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Sample records for triga users neuherberg

  1. 3. European conference of TRIGA users. Papers and abstracts

    International Nuclear Information System (INIS)

    1974-01-01

    The Third European Conference of TRIGA Users was held October 29-31, 1974, in Neuherberg near Munich, Germany under the sponsorship of the Gesellschaft fur Strahlen and Umweltforschung mbH, Physikalischen-Technische Ableilung. The main topics were: experience in reactor operation, maintenance, measurements, fuel management and fuel performance, neutron physical experiments and other research programs

  2. CD3 TRIGA users conference

    International Nuclear Information System (INIS)

    2008-01-01

    The Sixteenth European TRIGA Users Conference was held in Pitesti, Romania, on 25-28 September, 2000, under the sponsorship of the Institute for Nuclear Research at Pitesti. The papers which follow in this document are presented in the same order as listed in the Conference Program. All papers which were received for publication (44) have been included. Those papers which were presented but not received for publication are presented in abstract form (4 papers). It was very interesting for the Conference attendees from the West to learn about the large scope of excellent work conducted in Romania, especially at the Institute of Nuclear Research in Pitesti. Similarly, it was fortunate that a large attendance of Romanian researchers (53) from many institutes, universities, and government agencies could attend the Conference and interact with their counterparts from outside Romania. The European TRIGA9 Owners' Group was fortunate to be hosted by the owners and users of the world's largest TRIGA reactor - the 14-MW Romanian research and test reactor. The Opening Session talk was given by Radu Berceanu, Minister of Industries and Commerce. It was followed by the following presentations: R and D - Support for Nuclear Power Development by Ioan Rotaru (General Manager of SNNE); Overview of TRIGA Reactor and other Programs at GENERAL ATOMICS by Junaid Razvi (General Manager TRIGA Reactor at GA); Development strategies connected to National Power and Energy Program by Mircea Ionescu (Director Nuclear Energy Department of M.I.C.); Contribution of INR R and D Programs to Sustain Peaceful and Safe utilization of Nuclear Energy in Romania by Constantin Gheorghiu (Scientific Deputy Director at SCN). A Technical visit to TRIGA Reactor at INR Pitesti took place. Opening Session was followed by five sessions dedicated to the following subjects: Session 1 (8 papers) - TRIGA reactors operation, repair and maintenance; Session 2 (10 papers) - Future developments and future goals of

  3. 10th European TRIGA users conference

    International Nuclear Information System (INIS)

    Anon.

    1988-01-01

    Abstracts of 46 papers on various aspects of Triga reactors (mainly Triga Mark 2 reactors) are given, according to the main headings: reactor operation and maintenance experience; new developments and improvements of Triga components and systems, including instrumentation; fuel and fuel management; safety aspects, licensing and radiation protection; experiments with Triga reactors; radiochemistry, radioisotope production and NAA; reactor physics. (qui)

  4. 10. European TRIGA users conference. Papers and abstracts

    International Nuclear Information System (INIS)

    1988-01-01

    The Tenth European TRIGA Users Conference was held in Vienna, September 14-16, 1988 under the sponsorship of the Atominstitut. The main areas of discussions were: Reactor operation and maintenance experiences; New developments and improvements of TRIGA components and systems, including instrumentation; Fuel and fuel management; Safety aspects, licensing and radiation protection; Experiments with TRIGA reactors; Radiochemistry, radioisotope production and NAA; and Reactor physics

  5. Decommissioning of the Neuherberg Research Reactor (FRN)

    International Nuclear Information System (INIS)

    Demmeler, M.; Rau, G.; Strube, D.

    1982-01-01

    The Neuherberg Research Reactor is of type TRIGA MARK III with 1 MW steady state power and pulsable up to 2000 MW. During more than ten years of operation 12000 MWh and 6000 reactor pulses had been performed. In spite of its good technical condition and of permanent safe operation without any failures, the decommissioning of the Neuherberg research reactor was decided by the GSF board of directors to save costs for maintaining and personnel. As the mode of decommissioning the safe enclosure was chosen which means that the fuel elements will be transferred back to the USA. All other radioactive reactor components will be enclosed in the reactor block. Procedures for licensing of the decommissioning, dismantling procedures and time tables are presented

  6. Proceedings of the 4. World TRIGA Users Conference

    International Nuclear Information System (INIS)

    2008-01-01

    This document gathers 30 presentations given at the 2008 Conference of the World TRIGA reactor Users. Most presentations are in the form of slides only, and few ones have an additional summary or are presented as an article only. All aspects of TRIGA-type reactors are approached, from upgrading to decommissioning, from radiotherapy to isotope production, from research program management to training, etc

  7. 7. European conference of TRIGA reactor users. Conference papers

    International Nuclear Information System (INIS)

    1982-01-01

    At the Seventh European Conference of TRIGA Users, held in September 1982, in Istanbul, Turkey, the following aspects are discussed: safety aspects of TRIGA reactors; developments and improvements; operating and maintenance experiences; applications; reactor calculations; fuel cycle aspects and research programs

  8. 6. European conference of TRIGA reactor users. Conference papers

    International Nuclear Information System (INIS)

    1980-01-01

    The Sixth European Conference of TRIGA Users was held in September 1980, in Mainz, Germany under the joint sponsorship of INTERATOM and the Institut fur Kernchemie. The main areas of discussions were: Fuel cycle aspects; New reactor developments and improvements; TRIGA applications; Operating and maintenance experiences and Instrumentation

  9. Proceedings of the 4. World TRIGA Users Conference

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2008-10-29

    This document gathers 30 presentations given at the 2008 Conference of the World TRIGA reactor Users. Most presentations are in the form of slides only, and few ones have an additional summary or are presented as an article only. All aspects of TRIGA-type reactors are approached, from upgrading to decommissioning, from radiotherapy to isotope production, from research program management to training, etc.

  10. Twelfth European TRIGA users conference. Papers and abstracts

    International Nuclear Information System (INIS)

    2008-01-01

    The Twelfth European TRIGA Users Conference was held in Pitesti, Romania, on September 28 - October 1, 1992, under the sponsorship of the Institute for Nuclear Research. The papers which follow in this document are presented in the same order as listed in the Conference Program. All papers which were received for publication (44) have been included. Those papers which were presented but not received for publication are presented in abstract form (3). The European TRIGA9 Owners' Group was fortunate to be hosted by the owners and users of the world's largest TRIGA reactor - the 14-MW Romanian research and test reactor. For too many years it has been impossible to enjoy open interactions with the Romanian researchers. By hosting the 1992 European TRIGA Users' Conference in Romania, the Romanians accomplished a breakthrough in the exchange of TRIGA reactor technology. It was very interesting for the Conference attendees from the West to learn about the large scope of excellent work conducted in Romania, especially at the Institute of Nuclear Research in Pitesti. Similarly, it was fortunate that a large attendance of Romanian researchers from many institutes, universities, and government agencies could attend the Conference and interact with their counterparts from outside Romania. The proceedings of the conference were structured onto the following 6 subject matters: - Opening Session and Introduction; - Session I, Operating and Maintenance Experience (10 papers); - Session II, Reactor Physics And Fuel Utilization (11 papers); - Session III, Instrumentation and Control (5 papers); - Session IV, Irradiation Facilities, Experimental Accessories (8 papers); - Session V, Applications, New Development of TRIGA Concept (6 papers). The document is completed with the abstracts of 3 contributions. A number of 19 experts from Austria, Germany, Italy, United States, Turkey, Morocco, England, Slovenia and Albania, that use TRIGA reactors, and Romania attended the conference. The

  11. 4. European conference of TRIGA users. Papers and abstracts

    International Nuclear Information System (INIS)

    1976-01-01

    The main topics of the Conference are: TRIGA operating experience special experimental and maintenance experience; activation analysis and isotope production; research programs; special instrumentation testing and other TRIGA specific topics

  12. 9. European TRIGA users' conference. Papers and abstracts

    International Nuclear Information System (INIS)

    1986-01-01

    Operation and maintenance experience, new developments and improvements of TRIGA reactors, fuel management, radiation protection, licensing, uses for research and isotope production are discussed at the Conference

  13. 12. U.S. TRIGA users conference. Papers and abstracts

    International Nuclear Information System (INIS)

    1990-01-01

    The Conference presentations were devoted to the following topics: new developments and improvements, including modifications of TRIGA reactors and equipment; experiments with TRIGA reactors (Neutron Radiography); radiochemistry, radioisotope production and beam irradiations (experiment applications, simulation); reactor physics - fuel utilization; reactor operation and maintenance experience; safety aspects, licensing and radiation protection

  14. Thirteen[th] European TRIGA users conference. Proceedings

    International Nuclear Information System (INIS)

    1994-01-01

    The conference covers topic related to TRIGA reactor, such as reactor operation and maintenance experience - safety aspects; research activities and applications; boron neutron capture therapy (BNCT) projects and other design operational and safety issues

  15. 3. world TRIGA users conference. Papers and abstracts

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2006-07-01

    The Conference is focused on TRIGA reactors operation and applications. The main topics are: use of the reactor as a research tool; inspection of spent fuel elements; integrity of fuel rods cladding checks; evaluation of corrosion of aluminum-base fuel cladding materials; Pitting behavior of Aluminum alloys; Monte Carlo simulation of TRIGA: reactivity worth, burnup, flux and power; irradiation facilities; thermal hydraulics analyses etc.

  16. 2nd world TRIGA users conference. Conference volume

    International Nuclear Information System (INIS)

    2004-01-01

    This conference was organized by the Atomic Institute of the Austrian Universities (University of Technology Vienna), it was devoted to present results in the operation of TRIGA research reactors. The main general topics were: a) reactor operation experience, b)neutron and solid state physics, c) radiochemistry and activation analysis, d) medical applications (boron neutron capture therapy, labeled compounds), e) reactor related experiments and calculations, f) waste management and decommissioning of TRIGA reactors. (nevyjel)

  17. 14. U.S. TRIGA users conference. Final program and summary of papers

    International Nuclear Information System (INIS)

    1994-01-01

    The following papers were presented at the Conference: Early Development and Use of the TRIGA Reactor; Results of the MCNP Analysis of 20/20 LEU Fuel for the Oregon State University TRIGA Reactor; Upgradeable 2MW TRIGA Reactor Design for the Morocco Nuclear Energy Center McClellan Nuclear Radiation Center TRIGA Reactor: Four Years of Operations

  18. 7. biennial U.S. TRIGA users' conference. Papers and abstracts

    International Nuclear Information System (INIS)

    1980-01-01

    The conference covers the following topics: new developments in the TRIGA system; uses of microprocessors in control and monitoring and measurement of TRIGA performance parameters; safeguards, emergency planning, reactor standards; research facilities, fuel tests and calculations; TRIGA reactor parameters: emergency training

  19. NRF TRIGA packaging

    International Nuclear Information System (INIS)

    Clements, M.D.

