in KERMA factors by this method are considered the same as those in nuclear data of 20-30%. The KERMA factors by the energy balance method have several ten per cent uncertainties, because of deduction between large values each other. In fact since all nuclear data are not always consistent with each other and have uncertainties, we had better apply the direct method. In considering decay heat of short half life nuclides into KERMA factors, there are some problems such as application to shorter pulse operation than the half life of the nuclides. It is emphasized that nuclear heat and decay heat should be separately treated. (author)

  3. PSA-oriented analysis of transients in Atucha Unit II PHWR with a RELAP model

    International Nuclear Information System (INIS)

    This paper presents a model of the Atucha II (PHWR) nuclear power plant developed for RELAP5/MOD3.3. The nodalization was implemented to comply with the Probabilistic Safety Analysis required in the licensing, commissioning, and operating process. A data management platform is implemented that processes the geometric and physical data for nodalization and generates the code input. The steady state is analyzed by comparing calculated variables with their respective design values and previous calculations performed with other models/tools. Finally, a case of loss of heat sink caused by an electrical supply failure is analyzed. (author)

  4. Operation and utilizations of Dalat nuclear research reactor

    International Nuclear Information System (INIS)

    The reconstructed Dalat nuclear research reactor was commissioned in March 1984 and up to September 1988 more than 6200 hours of operation at nominal power have been recorded. The major utilizations of the reactor include radioisotope production, activation analysis, nuclear data research and training. A brief review of the utilizations of the reactor is presented. Some aspects of reactor safety are also discussed. (author)

  5. Ultrasonic flaw detection device in nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Yoshishige; Takabayashi, Jun-ichi

    1996-02-02

    Flaws on an outer circumferential surface of a shroud at inner side than jet pumps in a nuclear reactor are simply and reliably detected using ultrasonic waves. Ultrasonic waves are irradiated to the shroud which surrounds a reactor core at the inner side and has a plurality of jet pumps at the outer side at predetermined distances. An ultrasonic wave probe which detects flaws based on reflecting waves is suspended using a suspending rope. A jet nozzle is attached to a probe-attaching portion and water is jetted out to reactor water to move the probe-attaching portion in the reactor. Then, flaws can be detected easily and reliably using ultrasonic waves even at a narrow gap at the inner side of each jet pump. (N.H.).

  6. Ultrasonic flaw detection device in nuclear reactor

    International Nuclear Information System (INIS)

    Flaws on an outer circumferential surface of a shroud at inner side than jet pumps in a nuclear reactor are simply and reliably detected using ultrasonic waves. Ultrasonic waves are irradiated to the shroud which surrounds a reactor core at the inner side and has a plurality of jet pumps at the outer side at predetermined distances. An ultrasonic wave probe which detects flaws based on reflecting waves is suspended using a suspending rope. A jet nozzle is attached to a probe-attaching portion and water is jetted out to reactor water to move the probe-attaching portion in the reactor. Then, flaws can be detected easily and reliably using ultrasonic waves even at a narrow gap at the inner side of each jet pump. (N.H.)

  7. Space nuclear reactor power plants

    International Nuclear Information System (INIS)

    Requirements for electrical and propulsion power for space are expected to increase dramatically in the 1980s. Nuclear power is probably the only source for some deep space missions and a major competitor for many orbital missions, especially those at geosynchronous orbit. Because of the potential requirements, a technology program on space nuclear power plant components has been initiated by the Department of Energy. The missions that are foreseen, the current power plant concept, the technology program plan, and early key results are described

  8. Safety review, assessment and inspection on research reactors, experimental reactors and nuclear heating reactors

    International Nuclear Information System (INIS)

    The NNSA and its regional office step further strengthened the regulation on the safety of in-service research reactors in 1996. A lot of work has been done on the supervision of safe in rectifying the review and assessment of modified items, the review of operational documents, the treatment of accidents, the establishment of the system for operational experience feedback, daily and routine inspection on nuclear safety. The internal management of the operating organization on nuclear safety was further strengthened, nuclear safety culture was further enhanced, the promotion in nuclear safety and the safety situation for in-service research reactors were improved

  9. Safety review, assessment and inspection on research reactors, experimental reactors, nuclear heating reactors and critical facilities

    International Nuclear Information System (INIS)

    The NNSA organized mainly in 1999 to complete the verification loop in core of the high flux experimental reactor with the 2000 kW fuel elements, the re-starting of China Pulsed Reactor, review and assessment on nuclear safety for the restarting of the Uranium-water critical Facility and treat the fracture event with the fuel tubes in the HWRR

  10. ENS RRFM 2006: 10th international topical meeting on research reactor fuel management. Transactions

    International Nuclear Information System (INIS)

    The RRFM Conference is organized by the European Nuclear Society (ENS) with co-operation of the IAEA. It includes detailed scientific and technical reports reports on the following main topics: Fuel development, qualification, fabrication and licensing; Spent fuel management, back-end options and transportation; Reactor operation, fuel safety and core conversion; Innovative methods in research reactor analysis; Global Treat Reduction Initiative

  11. Software reliability and safety in nuclear reactor protection systems

    International Nuclear Information System (INIS)

    Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor

  12. Software reliability and safety in nuclear reactor protection systems

    Energy Technology Data Exchange (ETDEWEB)

    Lawrence, J.D. [Lawrence Livermore National Lab., CA (United States)

    1993-11-01

    Planning the development, use and regulation of computer systems in nuclear reactor protection systems in such a way as to enhance reliability and safety is a complex issue. This report is one of a series of reports from the Computer Safety and Reliability Group, Lawrence Livermore that investigates different aspects of computer software in reactor National Laboratory, that investigates different aspects of computer software in reactor protection systems. There are two central themes in the report, First, software considerations cannot be fully understood in isolation from computer hardware and application considerations. Second, the process of engineering reliability and safety into a computer system requires activities to be carried out throughout the software life cycle. The report discusses the many activities that can be carried out during the software life cycle to improve the safety and reliability of the resulting product. The viewpoint is primarily that of the assessor, or auditor.