    1995-11-01

    Training Reactor Isotopes, General Atomics (TRIGA reg-sign) Reactors are in use at four US Department of Energy (DOE) complex facilities and at least 23 university, commercial, or government facilities. The development of the Neutron Radiography Facility (NRF) TRIGA packaging system began in October 1993. The Hanford Site NRF is being shut down and requires an operationally user-friendly transportation and storage packaging system for removal of the TRIGA fuel elements. The NRF TRIGA packaging system is designed to remotely remove the fuel from the reactor and transport the fuel to interim storage (up to 50 years) on the Hanford Site. The packaging system consists of a cask and an overpack. The overpack is used only for transport and is not necessary for storage. Based upon the cask's small size and light weight, small TRIGA reactors will find it versatile for numerous refueling and fuel storage needs. The NRF TRIGA packaging design also provides the basis for developing a certifiable and economical packaging system for other TRIGA reactor facilities. The small size of the NRF TRIGA cask also accommodates placing the cask into a larger certified packaging for offsite transport. The Westinghouse Hanford Company NRF TRIGA packaging, as described herein can serve other DOE sites for their onsite use, and the design can be adapted to serve university reactor facilities, handling a variety of fuel payloads

  20. 10. biennial U.S. TRIGA users' conference. Papers and abstracts

    International Nuclear Information System (INIS)

    1986-01-01

    The conference cover the following main topics for TRIGA reactors: reactor instrumentation and measurements of reactor parameters, reactor operation and modifications, design innovation and service works, fast neutron spectrum, fuel examination, neutron flux, heat transfer, accidents analysis, corrosion problems, fuel failures and fuel management, mechanical problems and maintenance

  1. Development of the user interface for visualization of the auxiliary systems of the TRIGA Mark III nuclear reactor

    International Nuclear Information System (INIS)

    Merced D, J. E.

    2016-01-01

    The Instituto Nacional de Investigaciones Nucleares (ININ) has a nuclear research reactor type swimming pool with movable core cooled and moderate with light water. The nominal maximum power of the reactor is 1 MW in steady-state operation and can be pulsed at a maximum power of 2,000 MW for approximately 10 milliseconds. This reactor is mainly used to study the effects of radiation on various materials and substances. In 2001 the new control console of the nuclear reactor was installed which was based on two digital computers, one computer controls the bar management mechanisms and the other the systems to the reactor operator. In 2004, the control computer was replaced and the software was updated. Within the modernization and/or updating of the TRIGA Mark III reactor of ININ, is intended (theme of this work) to develop the user interface for the visualization of the auxiliary systems, through a Man-Machine Interface module for the renewal process of the control console. The man-machine interface system to be developed will have communication with the programmable logic controllers that will be constantly monitored and controlled to obtain real-time variables of the reactor behavior. (Author)

  2. Development of the user interface for visualization of the auxiliary systems of the TRIGA Mark III nuclear reactor; Desarrollo de la interface de usuario para la visualizacion de los sistemas auxiliares del reactor nuclear Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Merced D, J. E.

    2016-07-01

    The Instituto Nacional de Investigaciones Nucleares (ININ) has a nuclear research reactor type swimming pool with movable core cooled and moderate with light water. The nominal maximum power of the reactor is 1 MW in steady-state operation and can be pulsed at a maximum power of 2,000 MW for approximately 10 milliseconds. This reactor is mainly used to study the effects of radiation on various materials and substances. In 2001 the new control console of the nuclear reactor was installed which was based on two digital computers, one computer controls the bar management mechanisms and the other the systems to the reactor operator. In 2004, the control computer was replaced and the software was updated. Within the modernization and/or updating of the TRIGA Mark III reactor of ININ, is intended (theme of this work) to develop the user interface for the visualization of the auxiliary systems, through a Man-Machine Interface module for the renewal process of the control console. The man-machine interface system to be developed will have communication with the programmable logic controllers that will be constantly monitored and controlled to obtain real-time variables of the reactor behavior. (Author)

  3. Development of the user Interface of digital simulation system of the operational parameters of the TRIGA IPR-R1 Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Felippe, Adriano de A.M.; Lage, Aldo M.F.; Mesquita, Amir Z.

    2017-01-01

    The development of simulation systems has been increasingly improved to ensure security and reliability to the systems being associated. Computational tools, simulation systems and programming languages increasingly allow the diversification of control systems. With increasing concern about monitoring the key parameters involved in chain reactions inside a nuclear reactor, new technologies are being developed to ensure operations safety. This paper deals with a practical application of a work that is being developed in the Center for the Development of Nuclear Technology - CDTN, which intends to simulate the operation of the TRIGA-IPR-R1 nuclear research reactor using the LabVIEW® software, evaluating the evolution of the neutron flux and other related events. In this paper, the visual interface of the reactor control table, developed through virtual instruments that allow, in a vast repertoire of tools, replicating the panels of the control table in modern screens that can be operated by a user of an analogous form, but still more practical and complete. Since the innovations developed for research reactors can be replicated in power reactors, and because of their lower operating and maintenance costs, projects in this area allow the development of several technologies

  4. TRIGA reactor characteristics

    International Nuclear Information System (INIS)

    Boeck, H.; Villa, M.

    2007-01-01

    This module describes the general design, characteristics and parameters of TRIGA reactors and fuels. It is recommended that most of this information should be incorporated into any reactor operator training program and, in many cases, the facility Safety Analysis Report. It is oriented to teach the basics of the physics and mechanical design of the TRIGA fuel as well as its unique operational characteristics and the differences between TRIGA fuels and others more traditional reactor fuels. (nevyjel)

  5. PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Masood, Z.

    2016-01-01

    The PUSPATI TRIGA Reactor is the only research reactor in Malaysia. This 1 MW TRIGA Mk II reactor first reached criticality on 28 June 1982 and is located at the Malaysian Nuclear Agency premise in Bangi, Malaysia. This reactor has been mainly utilised for research, training and education and isotope production. Over the years several systems have been refurbished or modernised to overcome ageing and obsolescence problems. Major achievements and milestones will also be elaborated in this paper. (author)

  6. TRIGA low enrichment fuel

    International Nuclear Information System (INIS)

    Gietzen, A.

    1993-01-01

    Sixty TRIGA reactors have been sold and the earliest of these are now passing twenty years of operation. All of these reactors use the uranium-zirconium hydride fuel (UZrH) which provides certain unique advantages arising out of its large prompt negative temperature coefficient, very low fission product release, and high temperature capability. Eleven of these Sixty reactors are conversions from plate fuel to TRIGA fuel which were made as a result of these advantages. With only a few exceptions, TRIGA reactors have always used low-enriched-uranium (LEU) fuel with an enrichment of 19.9%. The exceptions have either been converted from the standard low-enriched fuel to the 70% enriched FLIP fuel in order to achieve extended lifetime, or are higher powered reactors which were designed for long life using 93%-enriched uranium during the time when the use and export of highly enriched uranium (HEU) was not restricted. The advent of international policies focusing attention on nonproliferation and safeguards made the HEU fuels obsolete. General Atomic immediately undertook a development effort (nearly two years ago) in order to be in a position to comply with these policies for all future export sales and also to provide a low-enriched alternative to fully enriched plate-type fuels. This important work was subsequently partially supported by the U.S. Department of Energy. The laboratory and production tests have shown that higher uranium densities can be achieved to compensate for reducing the enrichment to 20%, and that the fuels maintain the characteristics of the very thoroughly proven standard TRIGA fuels. In May of 1978, General Atomic announced that these fuels were available for TRIGA reactors and for plate-type reactors with power levels up to 15 MW with GA's standard commercial warranty

  7. Development of the user Interface of digital simulation system of the operational parameters of the TRIGA IPR-R1 Nuclear Research Reactor; Desenvolvimento da interface para usuário do sistema digital de simulação dos parâmetros operacionais do reator nuclear de pesquisa Triga IPR-R1

    Energy Technology Data Exchange (ETDEWEB)

    Felippe, Adriano de A.M., E-mail: adrianoamfelippe@gmail.com [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil); Lage, Aldo M.F.; Mesquita, Amir Z., E-mail: aldo@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    The development of simulation systems has been increasingly improved to ensure security and reliability to the systems being associated. Computational tools, simulation systems and programming languages increasingly allow the diversification of control systems. With increasing concern about monitoring the key parameters involved in chain reactions inside a nuclear reactor, new technologies are being developed to ensure operations safety. This paper deals with a practical application of a work that is being developed in the Center for the Development of Nuclear Technology - CDTN, which intends to simulate the operation of the TRIGA-IPR-R1 nuclear research reactor using the LabVIEW® software, evaluating the evolution of the neutron flux and other related events. In this paper, the visual interface of the reactor control table, developed through virtual instruments that allow, in a vast repertoire of tools, replicating the panels of the control table in modern screens that can be operated by a user of an analogous form, but still more practical and complete. Since the innovations developed for research reactors can be replicated in power reactors, and because of their lower operating and maintenance costs, projects in this area allow the development of several technologies.

  8. TRIGA International, a new TRIGA fuel fabrication facility at CERCA

    International Nuclear Information System (INIS)

    Harbonnier, G.

    1997-01-01

    At the time when General Atomics expressed its intention to cease fuel fabrication on its site of San Diego, CERCA has been chosen to carry on the fabrication of TRIGA fuel. After negotiations in 1994 and 1995, a partnership 50%/50% was decided and on July 1995, a new company was founded, with the name TRIGA INTERNATIONAL SAS, head office in Paris and fuel fabrication facility at CERCA in Romans. The intent of this presentation is, after a short reminder about TRIGA fuel design and fabrication to describe the new facility with special emphasis on the safety features associated with the modification of existing fabrication buildings. (author)

  9. 3. TRIGA owners' conference. Papers and abstracts

    International Nuclear Information System (INIS)

    1974-01-01

    The TRIGA Owners' Conference III was held February 25-27, 1974, in Albuquerque, New Mexico. Seventy representatives were in attendance from 26 TRIGA facilities in the United States, Mexico, Puerto Rico, Indonesia, and from interested government agencies and industrial concerns. The main topics, discussed at the Conference were: TRIGA operating experiences; analytical and experimental methods; limits on effluents release for research reactor; and TRIGA modifications

  10. European TRIGA owners' conference. Papers and abstracts

    International Nuclear Information System (INIS)

    1970-01-01

    The conference covers the following topics, concerning TRIGA reactors: Experience in the Operation and Maintenance and utilization of TRIGA reactors; reactor upgrading; irradiation facilities; fuel management; air-concentration measurements; nuclear tests; use of TRIGA in nuclear medicine and biology; reactor design, fuel and performance; failures and other research activities

  11. Operating experience with the Cornell University TRIGA reactor

    International Nuclear Information System (INIS)

    Aderhold, H.C.

    1970-01-01

    As a result of our investigations, we believed the damage to be mechanical in origin and not to cladding failure. A new handling tool of modified design was put into service in July 1963, and since that time one element S/N 3075 has been dropped. This we believe was caused by operator error. At the request of prospective users, a high intensity, high energy gamma-ray irradiation facility has been added to the TRIGA equipment. This apparatus is simple to construct and use, either temporarily or permanently, with the TRIGA. Adjustment of relative neutron and gamma ray fluxes is possible by either shielding or changing rate of water flow. No attempt was made to improve performance by guiding water flow through the core, and higher yields should be obtainable by this means and by increasing the size of the holdup tank

  12. TRIGA forced shutdowns analysis

    International Nuclear Information System (INIS)

    Negut, Gheorghe; Laslau, Florica

    2008-01-01

    The need for improving the operation leads us to use new methods and strategies. Probabilistic safety assessments and statistical analysis provide insights useful for our reactor operation. This paper is dedicated to analysis of the forced shutdowns during the first reactor operation period, between 1980 to 1989. A forced shutdown data base was designed using data on forced shutdowns collected from the reactor operation logbooks. In order to sort out the forced shutdowns the records have the following fields: - current number, date, equipment failed, failure type (M for mechanical, E for electrical, D for irradiation device, U for human factor failure; - scram mode, SE for external scram, failure of reactor cooling circuits and/or irradiation devices, SR for reactor scram, exceeding of reactor nuclear parameters, SB for reactor scram by control rod drop, SM for manual scram required by the abnormal reactor status; - scram cause, giving more information on the forced shutdown. This data base was processed using DBase III. The data processing techniques are presented. To sort out the data, one of the criteria was the number of scrams per year, failure type, scram mode, etc. There are presented yearly scrams, total operation time in hours, total unavailable time, median unavailable time period, reactor availability A. There are given the formulae used to calculate the reactor operational parameters. There are shown the scrams per year in the 1980 to 1989 period, the reactor operation time per year, the reactor shutdown time per year and the operating time versus down time per year. Total number of scrams in the covered period was 643 which caused a reactor down time of 4282.25 hours. In a table the scrams as sorted on the failure type is shown. Summarising, this study emphasized some problems and difficulties which occurred during the TRIGA reactor operation at Pitesti. One main difficulty in creating this data base was the unstandardized scram record mode. Some times

  13. Extension of TRIGA reactor capabilities

    International Nuclear Information System (INIS)

    Gietzen, A.J.

    1980-01-01

    The first TRIGA reactor went into operation at 10 kW about 22 years ago. Since that time 55 TRIGAs have been put into operation including steady-state powers up to 14,000 kW and pulsing reactors that pulse to 20,000,000 kW. Five more are under construction and a proposal will soon be submitted for a reactor of 25,000 kW. Along with these increases in power levels (and the corresponding fluxes) the experimental facilities have also been expanded. In addition to the installation of new TRIGA reactors with enhanced capabilities many of the older reactors have been modified and upgraded. Also, a number of reactors originally fueled with plate fuel were converted to TRIGA fuel to take advantage of the improved technical and safety characteristics, including the ability for pulsed operation. In order to accommodate increased power and performance the fuel has undergone considerable evolution. Most of the changes have been in the geometry, enrichment and cladding material. However, more recently further development on the UZrH alloy has been carried out to extend the uranium content up to 45% by weight. This increased U content is necessary to allow the use of less than 20% enrichment in the higher powered reactors while maintaining longer core lifetime. The instrumentation and control system has undergone remarkable improvement as the electronics technology has evolved so rapidly in the last two decades. The information display and the circuitry logic has also undergone improvements for enhanced ease of operation and safety. (author)

  14. Higher power density TRIGA research reactors

    International Nuclear Information System (INIS)

    Whittemore, W.L.

    2008-01-01

    The uranium zirconium hydride (U-ZrH) fuel is the fundamental feature of the TRIGA family of reactors that accounts for its widely recognized safety, good performance, economy of operation, and its acceptance worldwide. Of the 65 TRIGA reactors or TRIGA fueled reactors, several are located in hospitals or hospital complexes and in buildings that house university classrooms. These examples are a tribute to the high degree of safety of the operating TRIGA reactor. In the early days, the majority of the TRIGA reactors had power levels in the range from 10 to 250 kW, many with pulsing capability. An additional number had power levels up to 1 MW. By the late 1970's, seven TRIGA reactors with power levels up to 2 MW had been installed. A reduction in the rate of worldwide construction of new research reactors set in during the mid 1970's but construction of occasional research reactors has continued until the present. Performance of higher power TRIGA reactors are presented as well as the operation of higher power density reactor cores. The extremely safe TRIGA fuel, including the more recent TRIGA LEU fuel, offers a wide range of possible reactor configurations. A long core life is assured through the use of a burnable poison in the TRIGA LEU fuel. In those instances where large neutron fluxes are desired but relatively low power levels are also desired, the 19-rod hexagonal array of small diameter fuel rods offers exciting possibilities. The small diameter fuel rods have provided extremely long and trouble-free operation in the Romanian 14 MW TRIGA reactor

  15. TRIGA reactor owners' seminar. Papers and abstracts

    International Nuclear Information System (INIS)

    1970-01-01

    The TRIGA Reactor Owners' Conference was planned with the aim of bringing together a group of persons interested in the ownership and operation of TRIGA reactors in the hope that an interchange of viewpoints, information, and experience would prove of mutual benefit

  16. Illinois advanced TRIGA startup testing

    International Nuclear Information System (INIS)

    Beck, G.P.

    1970-01-01

    A decision was made to carry out a major modification of the University of Illinois Mark II TRIGA reactor in order to obtain higher steady-state power levels and to increase the pulsing capabilities. The main aspects of the upgrading would be to change from aluminum clad to stainless steel clad fuel elements, a forced convective cooling system with sufficient capacity for operations to 3 MW, two pulse rods to allow reactivity insertions up to $4.60, a hexagonal grid pattern, and several new beamports

  17. Radioactivity of spent TRIGA fuel

    Energy Technology Data Exchange (ETDEWEB)

    Usang, M. D., E-mail: mark-dennis@nuclearmalaysia.gov.my; Nabil, A. R. A.; Alfred, S. L.; Hamzah, N. S.; Abi, M. J. B.; Rawi, M. Z. M.; Abu, M. P. [Reactor Department, Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2015-04-29

    Some of the oldest TRIGA fuel in the Malaysian Reaktor TRIGA PUSPATI (RTP) is approaching the limit of its end of life with burn-up of around 20%. Hence it is prudent for us to start planning on the replacement of the fuel in the reactor and other derivative activities associated with it. In this regard, we need to understand all of the risk associated with such operation and one of them is to predict the radioactivity of the fuel, so as to estimate the safety of our working conditions. The radioactivity of several fuels are measured and compared with simulation results to confirm the burnup levels of the selected fuels. The radioactivity measurement are conducted inside the water tank to reduce the risk of exposure and in this case the detector wrapped in plastics are lowered under water. In nuclear power plant, the general practice was to continuously burn the fuel. In research reactor, most operations are based on the immediate needs of the reactor and our RTP for example operate periodically. By integrating the burnup contribution for each core configuration, we simplify the simulation of burn up for each core configuration. Our results for two (2) fuel however indicates that the dose from simulation underestimate the actual dose from our measurements. Several postulates are investigated but the underlying reason remain inconclusive.

  18. 5. TRIGA owners' conference. Papers and abstracts

    International Nuclear Information System (INIS)

    1977-01-01

    The main topics of the Conference are: research reactor licensing and regulation; standards and public relations programs; operating problems and operating programs of research reactors; security requirements for TRIGA reactors

  19. Oregon State University TRIGA Reactor annual report

    International Nuclear Information System (INIS)

    Anderson, T.V.; Johnson, A.G.; Bennett, S.L.; Ringle, J.C.

    1979-01-01

    The use of the Oregon State University TRIGA Reactor during the year ending June 30, 1979, is summarized. Environmental and radiation protection data related to reactor operation and effluents are included

  20. Research activities at the TRIGA Mainz

    International Nuclear Information System (INIS)

    Eberhardt, K.; Trautmann, N.

    2002-01-01

    The TRIGA Mark II reactor of the Mainz University became first critical on August 3, 1965. It can be operated in the steady state mode with a maximum power of 100 kW and in the pulse mode with a peak power of 250 MW. The TRIGA Mainz is mainly used for neutron activation analysis, isotope production, basic research in nuclear chemistry and nuclear physics as well as for education and training

  1. Research with Neutrons at the TRIGA Mainz

    International Nuclear Information System (INIS)

    Hampel, Gabriele

    2008-01-01

    The TRIGA Mark II reactor at the Institut fuer Kernchemie of the Johannes Gutenberg-Universitaet in Mainz became first critical on August 3, 1965 and is still intensively used for basic research, applied science and educational purposes. Considering the past and future operation schedule and the low burn-up of the fuel elements (∼4 g 235 U/a), the reactor can be operated for at least the next decade taking into account the fresh fuel elements on stock and without changing spent fuels. The operation of the TRIGA Mainz has been extended very recently until the year 2020. The TRIGA Mainz can be operated in the steady state mode with a maximum power of 100 kWth and in the pulse mode with a peak power of 250 MWth. Until now, more than 16600 pulses have been carried out without any fuel failure. For irradiations the TRIGA Mainz has a central experimental tube, three pneumatic transfer systems and a rotary specimen rack with 40 positions which allows the irradiation of 80 samples at the same time. In addition, the TRIGA Mainz includes four horizontal beam ports penetrating the concrete shielding and extending inside the pool towards the reflector. A graphite thermal column provides a source of well-thermalized neutrons suitable for physical research or biological and medical irradiations. Important projects for the future of the TRIGA Mainz are the production of ultracold neutrons (UCN) and experiments with UCN, high precision mass measurements and laser spectroscopy of short-lived fission products (TRIGA-TRAP), the production of radionuclides for fast chemical separations, medical and radiopharmaceutical applications, and the use of the neutron activation analysis for the application in archeometry, solar energy technique, criminalistics and vine analysis. Furthermore, studies are performed to judge if the Boron Neutron Capture Therapy (BNCT) can be applied at the TRIGA Mainz for cancer treatment of liver metastases. Also, the reactor facility is used for the training

  2. Research with Neutrons at the TRIGA Mainz

    Energy Technology Data Exchange (ETDEWEB)

    Hampel, Gabriele [University of Mainz, Fritz-Strassmann-Weg 2, D-55128 Mainz (Germany)

    2008-10-29

    The TRIGA Mark II reactor at the Institut fuer Kernchemie of the Johannes Gutenberg-Universitaet in Mainz became first critical on August 3, 1965 and is still intensively used for basic research, applied science and educational purposes. Considering the past and future operation schedule and the low burn-up of the fuel elements ({approx}4 g {sup 235}U/a), the reactor can be operated for at least the next decade taking into account the fresh fuel elements on stock and without changing spent fuels. The operation of the TRIGA Mainz has been extended very recently until the year 2020. The TRIGA Mainz can be operated in the steady state mode with a maximum power of 100 kWth and in the pulse mode with a peak power of 250 MWth. Until now, more than 16600 pulses have been carried out without any fuel failure. For irradiations the TRIGA Mainz has a central experimental tube, three pneumatic transfer systems and a rotary specimen rack with 40 positions which allows the irradiation of 80 samples at the same time. In addition, the TRIGA Mainz includes four horizontal beam ports penetrating the concrete shielding and extending inside the pool towards the reflector. A graphite thermal column provides a source of well-thermalized neutrons suitable for physical research or biological and medical irradiations. Important projects for the future of the TRIGA Mainz are the production of ultracold neutrons (UCN) and experiments with UCN, high precision mass measurements and laser spectroscopy of short-lived fission products (TRIGA-TRAP), the production of radionuclides for fast chemical separations, medical and radiopharmaceutical applications, and the use of the neutron activation analysis for the application in archeometry, solar energy technique, criminalistics and vine analysis. Furthermore, studies are performed to judge if the Boron Neutron Capture Therapy (BNCT) can be applied at the TRIGA Mainz for cancer treatment of liver metastases. Also, the reactor facility is used for the

  3. 78 FR 26811 - Dow Chemical Company, Dow TRIGA Research Reactor; License Renewal for the Dow Chemical TRIGA...

    Science.gov (United States)

    2013-05-08

    ... COMMISSION Dow Chemical Company, Dow TRIGA Research Reactor; License Renewal for the Dow Chemical TRIGA Research Reactor; Supplemental Information and Correction AGENCY: Nuclear Regulatory Commission. ACTION... Chemical TRIGA Research Reactor,'' to inform the public that the NRC is considering issuance of a renewed...

  4. Optimum burnup of BAEC TRIGA research reactor

    International Nuclear Information System (INIS)

    Lyric, Zoairia Idris; Mahmood, Mohammad Sayem; Motalab, Mohammad Abdul; Khan, Jahirul Haque

    2013-01-01

    Highlights: ► Optimum loading scheme for BAEC TRIGA core is out-to-in loading with 10 fuels/cycle starting with 5 for the first reload. ► The discharge burnup ranges from 17% to 24% of U235 per fuel element for full power (3 MW) operation. ► Optimum extension of operating core life is 100 MWD per reload cycle. - Abstract: The TRIGA Mark II research reactor of BAEC (Bangladesh Atomic Energy Commission) has been operating since 1986 without any reshuffling or reloading yet. Optimum fuel burnup strategy has been investigated for the present BAEC TRIGA core, where three out-to-in loading schemes have been inspected in terms of core life extension, burnup economy and safety. In considering different schemes of fuel loading, optimization has been searched by only varying the number of fuels discharged and loaded. A cost function has been defined and evaluated based on the calculated core life and fuel load and discharge. The optimum loading scheme has been identified for the TRIGA core, the outside-to-inside fuel loading with ten fuels for each cycle starting with five fuels for the first reload. The discharge burnup has been found ranging from 17% to 24% of U235 per fuel element and optimum extension of core operating life is 100 MWD for each loading cycle. This study will contribute to the in-core fuel management of TRIGA reactor

  5. Dynamics of TRIGA-3 Salazar Reactor

    International Nuclear Information System (INIS)

    Gallardo S, L.F.