  13. Use of nuclear reactors for seawater desalination

    International Nuclear Information System (INIS)

    The last International Atomic Energy Agency (IAEA) status report on desalination, including nuclear desalination, was issued nearly 2 decades ago. The impending water crisis in many parts of the world, and especially in the Middle East, makes it appropriate to provide an updated report as a basis for consideration of future activities. This report provides a state-of-the-art review of desalination and pertinent nuclear reactor technology. Information is included on fresh water needs and costs, environmental risks associated with alternatives for water production, and data regarding the technical and economic characteristics of immediately available desalination systems, as well as compatible nuclear technology. 68 refs, 60 figs, 11 tabs

  14. Modification and updating of documentation in equipment of panels of control room in nuclear power plant operation; Modificacion y actualizacion de documentacion en aparatos de paneles de sala de control en una central nuclear en operacion

    Energy Technology Data Exchange (ETDEWEB)

    Agudo Montero, L.

    2013-07-01

    The present paper describes a case very unique specific design of interactive 2D-CAD application, that has been developed by Empresarios Agrupados as engineering support to the nuclear power plants, aware of the problem that exists with the documentation of the instruments and devices that are on the panels of Control room, and that only have the documentation generated in its day by the manufacturers of these panels. To this end, an application (application DOPAB) has been developed to help solve the problem of management, design and modification of wiring and wiring devices existing in the Control room control panels.

  15. Reactor Structure Materials: Nuclear Fuel

    International Nuclear Information System (INIS)

    Progress and achievements in 1999 in SCK-CEN's programme on applied and fundamental nuclear fuel research in 1999 are reported. Particular emphasis is on thermochemical fuel research, the modelling of fission gas release in LWR fuel as well as on integral experiments

  16. Radiation protection in nuclear reactors

    International Nuclear Information System (INIS)

    Full text: People are exposed to ionizing radiation in many different forms: cosmic rays that penetrate earth atmosphere or radiation from soil and mineral resources are natural forms of ionizing radiation. Other forms are produced artificially using radioactive materials for various beneficial applications in medicine, industry and other fields. The greatest concerns about ionizing radiation are tied to its potential health effects and a system of radiation protection has been developed to protect people from harmful radiation. The promotion of radiation protection is one of the International Atomic Energy Agency main activities. Radiation protection concerns the protection of workers, members of public, and patients undergoing diagnosis and therapy against the harmful effects of ionizing radiation. The report covers the responsibility of radiation protection officer in Egypt Second Research Reactor (ETRR-2) in Inshas - Egypt, also presents the protection against ionizing radiation from external sources, including types of radiation, sources of radiation (natural - artificial), and measuring units of dose equivalent rate. Also covers the biological effects of ionizing radiation, personal monitoring and radiation survey instruments and safe transport of radioactive materials. The report describes the Egypt Second Research Reactor (ETRR-2), the survey instruments used, also presents the results obtained and gave a relations between different categories of data. (author)

  17. Full-scale vibration test of Atucha II N.P.P. Part VII: numerical simulation of steady state response by simplified axisymmetric Finite Element Method

    International Nuclear Information System (INIS)

    This work determines the dynamic characteristics of the reactor building of the heavy water reactor Atucha II and to checks certain analytical procedures used to validate full-scale vibration tests. Soil-structure interaction effects were incorporated by means of springs and dampers which represent the soil behavior. The full-scale vibration tests were reproduced using a simplified axisymmetric model. (author). 3 refs., 25 figs., 1 tab

  18. Perspective of nuclear energy and advanced reactors

    International Nuclear Information System (INIS)

    Future nuclear energy growth will be the result of the contributions of every single plant being constructed or projected at present as it is connected to the grid. As per IAEA, there exists presently 34 nuclear power plants under construction 81 with the necessary permits and funding and 223 proposed, which are plants seriously pursuing permits and financing. This means that in a few decades the number of nuclear power plants in operation will have doubled. This growth rate is characterised by the incorporation of new countries to the nuclear club and the gradually increasing importance of Asian countries. During this expansive phase, generation III and III+designs are or will be used. These designs incorporate the experience from operating plants, and introduce innovations on rationalization design efficiency and safety, with emphasis on passive safety features. In a posterior phase, generation IV designs, presently under development, will be employed. Generation IV consists of several types of reactors (fast reactors, very high temperature reactors, etc), which will improve further sustain ability, economy, safety and reliability concepts. The described situation seems to lead to a renaissance of the nuclear energy to levels hardly thinkable several years ago. (Author)

  19. Advanced integral reactor (SMART) for nuclear desalination

    International Nuclear Information System (INIS)