    1990-01-01

    The theoretical study of temporal behavior of a nuclear reactor is of great importance, since it allows to know, in advance, the conditions to which a reactor is going to be submitted. The reliability of two computer codes (AIREK-JEN and PLANKIN) designed to reproduce the temporal behavior of nuclear reactors, generally power reactors, when they are applied to reproduce the dynamic behavior of TRIGA-3 Salazar Reactor is analyzed. In the first chapters, the fundamental equations that solve this computer codes are deduced, and also the main characteristics of TRIGA-3 Salazar Reactor and the necessary data to run the programs are presented; later the results obtained with the computer codes and the experimental results reported in the operational logbook of the reactor are compared, with the result that such computer codes are applicable to the temporal study of TRIGA-3 Salazar Reactor. (Author)

  6. History, Development and Future of TRIGA Research Reactors

    International Nuclear Information System (INIS)

    2016-01-01

    Due to its particular fuel design and resulting enhanced inherent safety features, TRIGA reactors (Training, Research, Isotopes, General Atomics) constitute a ‘class of their own’ among the large variety of research reactors built world-wide. This publication summarizes in a single document the information on the past and present of TRIGA research reactors and presents an outlook in view of potential issues to be solved by TRIGA operating organizations in the near future. It covers the historical development and basic TRIGA characteristics, followed by utilization, fuel conversion and ageing management of TRIGA research reactors. It continues with issues and challenges, introduction to the global TRIGA research reactor network and concludes with future perspectives. The publication is complemented with a CD-ROM to illustrate the historical developments of TRIGA research reactors through individual facility examples and experiences

  7. Component failure data base of TRIGA reactors

    International Nuclear Information System (INIS)

    Djuricic, M.

    2004-10-01

    This compilation provides failure data such as first criticality, component type description (reactor component, population, cumulative calendar time, cumulative operating time, demands, failure mode, failures, failure rate, failure probability) and specific information on each type of component of TRIGA Mark-II reactors in Austria, Bangladesh, Germany, Finland, Indonesia, Italy, Indonesia, Slovenia and Romania. (nevyjel)

  8. Decommissioning of the Northrop TRIGA reactor

    International Nuclear Information System (INIS)

    Cozens, George B.; Woo, Harry; Benveniste, Jack; Candall, Walter E.; Adams-Chalmers, Jeanne

    1986-01-01

    An overview of the administrative and operational aspects of decommissioning and dismantling the Northrop Mark F TRIGA Reactor, including: planning and preparation, personnel requirements, government interfacing, costs, contractor negotiations, fuel shipments, demolition, disposal of low level waste, final survey and disposition of the concrete biological shielding. (author)

  9. [2. European] TRIGA owners' conference. Papers and abstracts

    International Nuclear Information System (INIS)

    1972-01-01

    The Second European TRIGA Owners' Conference was held in Pavia, Italy, September 1972. The meeting was organized by the University of Pavia Applied Nuclear Energy Laboratory (L.E.N.A.). Sixty-two attendees representing 12 TRIGA reactor centers in Europe, South America, and the United States were present at the Conference. The main areas of discussions were: Reactor operation and maintenance experience; Research programs and TRIGA technology development

  10. Aspek Proteksi Radiasi Pada Peningkatan Daya Reaktor Triga 2000

    OpenAIRE

    Oetami, Rini Heroe; Wiryosimin, S; Kurnia, E; Sofyan, S; Arifin, Z

    2002-01-01

    ASPEK PROTEKSI RADIASI FADA PENLNGKATAN DAYA REAKTOR TRIGA 2000. Reaktor TRIGA Mark II Bandung telah ditingkatkan dayanya dari 1 MW menjadi 2 MW dan saat ini namanya menjadi reaktor TRIGA 2000. Pekerjaan peningkatan daya meliputi tahap kegiatan pembongkaran, pemasangan dan uji nuklir yang memberikan risiko bahaya radiasi eksterna maupun interna. Bahaya radiasi dapat memberikan efek deterministik dan stokastik pada jaringan tubuh. Untuk mencegah efek deterministik dan menurunkan peluang terjad...

  11. NFR TRIGA package design review report

    International Nuclear Information System (INIS)

    Clements, M.D.

    1994-01-01

    The purpose of this document is to compile, present and document the formal design review of the NRF TRIGA packaging. The contents of this document include: the briefing meeting presentations, package description, design calculations, package review drawings, meeting minutes, action item lists, review comment records, final resolutions, and released drawings. This design review required more than two meeting to resolve comments. Therefore, there are three meeting minutes and two action item lists

  12. Monte Carlo modelling of TRIGA research reactor

    International Nuclear Information System (INIS)

    El Bakkari, B.; Nacir, B.; El Bardouni, T.; El Younoussi, C.; Merroun, O.; Htet, A.; Boulaich, Y.; Zoubair, M.; Boukhal, H.; Chakir, M.

    2010-01-01

    The Moroccan 2 MW TRIGA MARK II research reactor at Centre des Etudes Nucleaires de la Maamora (CENM) achieved initial criticality on May 2, 2007. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the neutronic analysis of the 2-MW TRIGA MARK II research reactor at CENM and validation of the results by comparisons with the experimental, operational, and available final safety analysis report (FSAR) values. The study was prepared in collaboration between the Laboratory of Radiation and Nuclear Systems (ERSN-LMR) from Faculty of Sciences of Tetuan (Morocco) and CENM. The 3-D continuous energy Monte Carlo code MCNP (version 5) was used to develop a versatile and accurate full model of the TRIGA core. The model represents in detailed all components of the core with literally no physical approximation. Continuous energy cross-section data from the more recent nuclear data evaluations (ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1, and JENDL-3.3) as well as S(α, β) thermal neutron scattering functions distributed with the MCNP code were used. The cross-section libraries were generated by using the NJOY99 system updated to its more recent patch file 'up259'. The consistency and accuracy of both the Monte Carlo simulation and neutron transport physics were established by benchmarking the TRIGA experiments. Core excess reactivity, total and integral control rods worth as well as power peaking factors were used in the validation process. Results of calculations are analysed and discussed.

  13. Monte Carlo modelling of TRIGA research reactor

    Science.gov (United States)

    El Bakkari, B.; Nacir, B.; El Bardouni, T.; El Younoussi, C.; Merroun, O.; Htet, A.; Boulaich, Y.; Zoubair, M.; Boukhal, H.; Chakir, M.

    2010-10-01

    The Moroccan 2 MW TRIGA MARK II research reactor at Centre des Etudes Nucléaires de la Maâmora (CENM) achieved initial criticality on May 2, 2007. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the neutronic analysis of the 2-MW TRIGA MARK II research reactor at CENM and validation of the results by comparisons with the experimental, operational, and available final safety analysis report (FSAR) values. The study was prepared in collaboration between the Laboratory of Radiation and Nuclear Systems (ERSN-LMR) from Faculty of Sciences of Tetuan (Morocco) and CENM. The 3-D continuous energy Monte Carlo code MCNP (version 5) was used to develop a versatile and accurate full model of the TRIGA core. The model represents in detailed all components of the core with literally no physical approximation. Continuous energy cross-section data from the more recent nuclear data evaluations (ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1, and JENDL-3.3) as well as S( α, β) thermal neutron scattering functions distributed with the MCNP code were used. The cross-section libraries were generated by using the NJOY99 system updated to its more recent patch file "up259". The consistency and accuracy of both the Monte Carlo simulation and neutron transport physics were established by benchmarking the TRIGA experiments. Core excess reactivity, total and integral control rods worth as well as power peaking factors were used in the validation process. Results of calculations are analysed and discussed.

  14. RIA Analysis of Unprotected TRIGA Reactor

    Directory of Open Access Journals (Sweden)

    M.H. Altaf

    2017-07-01

    Full Text Available An RIA (reactivity initiated accident analysis has been carried out for the TRIGA Mark II research reactor considering both step and ramp reactivity ranges within 0.5 % dk/k (< $1 to 2.0 % dk/k (>$2. The insertion time was set at 10 s. Based on the fact that a reactor becomes unprotected if scram does not work at the event of danger, to define unprotected conditions, the time to actuate scram (trip was taken as close to total simulation time. In this long duration of scram inactivity, it is obtained from the present analysis that the reactor remained safe to up to 1.8 % dk/k ($2.57 for step reactivity and 1.99 % dk/k ($2.84 for ramp reactivity. In addition to negative temperature coefficient of reativity, probably the longer time of reactivity insertion keeps TRIGA safe even at larger magnitudes of reactivity during unprotected reactor transients. Coupled point kinetics, neutronics, and thermal hydraulics code EUREKA-2/R has been utilized for this work. It appears that EUREKA-2/RR predicts the sequence of unprotected transient scenario of TRIGA core with good approximation and the results will definitely be helpful for the reactor operators.

  15. Industrial and commercial applications for a Triga reactor

    International Nuclear Information System (INIS)

    Green, D.

    1986-01-01

    The Physics and Radioisotope Services Group of ICI operates a Triga Reactor in support of a commercial, Industrial Radioisotope Technology Service. The technical and commercial development of this business is discussed in the context of operating a Triga Reactor in an Industrial Environment. (author)

  16. Enrichment measurement in TRIGA type fuels

    International Nuclear Information System (INIS)

    Aguilar H, F.; Mazon R, R.

    2001-05-01

    The Department of Energy of the United States of North America, through the program 'Idaho Operations Nuclear Spent Fuel Program' of the Idaho National Engineering and Environmental Laboratory (INEEL), in Idaho Falls; Idaho USA, hires to Global Technologies Inc. (GTI) to develop a prototype device of detection enrichment uranium (DEU Detection of Enrichment of Uranium) to determine quantitatively the enrichment in remainder U-235 in a TRIGA fuel element at the end of it useful life. The characteristics of the prototype developed by GTI are the following ones: It allows to carry out no-destructive measurements of TRIGA type fuel. Easily transportable due to that reduced of it size. The determination of the enrichment (in grams of U-235) it is obtained with a precision of 5%. The National Institute of Nuclear Research (ININ), in its facilities of the Nuclear Center of Mexico, it has TRIGA type fuel of high and low enrichment (standard and FLIP) fresh and with burnt, it also has the infrastructure (hot cells, armor-plating of transport, etc) and qualified personnel to carry out the necessary maneuvers to prove the operation of the DEU prototype. For this its would be used standard type fuel elements and FLIP, so much fresh as with certain burnt one. In the case of the fresh fuels the measurement doesn't represent any risk, the fuels before and after the measurement its don't contain a quantity of fission products that its represent a radiological risk in its manipulation; but in the case of the fuels with burnt the handling of the same ones represents an important radiological risk reason why for its manipulation it was used the transport armor-plating and the hot cells. (Author)

  17. Operating experiences at the Finnish TRIGA reactor

    International Nuclear Information System (INIS)

    Salmenhaara, Seppo

    1988-01-01

    The Finnish TRIGA reactor has been in operation since March 1962. There are still 57 original Al-clad fuel elements in the core. So far we have had only two fuel cladding failures in 1981 and 1988. The first one was an Al-clad element and the second one a SS-clad. The low rate of fuel cladding failures has made it possible to use continuously also the Al-clad fuel elements. Although some conventional irradiations of certain type have been repeated successfully tens of times, new and unexpected incidents can still take place. As an example an event of a leaking irradiation capsule is described

  18. Arkansas Tech University TRIGA nuclear reactor

    International Nuclear Information System (INIS)

    Sankoorikal, J.; Culp, R.; Hamm, J.; Elliott, D.; Hodgson, L.; Apple, S.