    At present, severe fresh water shortages are occurring in some regional areas of the Republic of Korea and the problem is expected to spread throughout the country within a decade unless appropriate and timely countermeasures are taken. Of these, nuclear sea water desalination is receiving much attention because the Republic of Korea has a firmly established nuclear environment and abundant sea water resources. In addition, nuclear plants provide cleaner energy than fossil plants, which is another important beneficial factor for countries as crowded as ours. With a view to applying nuclear desalination, development of SMART (system integrated modular advanced reactor) was initiated and is currently in progress. SMART is being developed as a 330 MW(th) integral reactor with passive safety features. The design of SMART is aimed at combining the firmly established commercial reactor design with new advanced technologies. This has led to the use of industry proven Korea optimized fuel assembly (KOFA) based fuels, while radically new technologies such as a self-pressurizing pressurizer, helical once-through steam generators and a new control concept are being developed. The current development status of SMART and its application to nuclear desalination are presented. (author)

  20. A nuclear reactor for district heating

    International Nuclear Information System (INIS)

    Global energy requirements are expected to double over the next 40 years. In the northern hemisphere, many countries consume in excess of 25 percent of their primary energy supply for building heating. Satisfying this need, within the constraints now being acknowledged for sustainable global development, provides an important opportunity for district heating. Fuel-use flexibility, energy and resource conservation, and reduced atmospheric pollution from acid gases and greenhouse gases, are important features offered by district heating systems. Among the major fuel options, only hydro-electricity and nuclear heat completely avoid emissions of combustion gases. To fill the need for an economical nuclear heat source, Atomic Energy of Canada Limited has designed a 10 MW plant that is suitable as a heat source within a network or as the main supply to large individual users. Producing hot water at temperatures below 100 degrees C, it incorporates a small pool-type reactor based on AECL's successful SLOWPOKE Research Reactor. A 2 MW prototype for the commercial unit is now being tested at the Whiteshell Nuclear Research Establishment in Manitoba. With capital costs of $7 million (Canadian), unit energy costs are projected to be $0.02/kWh for a 10 MW unit operating in a heating grid over a 30-year period. By keeping the reactor power low and the water temperature below 100 degrees C, much of the complexity of the large nuclear power plants can be avoided, thus allowing these small, safe nuclear heating systems to be economically viable

  1. Five Lectures on Nuclear Reactors Presented at Cal Tech

    Science.gov (United States)

    Weinberg, Alvin M.

    1956-02-10

    The basic issues involved in the physics and engineering of nuclear reactors are summarized. Topics discussed include theory of reactor design, technical problems in power reactors, physical problems in nuclear power production, and future developments in nuclear power. (C.H.)

  2. Radioactive target needs for nuclear reactor physics and nuclear astrophysics

    International Nuclear Information System (INIS)

    Nuclear reaction cross sections of short-lived nuclei are key inputs for new generation nuclear reactor simulations and for models describing the nucleosynthesis of elements. After discussing various topics of nuclear astrophysics and reactor physics where the demand of nuclear data on unstable nuclei is strong, we describe the general characteristics of the targets needed to measure the requested data. In some cases the half-life of the nucleus of interest is so short that it is not possible to produce a target and perform the measurement. However, some alternative methods have been developed that allow one to obtain neutron-induced cross sections of highly radioactive nuclei. One of these methods is the surrogate reaction technique. We explain the principle of the surrogate method and describe the characteristics of the targets used in surrogate experiments.

  3. Full scale dynamic tests of Atucha II NPP

    International Nuclear Information System (INIS)

    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)

  4. Fuel rod for nuclear reactors

    International Nuclear Information System (INIS)

    The fuel or breeder element with fission gas plenum has hollow spaces on the central part of the fuel or breeder material volume, which are filled with sodium. During operation, this sodium provides a second heat transport mechanism, as the sodium evaporates, rises into the fission gas plenum, condenses on the metal sleeve there and returns to the fuel or breeding zone under gravity or capillary effect. (DG)

  5. Technical aspects of the process of segmentation and packaging of the reactor vessel of Jose Cabrera NPP; Aspectos tecnicos del proceso de segmentacion y embalaje de la vasija del reactor de la central nuclear Jose Cabrera

    Energy Technology Data Exchange (ETDEWEB)

    Valdivieso, J. M.; Garcia Castro, R.

    2015-07-01

    Westinghouse is carrying out the segmentation of the reactor pressure vessel (RPV) within the framework of the Dismantling and Decommissioning Project of the Jose Cabrera NPP. The final concept is based on the comprehensive Westinghouse experience in the field of LWR pressure vessel and internals segmentation, and particularly in previous reactor internals segmentation project for Jose Cabrera NPP. This article shows the development of all the activities included: cutting method selection, preparatory works, cutting activities, waste characterization and packaging activities. (Author)

  6. The Nuclear Safety Council's Instruction IS-30 on program requirements of fire protection at nuclear power plants; La instruccion IS-30 del consejo de Seguridad Nuclear sobre requisitos del programa de proteccion contraincendios en centrales nucleares

    Energy Technology Data Exchange (ETDEWEB)

    Peco, J.