    1990-01-01

    This paper describes the TRIGA nuclear reactor (ATUTR) proposed for construction on the campus of Arkansas Tech University in Russellville, Arkansas. The reactor will be part of the Center for Energy Studies located at Arkansas Tech University. The reactor has a steady state power level of 250 kW and can be pulsed with a maximum reactivity insertion of $2.0. Experience gained in dismantling and transporting some of the components from Michigan State University, and the storage of these components will be presented. The reactor will be used for education, training, and research. (author)

  19. Safety Management at PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Ligam, A.S.; Zarina Masood; Ahmad Nabil Abdul Rahim

    2011-01-01

    Adequate safety measures and precautions, which follow relevant safety standards and procedures, should be in place so that personnel safety is assured. Nevertheless, the public, visitor, contractor or anyone who wishes to enter or be in the reactor building should be well informed with the safety measures applied. Furthermore, these same elements of safety are also applied to other irradiation facilities within the premises of Nuclear Malaysia. This paper will describes and explains current safety management system being enforced especially in the TRIGA PUSPATI Reactor (RTP) namely radiation monitoring system, safety equipment, safe work instruction, and interconnected internal and external health, safety and security related departments. (author)

  20. Decontamination of TRIGA Mark II reactor, Indonesia

    International Nuclear Information System (INIS)

    Suseno, H.; Daryoko, M.; Goeritno, A.

    2002-01-01

    The TRIGA Mark II Reactor in the Centre for Research and Development Nuclear Technique Bandung has been partially decommissioned as part of an upgrading project. The upgrading project was carried out from 1995 to 2000 and is being commissioned in 2001. The decommissioning portion of the project included disassembly of some components of the reactor core, producing contaminated material. This contaminated material (grid plate, reflector, thermal column, heat exchanger and pipe) will be sent to the Decontamination Facility at the Radioactive Waste Management Development Centre. (author)

  1. Some possibilities of utilisation of TRIGA reactors in the future

    International Nuclear Information System (INIS)

    Stegnar, Peter; Byrne, Anthony R.

    2008-01-01

    Full text. In this presentation, some possibilities for the future use of TRIGA reactors are discussed. The use and practical applications of neutron activation analysis, both in instrumental and radiochemical analysis, is presented based on the experience of the Institute's TRIGA Mark II Reactor in Ljubljana. The limited use of isotope production for medicine and industry is also discussed as well as some other potential applications, i.e. prompt gamma neutron activation analysis and an approach to BNCT (Boron Neutron Capture Therapy). The possibility of using TRIGA reactors for training in nuclear safety, radiological protection and other relevant fields of science and technology is also addressed in the presentation

  2. The role of TRIGA reactors in pure and applied nuclear research outside the United States in the last couple of years

    International Nuclear Information System (INIS)

    Rollier, M.A.

    1972-01-01

    The last two years trend of research in European TRIGA plants, which reported to the 1970 TRIGA Owners' Conference in Helsinki, is presented. The report discusses also new TRIGA plants in Europe, 1971-72; Research at TRIGA plants and new TRIGA reactors outside Europe and the U.S.A., 1971-72; Safety, health, environment, egomania and TRIGA reactors

  3. Quality control in nuclear medicine. Addendum to 'Quality assurance in nuclear medicine', World Health Organization, 1982. Following a training workshop in Neuherberg, Munich, Federal Republic of Germany, November 23-30, 1983

    International Nuclear Information System (INIS)

    Cradduck, T.D.; Busemann Sokole, E.

    2006-01-01

    Based on the results of a workshop, sponsored by the German Federal Ministry of the Interior (BMI) and the World Health Organization (WHO), held in Neuherberg, Munich, 1983, a revised concept of quality control (QC) in nuclear medicine is proposed. In a previous document published by the WHO in 1982, a set of rigorous daily, weekly, monthly, quarterly and half-quarterly QC protocols was recommended. The revised concept suggested by the 1983 workshop participants is that of a simple, total system, QC test that can be applied on a frequent, probably daily, basis. Any deviation of the results of such a test from the accepted norm will require further investigation and decision trees can then be applied to localize faults

  4. Operation of the low power reactor assembly in the University of Illinois TRIGA bulk shielding facility

    International Nuclear Information System (INIS)

    Beck, G.P.; Thayer, G.

    1972-01-01

    The Low Power Reactor Assembly (LOPRA) is a rectangular assembly of TRIGA fuel elements located in the bulk shielding facility of the Illinois Advanced TRIGA Reactor (TRIGA). The LOPRA is coupled to the TRIGA by a thermal column that extends from the TRIGA core to the bulk shielding facility. The LOPRA will be utilized for studying steady-state and transient coupling between the two reactors. Spatial variations in the flux at the LOPRA will be measured for different operating conditions of the TRIGA. These measurements will be made with both systems in low power critical states or in subcritical states

  5. The new supervisory system of the ENEA'S TRIGA

    International Nuclear Information System (INIS)

    Bessenyei, Z.; Businaro, T.; Rabbani, M.I.

    1986-01-01

    color changing is based on the comparison of the measured and modelled variables. The multilevel flow modelling is function oriented. In the point of view of modelling it is supposed that the system to be modelled was planned, constructed and is operated to achieve some goal(s), which was or were established by the designer and/or the user of the system. The goal(s) of the system can be achieved if its appropriate functional parts are working well. The functional parts (aggregates or nodes) can work well if their physical realisations are available. It is easy to recognise, in complex systems this triple can be repeated several times, from the main goal of the plant to the equipments, through the subsystems. The nodes are composed of elementary flow functions, which describe the mass/energy flow within an aggregate. A set of predefined flow functions are used to describe a complex plant. Choosing appropriate graphic symbols the nodes can be presented as hierarchically ordered pictures, where some color changing is used for qualitative diagnosis. Its base is the comparison of the process model and the measured variables which belong to the node considered. Up to now, the mimic presentation and a preliminary version of the multilevel flow model of the TRIGA (together with its equation model) have been developed at the ENEA TIB ISP INFO laboratory. During this work some methodological question were answered, and some new solution were introduced relative to the earlier mfm applications. This paper is a short summary of the work, and with respect to the topic of the conference, the way of the information presentation is emphasized. (author)

  6. The TRIGA reactor as chemistry apparatus

    International Nuclear Information System (INIS)

    Miller, G.E.

    1974-01-01

    At the Irvine campus of the University of California, the Mark I, 250 kilowatt TRIGA reactor is used as a regular teaching and research tool by the Department of Chemistry which operates the reactor. Students are introduced to radiochemistry and activation analysis in undergraduate laboratory courses and the relation of nuclear to chemical phenomena is emphasized even in Freshman chemistry. Special peripheral items have been developed for use in graduate and undergraduate research, including a fast pneumatic transfer system for studying short-lived isotopes and arrangements for irradiations at low temperatures. These and other unique features of a purely chemically oriented operation will be discussed and some remarks appended with regard to the merits of a low budget operation. (author)

  7. Status of the TRIGA-LASER experiment

    Energy Technology Data Exchange (ETDEWEB)

    Gorges, C., E-mail: chgorges@uni-mainz.de; Kaufmann, S., E-mail: s.kaufmann@uni-mainz.de [Technische Universität Darmstadt, Institut für Kernphysik (Germany); Geppert, Ch. [Johannes Gutenberg-Universität Mainz, Institut für Kernchemie (Germany); Krämer, J. [Technische Universität Darmstadt, Institut für Kernphysik (Germany); Sánchez, R. [GSI Helmholtzzentrum für Schwerionenforschung (Germany); Nörtershäuser, W. [Technische Universität Darmstadt, Institut für Kernphysik (Germany)

    2017-11-15

    We report on the newly developed control system called TRITON and the new data acquisition called TILDA as well as on improved isotope shift measurements of the isotopes {sup 40,42,44,48}Ca in the 4s 2S1/2 → 4p 2P3/2 (D2) transition at the TRIGA-LASER experiment in Mainz using collinear laser spectroscopy. Well known isotope shift measurements in the 4s 2S1/2 → 4p 2P1/2 (D1) transition act as calibration points to reduce the uncertainties in the D2-line to provide reference values for the determination of nuclear charge radii and quadrupole moments of neutron rich calcium isotopes at COLLAPS.

  8. The JSI TRIGA Mark II Reactor, Slovenia

    International Nuclear Information System (INIS)

    Smodiš, B.; Snoj, L.

    2016-01-01

    The TRIGA Mark II research reactor at Jožef Stefan Institute (JSI) in Ljubljana Slovenia has been playing an important role in developing nuclear technology and safety culture in Slovenia. In the 1970s-1980s the reactor was extensively used for production of radioactive isotopes. Since then the reactor has been extensively used for various applications, such as: irradiation of various samples, training and education, verification and validation of nuclear data and computer codes, testing and development of experimental equipment used for core physics tests at a nuclear power plant. The paper describes the aforementioned activities proving that even such small reactors are still indispensable in nuclear science and technology. (author)

  9. TRIGA reactor dynamics: Frequency response tests

    International Nuclear Information System (INIS)

    Firat, Coskun; Geckinli, Melih

    2008-01-01

    In this work, the results of frequency response tests conducted on ITU TRIGA Reactor are presented. To conduct the experiments, a special 'micro control rod' and its submersible stepping-motor drive mechanism was designed and constructed. The experiments cover a frequency range of 0.002 - 2 Hz., and 0.02, 4, 200 kW nominal power levels. Zero-power and at-power reactivity to % power transfer functions are presented as gain, and phase shift vs. frequency diagrams. Low power response is in close agreement with the point reactor zero-power transfer function. Response at 200 kW is studied with the help of a Nyquist diagram, and found to be stable. An elaboration on the main features of the feedback mechanism is also given. Power to reactivity feedback was measured to be just about 1.5 cent / % power change. (authors)

  10. Pneumatic transport systems for TRIGA reactors

    International Nuclear Information System (INIS)

    Bolton, John A.

    1970-01-01

    Main parameters and advantages of pneumatically operated systems, primarily those operated by gas pressure are discussed. The special irradiation ends for the TRIGA reactor are described. To give some idea of the complexity of some modern systems, the author presents the large system currently operating at the National Bureau of Standards in Washington. In this system, 13 stations are located throughout the radiochemistry laboratories and three irradiation ends are located in the reactor, which is a 14-megawatt unit. The system incorporates practically every fail-safe device possible, including ball valves located on all capsule lines entering the reactor area, designed to close automatically in the event of a reactor scram, and at that time capsules within the reactor would be diverted by means of switches located on the inside of the reactor wall. The whole system is under final control of a permission control panel located in the reactor control room. Many other safety accessories of the system are described

  11. TRIGA out of core gamma irradiation facility

    International Nuclear Information System (INIS)

    Rant, J.; Pregl, G.

    1988-01-01

    A possibility to irradiate extended objects in a gamma field inside the shielding water tank and above the core of operating TRIGA Mark II Reactor has been investigated. The irradiation cask is shielded with Cd cover to filter out thermal neutrons. The dose rate of the gamma field strongly depends on the distance of the irradiation position above the core. At 25 cm above the core, the gamma dose rate is 2.2 Gy/s and epithermal neutron flux is ∼ 8.10 6 ncm -2 s -1 ∼ 3 as measured by TLD (CaF 2 : Mn) dosimeters and Au foils respectively. Tentative applications of the gamma irradiation facility are in the studies of radiation induced accelerated aging and within the Nuclear Power Plant Equipment Qualification Program (EQP). A complete characterization of the neutron spectrum and optimization of the 7 radiation field within the cask has still to be performed. (author)

  12. Inspection of the TRIGA Reactor Tank

    International Nuclear Information System (INIS)

    Jerman, Jernej; Lesnjak, Andrej; Snoj, Luka; Smodiš, Borut

    2016-01-01

    Nuclear components are under strict supervision of operators and safety authorities. The Reactor Centre of the Jožef Stefan Institute decided to make an inspection of its TRIGA Mark II research reactor to verify the conditions for long-term future operation within the on-going periodic safety review. Two main inspection methods were used: ultrasonic and visual inspection. Ultrasonic inspection was selected to prove that there is no significant reduction of wall thickness anywhere in the tank. The inspection confirmed that the reactor tank has not been degraded or corroded. In the future such inspection will take place every 10 years within the periodic safety review in order to monitor every 10 years the reactor tanks condition. (author)

  13. The TRIGA in virtual classroom for training

    International Nuclear Information System (INIS)

    Plata M, A. C.; Morales S, J. B.; Salazar S, E.

    2008-01-01

    The research nuclear reactors have been fundamental part in the evolution of the nuclear power plants and they have been used in the training for the obtaining of operation licenses of radioactive facilities. For purposes of training of professionals in nuclear engineering, it is interesting to know the benefit that can be obtained by means of the virtual representation of a research nuclear reactor TRIGA, with which they are possible the practice to be realized them but common that to date they are carried out in different nuclear facilities of training throughout the world. The simulation has become a valuable tool in the personal preparation, having obtained ambient and very approximate situations to the reality. The physical models of kinetics of neutrons, heat transfer, Cherenkov effect, dynamics of the xenon, as well as the virtual instrumentation is contemplated in this development. The instrumentation and control panels of a research reactor, failures waited for in the use of this equipment, physical consequences to instruments, virtual personnel and facilities, as well as the administrative and legal aspects that it requires to meet an authorized operator, must be available and they are considered in the first virtual approach. The obtaining of the reactor time constant comprises of the mathematical model that provides to the operator of a direct way the knowledge of the changes of power. The coolant and moderator are modeled as well as the retardations that appear in the measurements and controls that can be introduced from the virtual console. In the simulator the four possible states of operation of the TRIGA can be had. At the moment also the monitoring can be realized and control in remote form, thus the control and supervision interface for the remote operation will be analyzed in their benefits and possible risks in the instruction processes. (Author)

  14. Fuel management for TRIGA reactor operators

    International Nuclear Information System (INIS)

    Totenbier, R.E.; Levine, S.H.

    1980-01-01

    One responsibility of the Supervisor of Reactor Operations is to follow the TRIGA core depletion and recommend core loading changes for refueling and special experiments. Calculations required to analyze such changes normally use digital computers and are extremely difficult to perform for one who is not familiar with computer language and nuclear reactor diffusion theory codes. The TRICOM/SCRAM program developed to perform such calculations for the Penn State TRIGA Breazeale Reactor (PSBR), has a very simple input format and is one which can be used by persons having no knowledge of computer codes. The person running the program need not understand computer language such as Fortran, but should be familiar with reactor core geometry and effects of loading changes. To further simplify the input requirements but still allow for all of the studies normally needed by the reactor operations supervisor, the options required for input have been isolated to two. Given a master deck of computer cards one needs to change only three cards; a title card, core energy history information card and one with core changes. With this input, the program can provide individual fuel element burn-up for a given period of operation and the k eff of the core. If a new loading is desired, a new master deck containing the changes is also automatically provided. The life of a new core loading can be estimated by feeding in projected core burn-up factors and observing the resulting loss in individual fuel elements. The code input and output formats have now been made sufficiently convenient and informative as to be incorporated into a standard activity for the Reactor Operations Supervisor. (author)

  15. Development of a simulator for design and test of power controllers in a TRIGA Mark III reactor; Desarrollo de un simulador para diseno y prueba de controladores de potencia en un reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Perez M, C.; Benitez R, J.S.; Lopez C, R. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2003-07-01

    The development of a simulator that uses the Runge-Kutta-Fehlberg method to solve the model of the punctual kinetics of a nuclear research reactor type TRIGA. The simulator includes an algorithm of power control of the reactor based on the fuzzy logic, a friendly graphic interface which responds to the different user's petitions and that it shows numerical and graphically the results in real time. The user can modify the demanded power and to visualize the dynamic behavior of the one system. This simulator was developed in Visual Basic under an open architecture with which its will be prove different controllers for its analysis. (Author)

  16. Development of a simulator for design and test of power controllers in a TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Perez M, C.; Benitez R, J.S.; Lopez C, R.

    2003-01-01

    The development of a simulator that uses the Runge-Kutta-Fehlberg method to solve the model of the punctual kinetics of a nuclear research reactor type TRIGA. The simulator includes an algorithm of power control of the reactor based on the fuzzy logic, a friendly graphic interface which responds to the different user's petitions and that it shows numerical and graphically the results in real time. The user can modify the demanded power and to visualize the dynamic behavior of the one system. This simulator was developed in Visual Basic under an open architecture with which its will be prove different controllers for its analysis. (Author)

  17. Enrichment measurement in TRIGA type fuels; Medicion de enriquecimiento en combustibles tipo Triga

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F.; Mazon R, R. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)

    2001-05-15

    The Department of Energy of the United States of North America, through the program 'Idaho Operations Nuclear Spent Fuel Program' of the Idaho National Engineering and Environmental Laboratory (INEEL), in Idaho Falls; Idaho USA, hires to Global Technologies Inc. (GTI) to develop a prototype device of detection enrichment uranium (DEU Detection of Enrichment of Uranium) to determine quantitatively the enrichment in remainder U-235 in a TRIGA fuel element at the end of it useful life. The characteristics of the prototype developed by GTI are the following ones: It allows to carry out no-destructive measurements of TRIGA type fuel. Easily transportable due to that reduced of it size. The determination of the enrichment (in grams of U-235) it is obtained with a precision of 5%. The National Institute of Nuclear Research (ININ), in its facilities of the Nuclear Center of Mexico, it has TRIGA type fuel of high and low enrichment (standard and FLIP) fresh and with burnt, it also has the infrastructure (hot cells, armor-plating of transport, etc) and qualified personnel to carry out the necessary maneuvers to prove the operation of the DEU prototype. For this its would be used standard type fuel elements and FLIP, so much fresh as with certain burnt one. In the case of the fresh fuels the measurement doesn't represent any risk, the fuels before and after the measurement its don't contain a quantity of fission products that its represent a radiological risk in its manipulation; but in the case of the fuels with burnt the handling of the same ones represents an important radiological risk reason why for its manipulation it was used the transport armor-plating and the hot cells. (Author)

  18. Operational reactivity considerations of Texas A and M TRIGA

    International Nuclear Information System (INIS)

    Randall, John D.; Schad, Douglas R.

    1970-01-01

    The Texas A and M University Nuclear Science Center Reactor was converted from a 100 kW MTR type core to a modified 1 MW TRIGA Mark III core in August, 1968. The TRIGA core has solved many of the low neutron flux problems without requiring a forced cooled reactor. However, the TRIGA core has presented some operational problems as the results of its reactivity requirements. Several solutions have been proposed to eliminate the problem of burnup. The use of fueled followers is presently being implemented at Texas A and M, but this is only a temporary solution. A final solution is the use of FLIP fuel which is presently under development. It incorporates a burnable poison in the fuel matrix, which results in a predicted lifetime of 9 MW-years. A decision to use this fuel will depend on the results of the PRNC conversion

  19. History, Development and Future of TRIGA Research Reactors. Companion CD-ROM

    International Nuclear Information System (INIS)

    2016-01-01

    Due to its particular fuel design and resulting enhanced inherent safety features, TRIGA reactors (Training, Research, Isotopes, General Atomics) constitute a ‘class of their own’ among the large variety of research reactors built world-wide. This publication summarizes in a single document the information on the past and present of TRIGA research reactors and presents an outlook in view of potential issues to be solved by TRIGA operating organizations in the near future. It covers the historical development and basic TRIGA characteristics, followed by utilization, fuel conversion and ageing management of TRIGA research reactors. It continues with issues and challenges, introduction to the global TRIGA research reactor network and concludes with future perspectives. This CD-ROM illustrates the historical developments of TRIGA research reactors through individual facility examples and experiences

  20. Designing a TRIGA for environmental compatibility

    International Nuclear Information System (INIS)

    Profio, A.E.

    1972-01-01

    A 10 kW TRIGA Mark I teaching reactor is programmed for the new Engineering Unit 2 building on the main campus of UCSB. A special effort has been made to keep radioactivity in the reactor room air and exhaust air to a minimum. The regular pneumatic transfer system is being replaced by an inert-gas pressure system, or a closed-cycle air system, to avoid argon activation and release. Other experimental facilities will be filled with water or carbon dioxide, or sealed, to prevent release of argon-41. The remaining source of Ar-41, from the air dissolved in the pool water, cannot be completely eliminated. The campus has a policy of not discharging liquid radioactive wastes to the sewer. Experience at other installations indicates that the ion exchanger in the purification loop is effective in removing activated corrosion products or impurities, and aside from the short-lived N-16 and some Ar-41, radioactivity of the pool water should be quite low. Solid wastes, which would include the spent ion exchange resin, are collected by an independent firm for burial in a special nuclear disposal site in Nevada. Other important considerations in siting the UCSB reactor are the proximity to Santa Barbara Municipal Airport, and the seismic activity. We plan to project the fuel elements against damage by falling aircraft or parts by addition of a thick steel cover over the pool. Earthquakes have been taken into account in the design of the building, and especially the below-ground, reinforced-concrete, aluminum-lined pool

  1. Probabilistic Safety Assessment Of It TRIGA Mark-II Reactor

    International Nuclear Information System (INIS)

    Ergun, E; Kadiroglu, O.S.

    1999-01-01

    The probabilistic safety assessment for Istanbul Technical University (ITU) TRIGA Mark-II reactor is performed. Qualitative analysis, which includes fault and event trees and quantitative analysis which includes the collection of data for basic events, determination of minimal cut sets, calculation of quantitative values of top events, sensitivity analysis and importance measures, uncertainty analysis and radiation release from fuel elements are considered

  2. Irradiation routine in the IPR-R1 Triga reactor

    International Nuclear Information System (INIS)

    Maretti Junior, F.

    1980-01-01

    Information about irradiations in the IPR-R1 TRIGA reactor and procedures necessary for radioisotope solicitation are presented All procedures necessary for asking irradiation in the reactor, shielding types, norms of terrestrial and aerial expeditions, payment conditions, and catalogue of disposable isotopes with their respective saturation activities are described. (M.C.K.)