    2015-07-01

    The Nuclear Safety Councils Instrumentation IS-30 is the standard that establishes the fire protection program requirements for the Spanish nuclear power plants with operating license in order to satisfy the two fire protection objectives, which are the adoption of the defense-in-depth principle for fire protection and, by fire area confinement, to ensure that one train of components needed to achieve and maintain the safe shutdown conditions is free of fire damage, and that radioactive liberation is minimized. (Author)

  7. Research nuclear reactor RA, Annual Report 2001

    International Nuclear Information System (INIS)

    During 2001, activities at the RA research nuclear reactor in were performed according to the Contract about financing of the RA reactor for the period January-December 2001, signed by the Ministry of Science, technology and development of the Republic of Serbia. RA reactor was not operated since shutdown in August 1984. Although, the most of the planned reconstruction activities were finished until 1991, the most important, which was concerned with exchange of the reactor instrumentation, financed by the IAEA, was interrupted due to international sanctions imposed on the country. Since 1992, all the renewal and reconstruction activities were ceased. Continuous aging and degradation of the equipment and facilities demand decision making about the future status of the Ra reactor. Until this decision is made it is an obligation to maintain control and maintenance of ventilation system, power supply, internal transportation system, spent fuel storage, hot cells, electronic fuel surveillance system, and part of the stationary dosimetry system. In 2001, apart from the mentioned activities, actions were undertaken related to maintenance of the reactor building and installations. The most important tasks fulfilled were: protection of the roof of the ventilation system building, purchase and installing the fire protection system and twelve new battery cells in the reactor building. There were no actions concerned with improvement of the conditions for intermediate spent fuel storage. With the support of IAEA, actions were initiated for possible transport of the spent fuel tu Russia. At the end of 2001, preparations were started for possible future decommissioning of the RA reactor. After, renewal of the membership of our country in the IAEA, the Government of Yugoslavia has declared its attitude about the intention of RA reactor decommissioning at the General Conference in September 2001

  8. Radiological surveillance in Mexico, derived of the accident of the Fukushima Daiichi nuclear power plant; Vigilancia radiologica en Mexico, derivado del accidente en la central nuclear de Fukushima Daiichi

    Energy Technology Data Exchange (ETDEWEB)

    Aguirre G, J.; Nohpal J, X., E-mail: jaguirre@cnsns.gob.mx [Comision Nacional de Seguridad Nuclear y Salvaguardias, Departamento de Vigilancia Radiologica, Dr. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

    2012-10-15

    March 11, 2011 an earthquake of 9.0 grades in the Richter scale, originated in the coast of Tohoku, Japan, in the Pacific Ocean gave origin to a tsunami that caused an accident in the Fukushima Daiichi nuclear power plant. Due to this accident, derived of the loss of the reactor cooling system, as well as of the prolonged absence of alternating and direct current, radiological protection actions were realized without being able to avoid the liberation of radioactive material to the atmosphere and ocean. The radiological impact of these liberations, not only in Japan but around the world, mainly in the north hemisphere of the Earth, was analyzed by means of environmental dose measurements and radionuclide concentrations in soil and water, among others. In the Mexico case, air samples data were obtained, as well as environmental dose celerity and full-length counts of the people coming from Japan near the disaster area. The present work contains the obtained results of the realized measurements in Mexico, same that have been used to make a summary and analysis of the dispersion in the environment in several countries of the world. (Author)

  9. Corrosion in nuclear power plants and it implication in the leak before break criteria; Corrosion en centrales nucleares y su implicacion en el criterio de fuga antes de ruptura

    Energy Technology Data Exchange (ETDEWEB)

    Martinez M, E

    1992-05-15

    The corrosion in a general way can be defined like a chemical or electrochemical reaction, which is carried out in the surface of the metallic materials exposed to a specific medium. Due to the operation conditions in the nuclear power plants are practically fixed and its modification in most of the cases is difficult or expensive, the natural tendency to prevent the corrosion has been generally directed toward the selection of materials. Numerous materials have been employees as substitute of the traditional steels, among other the stabilized stainless steels, those of extra low carbon and numerous nickel base alloys. The basic evaluation that establishes the approach Leak before break (LBB) it involves the analysis by means of 'fracture mechanics' of a postulated crevice that it crosses the thickness of the material, which causes a same flight, with a margin of safety, to the detection limit that it has for a shutdown of the reactor. Due to the crevice size postulated, it cannot be established highly starting from the mechanical properties of the material since these its will be affected by the corrosion mechanisms that can settle down, it was determined that the implementation of the LBB criteria, it cannot be established for components or systems that its are susceptible of suffering corrosion. (Author)

  10. Dismantling technology of nuclear reactors

    International Nuclear Information System (INIS)

    Because of nuclear power plants include very high radioactive inventories in their components, both a large scale steel and a massive reinforced concrete at the end of the plant lifetime, these radioactivities make it difficult to dismantle them. To attack these problems the state-of-the-art dismantling technologies of these components were surveyed and reviewed, and a case study of immediate dismantling and safe storage on a small scale BWR power station was carried out and disscussed. Immediate dismantling and safe storage on a large scale BWR power station was also surveyed and reviewed. In-site entombment applied to a small and a large scale BWR power stations was also discussed. The results surveyed and discussed were described in this report as the third year activity of dismantling technology subcommittee of the Committee on study of Decommissioning of Nuclear Facilities. (author)

  11. Erection device for nuclear reactors

    International Nuclear Information System (INIS)

    The assembly and crane unit, respectively, can be used in the erection of the reactor plant proper and in operation jobs. It meets the safety requirements under which cables must be installed in duplicate. The device mainly consists of two cable drums, power transmissions and hoists with one cable each for a certain load. A variable transmission with exchangeable wheel pairs sets various lifting speeds. One of the power transmission can be coupled with the variable transmission by a driving connection with exchangeable shafts of different lengths. A common motor is provided for the hoists. A rigid connection covering various distances connects the two hoists. Moreover, there is a cable equalizer with an adjustment device, e.g., a spring or a lever mechanism, for one of the cables. (DG/RF)

  12. Simulation of a marine nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kusunoki, Tsuyoshi; Kyouya, Masahiko; Kobayashi, Hideo; Ochiai, Masaaki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Office of Nuclear Ship Research and Development

    1995-02-01

    A Nuclear-powered ship Engineering Simulation SYstem (NESSY) has been developed by the Japan Atomic Energy Research Institute as an advanced design tool for research and development of future marine reactors. A marine reactor must respond to changing loads and to the ship`s motions because of the ship`s maneuvering and its presence in a marine environment. The NESSY has combined programs for the reactor plant behavior calculations and the ship`s motion calculations. Thus, it can simulate reactor power fluctuations caused by changing loads and the ship`s motions. It can also simulate the behavior of water in the pressurizer and steam generators. This water sloshes in response to the ship`s motions. The performance of NESSY has been verified by comparing the simulation calculations with the measured data obtained by experiments performed using the nuclear ship Mutsu. The effects of changing loads and the ship`s motions on the reactor behavior can be accurately simulated by NESSY.

  13. Nuclear reactor control room construction

    International Nuclear Information System (INIS)

    A control room for a nuclear plant is disclosed. In the control room, objects labelled 12, 20, 22, 26, 30 in the drawing are no less than four inches from walls labelled 10.2. A ceiling contains cooling fins that extend downwards toward the floor from metal plates. A concrete slab is poured over the plates. Studs are welded to the plates and are encased in the concrete. 6 figures

  14. Central Bureau for Nuclear Measurements

    International Nuclear Information System (INIS)

    The main task of CBNM is defined as the specific programme Nuclear Measurements and Reference Materials. In the field of neutron data for standards, for fission and for fusion application, the nuclear charge distribution and odd-even effects for mass, charge and neutron number in the cold spontaneous fission of 252Cf were determined. X- and γ-ray emission probabilities were evaluated in the frame of an IAEA coordinated Research Project. The subthermal fission cross section measurements of 235U, 233U and 239Pu, were finalised. The dependence of the experimental weighting function of C6D6 detectors on thickness of several 56Fe samples was determined. Fusion data studies involved the development of a light-ion telescope with improved time - and energy resolution. Double differential cross-sections of 9Be were analysed. Radionuclide metrology dealt with the response of silicon detectors, as well as with the standardization of 192Ir sources. Project Reference Materials reports the EC Certification of nuclear reference materials 210 (PuO2), 523 (Al), 525 (Nb) and 526 (Nb). Progress was achieved in the preparation of dried solid spikes of uranium and plutonium for undiluted reprocessing input solution analysis. 10B and 6Li deposits were prepared for a redetermination of the neutron lifetime. Preliminary studies on speciation of trace metals in biological fluids were successful. Radioactive waste barrels were analysed by γ-scanning and blood samples were irradiated with 0.6 MeV neutrons. Exploratory research resulted in first measurements of transition radiation properties

  15. Method and apparatus for increasing fuel efficiency in nuclear reactors

    International Nuclear Information System (INIS)

    This patent describes an improved method of producing a spectral shift in a nuclear reactor to achieve increased nuclear fuel efficiency, the nuclear reactor containing a fluid moderator juxtaposed with fuel elements containing the nuclear fuel, which comprises disposing within the fluid moderator stationary non-poison displacer rods for achieving the spectral shift, the displacer rods exhibiting a continuous reduction in volume during operation of the nuclear reactor whereby the fluid moderator increases in volume as the nuclear fuel is burned in the nuclear reactor

  16. Cold nuclear fusion reactor and nuclear fusion rocket

    Directory of Open Access Journals (Sweden)

    Huang Zhenqiang

    2013-10-01

    Full Text Available "Nuclear restraint inertial guidance directly hit the cold nuclear fusion reactor and ion speed dc transformer" [1], referred to as "cold fusion reactor" invention patents, Chinese Patent Application No. CN: 200910129632.7 [2]. The invention is characterized in that: at room temperature under vacuum conditions, specific combinations of the installation space of the electromagnetic field, based on light nuclei intrinsic magnetic moment and the electric field, the first two strings of the nuclei to be bound fusion on the same line (track of. Re-use nuclear spin angular momentum vector inherent nearly the speed of light to form a super strong spin rotation gyro inertial guidance features, to overcome the Coulomb repulsion strong bias barrier to achieve fusion directly hit. Similar constraints apply nuclear inertial guidance mode for different speeds and energy ion beam mixing speed, the design of ion speed dc transformer is cold fusion reactors, nuclear fusion engines and such nuclear power plants and power delivery systems start important supporting equipment, so apply for a patent merger

  17. Acoustic transducer for nuclear reactor monitoring

    International Nuclear Information System (INIS)