  3. Small Angle Neutron Scattering instrument at Malaysian TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shukri Mohd; Razali Kassim; Zal Uyun Mahmood [Malaysian Inst. for Nuclear Technology Research (MINT), Bangi, Kajang (Malaysia); Shahidan Radiman

    1998-10-01

    The TRIGA MARK II Research reactor at the Malaysian Institute for Nuclear Research (MINT) was commissioned in July 1982. Since then various works have been performed to utilise the neutrons produced from this steady state reactor. One of the project involved the Small Angle Neutron Scattering (SANS). (author)

  4. The TRIGA in virtual classroom for training; El TRIGA en aula virtual para entrenamiento

    Energy Technology Data Exchange (ETDEWEB)

    Plata M, A. C.; Morales S, J. B.; Salazar S, E. [UNAM, DEPFI Campus Morelos, Jiutepec Morelos 62550 (Mexico)]. e-mail: yoyuclof@hotmail.com

    2008-07-01

    The research nuclear reactors have been fundamental part in the evolution of the nuclear power plants and they have been used in the training for the obtaining of operation licenses of radioactive facilities. For purposes of training of professionals in nuclear engineering, it is interesting to know the benefit that can be obtained by means of the virtual representation of a research nuclear reactor TRIGA, with which they are possible the practice to be realized them but common that to date they are carried out in different nuclear facilities of training throughout the world. The simulation has become a valuable tool in the personal preparation, having obtained ambient and very approximate situations to the reality. The physical models of kinetics of neutrons, heat transfer, Cherenkov effect, dynamics of the xenon, as well as the virtual instrumentation is contemplated in this development. The instrumentation and control panels of a research reactor, failures waited for in the use of this equipment, physical consequences to instruments, virtual personnel and facilities, as well as the administrative and legal aspects that it requires to meet an authorized operator, must be available and they are considered in the first virtual approach. The obtaining of the reactor time constant comprises of the mathematical model that provides to the operator of a direct way the knowledge of the changes of power. The coolant and moderator are modeled as well as the retardations that appear in the measurements and controls that can be introduced from the virtual console. In the simulator the four possible states of operation of the TRIGA can be had. At the moment also the monitoring can be realized and control in remote form, thus the control and supervision interface for the remote operation will be analyzed in their benefits and possible risks in the instruction processes. (Author)

  5. Application of RELAP/SCDAPSIM to TRIGA system thermal hydraulic analysis

    International Nuclear Information System (INIS)

    Allison, C.M.; Hohorst, J.K.; Quamrul Huda, Md.

    2010-01-01

    The RELAP/SCDAPSIM/MOD4.0 code, designed to predict the behavior of reactor systems during normal and accident conditions, is being developed as part of the international SCDAP Development and Training Program (SDTP). RELAP/SCDAPSIM/MOD4.0, which is the first version of RELAP5 completely rewritten to FORTRAN 90/95/2000 standards, uses publicly available RELAP5 and SCDAP models in combination with advanced programming and numerical techniques and other SDTP-member modeling/user options. This paper describes the development of a representative input model for the 3MW TRIGA research reactor at AERE Bangladesh, describes the testing and qualification of the model using MOD4.0 advanced input checking and graphical display options, and then presents representative results for selected calculations. (author)

  6. Five years of operating the TRIGA Mainz reactor

    International Nuclear Information System (INIS)

    Benedict, Georg

    1970-01-01

    Considerable obstacles had to be surmounted before TRIGA MAINZ, first TRIGA reactor built in Germany, reached initial criticality in 1965. Subsequent five years' operation did not raise any major problems. The facility has proven quite reliable and particularly well suited for the purposes of the nuclear chemistry research program pursued at Mainz University. Extensive use is made of the pulse mode of operation. As a result, fuel elements are obviously somewhat overstressed, even though most pulses performed are of the 1.50 dollar size. Maximum licensed steady state power of 100 kW till now has met the requirements of most experiments. However, efforts are in progress to improve irradiation conditions by increasing the reactor power to 300 kW. (author)

  7. Operation of the AFRRI-TRIGA reactor for radiobiology research

    International Nuclear Information System (INIS)

    McKenzie, Jerry L.; Joseph, Richard J.

    1976-01-01

    The AFRRI TRIGA Mark F reactor has, since 1965, played an important role in the radiobiological research conducted at AFRRI. The purpose of this paper is to describe the experimental facilities and dosimetry techniques used in support of these radiobiology programs and to present a brief summary of current programs. The AFRRI TRIGA exposure facilities were designed to be especially suitable for biological irradiations and, as a result, exhibit several important and interesting features. Reactor dosimetry in the exposure rooms includes the measurement of neutron and gamma reactor spectra, dose, dose rate, neutron-to-gamma ratio and depth dose distribution. These parameters are routinely measured at AFRRI using: threshold detectors; paired 50 cm 3 ; 0.5 cm 3 and miniature 0.05 cm 3 ionization chambers; tissue equivalent phantoms; and specimen cadavers. (author)

  8. Operation and maintenance of 1MW PUSPATI TRIGA reactor

    International Nuclear Information System (INIS)

    Adnan Bokhari; Mohammad Suhaimi Kassim

    2006-01-01

    The Malaysian Research Reactor, Reactor TRIGA PUSPATI (RTP) has been successfully operated for 22 years for various experiments. Since its commissioning in June 1982 until December 2004, the 1MW pool-type reactor has accumulated more than 21143 hours of operation, corresponding to cumulative thermal energy release of about 14083 MW-hours. The reactor is currently in operation and normally operates on demand, which is normally up to 6 hours a day. Presently the reactor core is made up of standard TRIAGA fuel element consists of 8.5 wt%, 12 wt% and 20 wt% types; 20%-enriched and stainless steel clad. Several measures such as routine preventive maintenance and improving the reactor support systems have been taken toward achieving this long successful operation. Besides normal routine utilization like other TRIGA reactors, new strategies are implemented for effective increase in utilization. (author)

  9. Isothermal temperature reactivity coefficient measurement in TRIGA reactor

    International Nuclear Information System (INIS)

    Zagar, T.; Ravnik, M.; Trkov, A.

    2002-01-01

    Direct measurement of an isothermal temperature reactivity coefficient at room temperatures in TRIGA Mark II research reactor at Jozef Stefan Institute in Ljubljana is presented. Temperature reactivity coefficient was measured in the temperature range between 15 o C and 25 o C. All reactivity measurements were performed at almost zero reactor power to reduce or completely eliminate nuclear heating. Slow and steady temperature decrease was controlled using the reactor tank cooling system. In this way the temperatures of fuel, of moderator and of coolant were kept in equilibrium throughout the measurements. It was found out that TRIGA reactor core loaded with standard fuel elements with stainless steel cladding has small positive isothermal temperature reactivity coefficient in this temperature range.(author)

  10. Seed irradiation facilities at TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Najzer, M.

    1972-01-01

    Fast neutrons and gamma-rays with their high and low LET respectively are excellent complementary tools for investigation of the effect of different types of mutations. TRIGA Irradiation Facility and Thermal Column Irradiation Facility were designed and installed for the first time in the TRIGA tank and thermal column respectively. The basic idea of design was the use of depleted uranium as gamma-ray and thermal neutron shield and simultaneously as thermal to fast neutron converter. Low LET radiation, due to direct and thermal neutron capture gamma-rays, is strongly attenuated while fast neutron flux is increased. GIF is made of a cadmium tube inserted in a graphite block. It is located in the central thermal column channel. The basic idea is to convert thermal neutrons to gamma-rays by capture in the cadmium

  11. Research projects at the TRIGA-reactor Vienna

    International Nuclear Information System (INIS)

    Boeck, H.; Buchberger, T.; Buchtela, K.; Hammer, J.; Miksovsky, A.; Veider, A.; Weber, H.W.; Zugarek, G.

    1986-01-01

    In 1985 the thermalizing column was modified to a beam tube with a conical collimator for neutron radiography. A highly sophisticated sample and cassette changer will be constructed in the next months. The central channel of the thermal column is also used for neutron radiography especially for small objects. The four beam tubes of the TRIGA-reactor are intensively used for neutron spectroscopy, small angle scattering, neutron interferometry and investigations of magnetic structures with polarized neutrons. The neutron activation installation in the piecing beam tube is permanently used for various sample analysis using a ultrafast pneumatic transfer system. In addition to these experiments directly related to the TRIGA-reactor other research projects are carried out, some of them under an IAEA research contract which are mostly focused towards nuclear safeguards such as the magnetic scanning of power reactor fuel assemblies or the laser surveillance system of spent fuel pools. (author)

  12. Status of the TRIGA shipments to the INEEL from Asia

    International Nuclear Information System (INIS)

    Tyacke, M.; George, W.; Petrasek, A.; Stump, R.C.; Patterson, J.

    1997-01-01

    This paper will report on preparations being made for returning Training, Research, Isotope, General Atomics (TRIGA) foreign research reactor (FRR) spent fuel from South Korea and Indonesia to the Idaho National Engineering and Environmental Laboratory (INEEL). The roles of US Department of Energy, INEEL, and NAC International in implementing a safe shipment are provided. Special preparations necessitated by making a shipment through a west coast port of the US to the INEEL will be explained. The institutional planning and actions needed to meet the unique political and operational environment for making a shipment from Asia to INEEL will be discussed. Facility preparation at both the INEEL and the FRRs is discussed. Cask analysis needed to properly characterize the various TRIGA configurations, compositions, and enrichments is discussed. Shipping preparations will include an explanation of the integrated team of spent fuel transportation specialists, and shipping resources needed to retrieve the fuel from foreign research reactor sites and deliver it to the INEEL

  13. Development of the Fuel Element Database of PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Nurhayati Ramli; Naim Syauqi Hamzah; Nurfazila Husain; Yahya Ismail; Mat Zin Mat Husin; Mohd Fairus Abd Farid

    2015-01-01

    Since June 28th, 1982, the PUSPATI TRIGA Reactor (RTP) operates safely with an accumulated energy release of about 17,200 MWhr, which corresponds to about 882 g of uranium burn-up. The reactor core has been reconfigured 15th times. Presently, there are 111 TRIGA fuel elements in the core, which 66 of the fuel elements are from the initial criticality while the rest of the fuel elements have been added to compensate the uranium consumption. As 59 % of the fuel elements are older than 30 years old, it is necessary to put the history of every fuel element in a database for easy access of the fuel element movement, inspection results history and integrity status. This paper intends to describe how the fuel element database is developed and related formulae used in determining the RTP fuel element elongation. (author)

  14. TRIGA FUEL PHASE I AND II CRITICALITY CALCULATION

    International Nuclear Information System (INIS)

    Angers, L.

    1999-01-01

    The purpose of this calculation is to characterize the criticality aspect of the codisposal of TRIGA (Training, Research, Isotopes, General Atomic) reactor spent nuclear fuel (SNF) with Savannah River Site (SRS) high-level waste (HLW). The TRIGA SNF is loaded into a Department of Energy (DOE) standardized SNF canister which is centrally positioned inside a five-canister defense SRS HLW waste package (WP). The objective of the calculation is to investigate the criticality issues for the WP containing the five SRS HLW and DOE SNF canisters in various stages of degradation. This calculation will support the analysis that will be performed to demonstrate the viability of the codisposal concept for the Monitored Geologic Repository (MGR)

  15. Thermal spectra of the TRIGA Mark III reactor; El espectro termico del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Macias B, L.R.; Palacios G, J. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    1998-07-01

    The diffraction phenomenon is gave in observance of the well known Bragg law in crystalline materials and this can be performance by mean of X-rays, electrons and neutrons among others, which allows to do inside the field of each one of these techniques the obtaining of measurements focussed at each one of them. For the present work, it will be mentioned only the referring to X-ray and neutron techniques. The X-ray diffraction due to its properties just it does measurements which are known in general as superficial measurements of the sample material but for the properties of the neutrons, this diffraction it explores in volumetric form the sample material. Since the neutron diffraction process depends lots of its intensity, then it is important to know the neutron source spectra that in this case is supplied by the TRIGA Mark III reactor. Within of diffraction techniques a great number of them can be found, however some of the traditional will be mentioned such as the identification of crystalline samples, phases identification and the textures measurement. At present this last technique is founded on the dot of a minimum error and the technique of phases identification performs but not compete with that which is obtained by mean of X-rays due to this last one has a major resolution. (Author)

  16. Reactivity worth of the TRIGA Mark III central irradiation facility

    International Nuclear Information System (INIS)

    Ruby, L.; El-Zeftawy, M.M.; Elangovan, R.