    Disclosed is a transducer to monitor a parameter and produce an acoustic signal from which the monitored parameter can be recovered. The transducer comprises a modified Galton whistle which emits a narrow band acoustic signal having a frequency dependent upon the parameter being monitored, such as the temperature of the cooling media of a nuclear reactor. Multiple locations within a reactor are monitored simultaneously by a remote acoustic receiver by providing a plurality of transducers each designed so that the acoustic signal it emits has a frequency distinct from the frequencies of signals emitted by the other transducers, whereby each signal can be unambiguously related to a particular transducer. 8 claims, 1 figure

  18. Digital instrumentation for nuclear research reactors

    International Nuclear Information System (INIS)

    The problem in measuring the neutron flux in a nuclear reactor start-up pulse channel is related to the statistical fluctuation in a wide measuring range, from a few cps up to 106 cps. This article presents the development of a star up pulse channel with digital filtering of the neutron flux and its rate variation to avoid the statistical fluctuation and obtain stable readings with a response time of 50 ms. Results are shown compared with the traditional analog instrumentation of Argonauta reactor. (author (author)

  19. Liquid metal cooled fast breeder nuclear reactor

    International Nuclear Information System (INIS)

    A liquid metal cooled fast breeder nuclear reactor has a core comprising a plurality of fuel assemblies supported on a diagrid and submerged in a pool of liquid metal coolant within a containment vessel, the diagrid being of triple component construction and formed of a short cylindrical plenum mounted on a conical undershell and loosely embraced by a fuel store carrier. The plenum merely distributes coolant through the fuel assemblies, the load of the assemblies being carried by the undershell by means of struts which penetrate the plenum. The reactor core, fuel store carrier and undershell provide secondary containment for the plenum. (UK)

  20. Distributed computing and nuclear reactor analysis

    International Nuclear Information System (INIS)

    Large-scale scientific and engineering calculations for nuclear reactor analysis can now be carried out effectively in a distributed computing environment, at costs far lower than for traditional mainframes. The distributed computing environment must include support for traditional system services, such as a queuing system for batch work, reliable filesystem backups, and parallel processing capabilities for large jobs. All ANL computer codes for reactor analysis have been adapted successfully to a distributed system based on workstations and X-terminals. Distributed parallel processing has been demonstrated to be effective for long-running Monte Carlo calculations

  1. Emergency cooling system for nuclear reactors

    International Nuclear Information System (INIS)

    Upon the occasion of loss of coolant in a nuclear reactor as when a coolant supply or return line breaks, or both lines break, borated liquid coolant from an emergency source is supplied in an amount to absorb heat being generated in the reactor even after the control rods have been inserted. The liquid coolant flows from pressurized storage vessels outside the reactor to an internal manifold from which it is distributed to unused control rod guide thimbles in the reactor fuel assemblies. Since the guide thimbles are mounted at predetermined positions relative to heat generating fuel elements in the fuel assemblies, holes bored at selected locations in the guide thimble walls, sprays the coolant against the reactor fuel elements which continue to dissipate heat but at a reduced level. The cooling water evaporates upon contacting the fuel rods thereby removing the maximum amount of heat (970 BTU per pound of water) and after heat absorption will leave the reactor in the form of steam through the break which is the cause of the accident to help assure immediate core cooldown

  2. Medical Radioisotopes Production Without A Nuclear Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Van der Keur, H.

    2010-05-15

    This report is answering the key question: Is it possible to ban the use of research reactors for the production of medical radioisotopes? Chapter 2 offers a summarized overview on the history of nuclear medicine. Chapter 3 gives an overview of the basic principles and understandings of nuclear medicine. The production of radioisotopes and its use in radiopharmaceuticals as a tracer for imaging particular parts of the inside of the human body (diagnosis) or as an agent in radiotherapy. Chapter 4 lists the use of popular medical radioisotopes used in nuclear imaging techniques and radiotherapy. Chapter 5 analyses reactor-based radioisotopes that can be produced by particle accelerators on commercial scale, other alternatives and the advantages of the cyclotron. Chapter 6 gives an overview of recent developments and prospects in worldwide radioisotopes production. Chapter 7 presents discussion, conclusions and recommendations, and is answering the abovementioned key question of this report: Is it possible to ban the use of a nuclear reactor for the production of radiopharmaceuticals? Is a safe and secure production of radioisotopes possible?.

  3. Nuclear safety cooperation for Soviet designed reactors

    International Nuclear Information System (INIS)

    The nuclear accident at the Chernobyl nuclear power plant in 1986 first alerted the West to the significant safety risks of Soviet designed reactors. Five years later, this concern was reaffirmed when the IAEA, as a result of a review by an international team of nuclear safety experts, announced that it did not believe the Kozloduy nuclear power plants in Bulgaria could be operated safely. To address these safety concerns, the G-7 summit in Munich in July 1992 outlined a five point program to address the safety problems of Soviet Designed Reactors: operational safety improvement; near-term technical improvements to plants based on safety assessment; enhancing regulatory regimes; examination of the scope for replacing less safe plants by the development of alternative energy sources and the more efficient use of energy; and upgrading of the plants of more recent design. As of early 1994, over 20 countries and international organizations have pledged hundreds of millions of dollars in financial assistance to improve safety. This paper summarizes these assistance efforts for Soviet designed reactors, draws lessons learned from these activities, and offers some options for better addressing these concerns

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

    International Nuclear Information System (INIS)

    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)

  5. Lubrication greases for nuclear reactors

    International Nuclear Information System (INIS)

    Lubricating greases are essential components of many machines used in nuclear power plants. Where these machines are subject to radiation the life of the grease will be reduced due to deterioration of the components of the grease. According to the chemical nature of the grease used a greater or lesser resistance to radiation will be observed. Tests and techniques to evaluate the performance of greases before and after irradiation are described. The results of these tests show that conventional premium greases will resist comparatively low levels of irradiation, whilst greases formulated from correctly selected components can tolerate quite high levels of radiation permitting the machines they lubricate to attain their designed service lives

  6. Determination of uranium traces in nuclear cans of nuclear reactors

    International Nuclear Information System (INIS)

    To quantify the uranium content as impurity can be found in zirconium alloys and zircaloy, utilized to construct the sheaths containing fuels of the reactors of nuclear plants. The determination by fluorescence spectroscopy was employed as quality control measurement, at once the corrosion resistance, diminish with the increase of the uranium content in the alloys. (Author)

  7. 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)

  8. Exploitation continuation of Fessenheim nuclear plant nr 1 reactor after thirty years of operation; Poursuite d'exploitation du reacteur n.1 de la centrale nucleaire de fessenheim apres trente annees de fonctionnement

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-07-01

    After having recalled the regulatory framework, this report indicates how the Fukushima accident has been taken into account by the French nuclear safety authority (ASN) for the decision of keeping on operating the Fessenheim nuclear plant. Then, after a general presentation of nuclear installations, the report describes some peculiarities of the Fessenheim power plant with respect to the other French nuclear plants. It comments and discusses various issues: reactor exploitation, fuel management, vessel exploitation, exploitation of the main secondary circuits, of the confinement enclosure, and of other equipment. It recalls significant events, exploitation rules, and modifications brought to the reactor. It gives a global assessment. The authors report the safety re-examination (approach, compliance examination, security re-assessment), controls performed during decennial inspection (main controls and tests, implementation of modifications foreseen by safety re-examination, significant events, monitoring by the ASN, reactor restarting after the third decennial inspection). Perspectives are then discussed for the ten following years in terms of maintenance policy, ageing management, reactor vessel serviceability, and additional actions within the frame of ageing management. The operation continuation is then discussed

  9. Nuclear reactors for research and radioisotope production in Argentina

    International Nuclear Information System (INIS)

    In Argentina, the construction, operation, and use of research and radioisotope production reactors is and has been an important method of personnel preparation for the nuclear power program. Moreover, it is a very suitable means for technology transfer to countries developing their own nuclear programs. At present, the following research reactors are in operation in Argentina: Argentine Reactor 0 (RA-0); Argentine Reactor 1 (RA-1); Argentine Reactor 2 (RA-2); Argentine Reactor 3 (RA-3); Argentine Reactor 4 (RA-4). The Argentine Reactor 6 (RA-6), under construction, should reach criticality in 1981

  10. Nuclear safety in the newly independent states in central and eastern Europe

    International Nuclear Information System (INIS)

    The Chernobyl nuclear reactor accident in 1986 has led to a reassessment of safety issues in the nuclear industry's of the Commonwealth of Independent States (CIS) and in central and eastern Europe. Three aspects of the problem are explained and addressed here, design inadequacies in the RBMK and other reactor types, less than adequate operational safety procedures and lack of independent government regulation of state utilities, allowing economic targets to overcome safe operation of plants. (UK)

  11. Proceedings of DAE-BRNS theme meeting on advances in reactor physics: design, analysis and operation of nuclear reactors

    International Nuclear Information System (INIS)

    Reactor Physics forms a very vital part of Nuclear Engineering. It often initiates and contributes to new reactor concepts and designs. Reactor physics starts with nuclear physics aspects in the form of nuclear data required for analysis and interacts intensively with engineering aspects of design. Apart from design, physics aspects of operation, safety analysis, fuel management, fuel cycle studies etc are areas in which contributions are demanded from reactor physics. Efficient, economic and safe operation of any nuclear reactor requires good understanding of the physics aspects, in terms of operating regimes, available control and safety margins all the time. Due to such a central role, regulatory aspects of reactor physics are also important. For these tasks efficient and accurate modeling and computation tools are vital. The theoretical computations require experimental validation, which again requires special capabilities in nuclear detectors, electronics and analysis. The discussions in this theme meeting are on the recent advancements in reactor physics with reference to design, safety and operation. Papers relevant to INIS are indexed separately

  12. Report of Nuclear Fusion Reactor Engineering Research Meeting. 6. Advanced reactor engineering technology for nuclear fusion demonstration reactor

    International Nuclear Information System (INIS)

    This research meeting has been held every year, and the 6th meeting was held on January 17, 1995 at University of Tokyo. As the type of a demonstration reactor, tokamak type and helical type were set up, and the topics on the various subjects of their reactor engineering technology were presented, and active discussion was carried out. At the meeting, lectures were given on the reactor engineering technology required for a prototype reactor, the material technology supposed for a demonstration reactor, thermal-electric conversion and the direct electricity generation using Nernst effect, the advanced manufacturing technology of functional, structural materials, the application of high temperature superconductors to nuclear fusion reactors, the reactor engineering technology required for a helical type demonstration reactor, and tokamak demonstration reactor and the common technology of fission and fusion. This report is the summary of these lecture materials. The useful knowledges were obtained for considering the development of nuclear fusion reactor technology hereafter in this meeting. (K.I.)

  13. Development of smart nuclear instrumentation for reactors

    International Nuclear Information System (INIS)

    Variety of nuclear instruments are required for different applications in reactors such as reactor start-up, reactor protection and regulating system, area monitoring, failed fuel detection, stack monitoring etc. Attempts are made to develop a standardized microcomputer based hardware for configuring different types of instruments. PC architecture is chosen due to easy availability of components/boards and software. These instruments have dual redundant Network Interface Cards for connecting to a Primary Radiation Data LAN which in turn can be connected to Plant Information Bus through Gateways. These SMART instruments can be tested/calibrated through specific commands from remote computers connected over the LAN. This paper describes the various issues involved and the design details. (author)

  14. Dynamic detection of nuclear reactor core incident

    International Nuclear Information System (INIS)

    Surveillance, safety and security of evolving systems area challenge to prevent accident. The dynamic detection of a hypothetical and theoretical blockage incident in the Phenix nuclear reactor is investigated. Such an incident is characterized by abnormal temperature rises in the neighbourhood of the concerned reactor core assembly. The data set is the output temperature map of the reactor, it is provided by the Atomic Energy and Alternative Energies Commission (CEA). A real time approach is proposed, based on a sliding temporal window, it is divided into two steps. The first one behaves like a sieve, its function is to detect simultaneous temperature evolutions in a close neighbourhood which may induce a potential incident. When such evolutions are detected, the second step computes the temperature contrast between each assembly having these evolutions and its neighbourhood. This method permits to monitor the system evolution in real time while only few observations are required. Results are validated on various noisy realistic simulated perturbations. (authors)

  15. Nuclear energy center site survey reactor plant considerations

    Energy Technology Data Exchange (ETDEWEB)

    Harty, H.

    1976-05-01

    The Energy Reorganization Act of 1974 required the Nuclear Regulatory Commission (NRC) to make a nuclear energy center site survey (NECSS). Background information for the NECSS report was developed in a series of tasks which include: socioeconomic inpacts; environmental impact (reactor facilities); emergency response capability (reactor facilities); aging of nuclear energy centers; and dry cooled nuclear energy centers.

  16. Nuclear energy center site survey reactor plant considerations

    International Nuclear Information System (INIS)

    The Energy Reorganization Act of 1974 required the Nuclear Regulatory Commission (NRC) to make a nuclear energy center site survey (NECSS). Background information for the NECSS report was developed in a series of tasks which include: socioeconomic inpacts; environmental impact (reactor facilities); emergency response capability (reactor facilities); aging of nuclear energy centers; and dry cooled nuclear energy centers

  17. An overview of future sustainable nuclear power reactors

    OpenAIRE

    Andreas Poullikkas

    2013-01-01

    In this paper an overview of the current and future nuclear power reactor technologies is carried out. In particular, the nuclear technology is described and the classification of the current and future nuclear reactors according to their generation is provided. The analysis has shown that generation II reactors currently in operation all around the world lack significantly in safety precautions and are prone to loss of coolant accident (LOCA). In contrast, generation III reactors, which are ...

  18. Nuclear data requirements for fission reactor decommissioning

    International Nuclear Information System (INIS)

    The meeting was attended by 13 participants from 8 Member States and 2 International Organizations who reviewed the status of the nuclear data libraries and computer codes used to calculate the radioactive inventory in the reactor unit components for the decommissioning purposes. Nuclides and nuclear reactions important for determination of the radiation fields during decommissioning and for the final disposal of radioactive waste from the decommissioned units were identified. Accuracy requirements for the relevant nuclear data were considered. The present publication contains the text of the reports by the participants and their recommendations to the Nuclear Data Section of the IAEA. A separate abstract was prepared for each of these reports. Refs, figs and tabs

  19. Quality assurance of ECCS in nuclear reactors

    International Nuclear Information System (INIS)

    Size and shape of split or rupture in clad increases the whole body radiation exposure to the staff of the nuclear reactors. Suggests that a plant operating with 0.125 percent pin-hole fuel cladding defects showed a general five-fold increase in whole-body radiation exposure rates in some areas of the plant when compared to a sister plant with high-integrity fuel. Therefore Quality Assurance (QA) checks on Emergency Core Cooling System (ECCS) in Nuclear Reactors are very important to ensure minimum radiation hazard during Loss of Coolant Accident (LOCA). These checks will protect environment and public from radiation to great extent. The rate of rise of fuel temperature subsequent to LOCA should be lower than 5.5℃/s

  20. The regulatory challenges of decommissioning nuclear reactors

    International Nuclear Information System (INIS)

    Each nuclear power plant, fuel cycle facility and nuclear research and test facility that is operating today will eventually reach the end of its useful life and cease operation. During the period of its decommissioning, it is important to properly manage the health and environmental hazards and physical protection measures of the shutdown facility in order to protect the health and safety of the public and workers and to safeguard any nuclear materials. In this regard, the nuclear safety regulatory body is responsible for independently assuring that decommissioning activities are conducted safely, that radioactive materials and spent nuclear fuel are disposed of properly and that the site is in an acceptable end state. The purpose of this report is to describe the broad range of safety, environmental, organisational, human factors and public policy issues that may arise during the decommissioning of nuclear reactors and that the regulatory body should be prepared to deal with in the framework of its national regulatory system. The intended audience is primarily nuclear regulators, although the information and ideas may also be of interest to government authorities, environmental regulators, nuclear operating organisations, technical expert organisations and the general public. (author)