    1972-01-01

    This note reports on a calculation of the reactivity worth of the central irradiation facility, made specifically for the Berkeley TRIGA Mark-Ill Reactor. The results are in good qualitative agreement with the previous calculations (West and Larsen) and in excellent quantitative agreement with a single point which obtained for the Mark III. The results also indicate that the l-dimensional FOG calculation seriously underestimates the leakage from the completely-voided central region

  17. Operational aspects of TRIGA shipment from South Korea to INEEL

    International Nuclear Information System (INIS)

    Shelton, Thomas

    1999-01-01

    A shipment of 299 irradiated TRIGA fuel elements was made from South Korea to the United States in July 1998. The shipment was from two facilities in Korea and was received at the Irradiated Fuel Storage Facility (IFSF) at the Idaho National Engineering and Environmental Laboratory (INEEL). Fuel types shipped included aluminum and stainless steel clad standard fuel elements, instrumented and fuel follower control elements, as well as FLIP elements and failed fuel elements. Modes of transport included truck, rail and ship. (author)

  18. Neutron Beam Utilization At The TRIGA Mark II Reactor Vienna

    International Nuclear Information System (INIS)

    Villa, M.; Boeck, H.; Buchelt, R.J.; Koerner, S.; Rauch, H.

    2008-01-01

    A review is given about the research activities around the 250 kW TRIGA reactor Vienna, which are adequate to other neutron sources of comparable or bigger size. The topics selected for presentation range from neutron radiography, materials irradiation, neutron small-angle scattering, neutron activation analysis, neutron polarisation to neutron interferometry. It is the aim of this presentation to stimulate programs for more efficient use around TRIGA research reactors with neutron flux densities of 10 13 cm -2 s -1 at the centre of the reactor core. One briefly describes the experimental facilities installed at the 250 kW TRIGA reactor of the Austrian Universities in Vienna and presented a great part of the current research activities performed with them. Most of the techniques and experiments presented are adequate for implementation to other reactors of similar or even higher power. Those technologies which require extremely specialized know-how not generally available at every research institute are not treated here or are just mentioned without any further details. It is common knowledge that due to the relatively low neutron fluxes of such reactors one of the most important applications of neutron scattering on condensed matter, namely the study of atomic and molecular dynamics of solids and liquids, a priori must remain out of consideration. However, this does not mean that it is in general impossible to develop new or to improve existing techniques for such experiments at TRIGA research reactors. In fact such developing work has always been a crucial point of the research efforts in the variety of fields of applied and fundamental neutron physics. On the other hand, a small reactor facility is optimally suited to perform neutron activation analysis due to the rather short transfer distances of the sample into the reactor core. (authors)

  19. Neutron beam utilization at the TRIGA Mark II reactor Vienna

    International Nuclear Information System (INIS)

    Villa, M.; Boeck, H.; Ismail, S.; Koerner, S.; Baron, M.; Hainbuchner, M.; Badurek, G.; Buchelt, R.J.

    1999-01-01

    A review is given about the research activities around the 250 kw TRIGA reactor Vienna, which are adequate to other neutron sources of comparable or bigger size. The topics selected for presentation range from neutron radiography, materials irradiation, neutron small-angle scattering, neutron activation analysis, neutron polarization to neutron interferometry. It is the aim of this presentation to stimulate programs for more efficient use around TRIGA research reactors with neutron flux densities of 1013 cm-2a-1 at the center of the reactor core. We briefly describe the experimental facilities installed at the 250 kw TRIGA reactor of the Austrian Universities in Vienna and present a great part of the current research activities performed with them. We believe that most of the techniques and experiments presented here are adequate for implementation to other reactors of similar or even higher power. Those technologies which require extremely specialized know-how not generally available at every research Inst.e will not be treated here or are just mentioned without any further details.(author)

  20. Coordination of motor systems of the control bars of TRIGA Mark III reactor, through the use of a PLC; Coordinacion de los sistemas motrices de las barras de control del reactor Triga Mark III, mediante el uso de un PLC

    Energy Technology Data Exchange (ETDEWEB)

    Castro R, P. R.

    2016-07-01

    The use of programmable logic controllers (PLCs) has a wide field in the realization of automatic systems, since is sought that the form of control is easy for any user. In this work, the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) is intended to operate with a programming code in PLC for the automation of the control rods, having uniform wear according to the power required by the user. On the other hand, is proposed to develop an HMI graphical interface for communication via Ethernet, allowing supervision during the reactor operation process and greater protection of operators during reactor startup. The accuracy of the new actuators, as well as their durability, will allow a good performance of the reactor for many years to come. (Author)

  1. TRIGA-SPEC: A setup for mass spectrometry and laser spectroscopy at the research reactor TRIGA Mainz

    Energy Technology Data Exchange (ETDEWEB)

    Ketelaer, J. [Institut fuer Physik, Johannes Gutenberg-Universitaet Mainz, Staudinger Weg 7, D-55128 Mainz (Germany)], E-mail: ketela@uni-mainz.de; Kraemer, J. [Institut fuer Kernchemie, Johannes Gutenberg-Universitaet Mainz, Fritz-Strassmann-Weg 2, D-55128 Mainz (Germany); Beck, D. [Gesellschaft fuer Schwerionenforschung mbH, Planckstrasse 1, D-64291 Darmstadt (Germany); Blaum, K. [Institut fuer Physik, Johannes Gutenberg-Universitaet Mainz, Staudinger Weg 7, D-55128 Mainz (Germany); Gesellschaft fuer Schwerionenforschung mbH, Planckstrasse 1, D-64291 Darmstadt (Germany); Max-Planck-Institut fuer Kernphysik, Saupfercheckweg 1, D-69117 Heidelberg (Germany); Block, M. [Gesellschaft fuer Schwerionenforschung mbH, Planckstrasse 1, D-64291 Darmstadt (Germany); Eberhardt, K. [Institut fuer Kernchemie, Johannes Gutenberg-Universitaet Mainz, Fritz-Strassmann-Weg 2, D-55128 Mainz (Germany); Eitel, G.; Ferrer, R. [Institut fuer Physik, Johannes Gutenberg-Universitaet Mainz, Staudinger Weg 7, D-55128 Mainz (Germany); Geppert, C. [Institut fuer Kernchemie, Johannes Gutenberg-Universitaet Mainz, Fritz-Strassmann-Weg 2, D-55128 Mainz (Germany); Gesellschaft fuer Schwerionenforschung mbH, Planckstrasse 1, D-64291 Darmstadt (Germany); George, S. [Institut fuer Physik, Johannes Gutenberg-Universitaet Mainz, Staudinger Weg 7, D-55128 Mainz (Germany); Gesellschaft fuer Schwerionenforschung mbH, Planckstrasse 1, D-64291 Darmstadt (Germany); Herfurth, F. [Gesellschaft fuer Schwerionenforschung mbH, Planckstrasse 1, D-64291 Darmstadt (Germany); Ketter, J.; Nagy, Sz.; Neidherr, D. [Institut fuer Physik, Johannes Gutenberg-Universitaet Mainz, Staudinger Weg 7, D-55128 Mainz (Germany); Neugart, R. [Institut fuer Kernchemie, Johannes Gutenberg-Universitaet Mainz, Fritz-Strassmann-Weg 2, D-55128 Mainz (Germany)] (and others)

    2008-09-01

    The research reactor TRIGA Mainz is an ideal facility to provide neutron-rich nuclides with production rates sufficiently large for mass spectrometric and laser spectroscopic studies. Within the TRIGA-SPEC project, a Penning trap as well as a beamline for collinear laser spectroscopy are being installed. Several new developments will ensure high sensitivity of the trap setup enabling mass measurements even on a single ion. Besides neutron-rich fission products produced in the reactor, also heavy nuclides such as {sup 235}U or {sup 252}Cf can be investigated for the first time with an off-line ion source. The data provided by the mass measurements will be of interest for astrophysical calculations on the rapid neutron-capture process as well as for tests of mass models in the heavy-mass region. The laser spectroscopic measurements will yield model-independent information on nuclear ground-state properties such as nuclear moments and charge radii of neutron-rich nuclei of refractory elements far from stability. TRIGA-SPEC also serves as a test facility for mass and laser spectroscopic experiments at SHIPTRAP and the low-energy branch of the future GSI facility FAIR. This publication describes the experimental setup as well as its present status.

  2. Coordination of motor systems of the control bars of TRIGA Mark III reactor, through the use of a PLC

    International Nuclear Information System (INIS)

    Castro R, P. R.

    2016-01-01

    The use of programmable logic controllers (PLCs) has a wide field in the realization of automatic systems, since is sought that the form of control is easy for any user. In this work, the TRIGA Mark III reactor of Instituto Nacional de Investigaciones Nucleares (ININ) is intended to operate with a programming code in PLC for the automation of the control rods, having uniform wear according to the power required by the user. On the other hand, is proposed to develop an HMI graphical interface for communication via Ethernet, allowing supervision during the reactor operation process and greater protection of operators during reactor startup. The accuracy of the new actuators, as well as their durability, will allow a good performance of the reactor for many years to come. (Author)

  3. Sensitivity analysis to a RELAP5 nodalization developed for a typical TRIGA research reactor

    International Nuclear Information System (INIS)

    Reis, Patrícia A.L.; Costa, Antonella L.; Pereira, Claubia; Silva, Clarysson A.M.; Veloso, Maria Auxiliadora F.

    2012-01-01

    Highlights: ► We investigated how much the code results are affected by the code user. ► Two essential modifications were made on a previously validated nodalization. ► We used the RELAP5 code to predict the results. ► Results highlight the necessity of sensitivity analysis to have the ideal modeling. - Abstract: The main aim of this work is to identify how much the code results are affected by the code user in the choice of, for example, the number of thermal hydraulic channels in a nuclear reactor nodalization. To perform this, two essential modifications were made on a previously validated nodalization for analysis of steady-state and forced recirculation off transient in the IPR-R1 TRIGA research reactor. Experimental data were taken as reference to compare the behavior of the reactor for two different types of modeling. The results highlight the necessity of sensitivity analysis to obtain the ideal modeling to simulate a specific system.

  4. A Study on the Distribution of Dose at PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Nurhayati Ramli; Norfarizan Mohd Said; Zaredah Hashim; Muhammad Khairol Ariff Mustafa

    2016-01-01

    Full text: The main functions of TRIGA Reactor are to provide Training, Research and Isotope Production. In recent years, RTP actively facilitate various experiments for instance operational of sample irradiation and beam utilizations. Besides, RTP also support education and training program due to overwhelming interest from local and international students and researchers. All these activities involve series of experiments utilizing the RTP that indirectly cause radiation exposure to the users. Concerning the amount of radiation exposed, the aim of this paper is to investigate the operational dose exposure for the safe of RTP operation. The data is collected from 5 channels of Area Radiation Monitoring (ARM) since 2014 until Jun 2016. The installed ARMs are located at the most potential radiation release whereby four channels are mounted at the wall closed to the beam ports and one channel is located at the reactor pool top. The results observed shows that the exposure dose did not exceed the radiation release limit of 5 μSv/ hr as stipulated in RTP Safety Analysis Report (SAR). Moreover, the measured parameters also provide information to users regarding releases of radioactivity and any existing danger from radiation. (author)

  5. Transport and reprocessing of irradiated TRIGA fuel elements

    International Nuclear Information System (INIS)

    Staake, Theo R.

    1980-01-01

    This paper is intended to provide a review of the transport of irradiated TRIGA fuel elements, and the selection of the transport casks. The information presented is based on the experience TRANSNUKLEAR GmbH has gained over the last 14 years in the transport and reprocessing of irradiated MTR fuel elements. During this period over 2000 fuel elements were delivered to various destinations for reprocessing (about 1000 fuel elements - to European reprocessing plants (Eurochemic in Belgium, and CEA Marcoule in France), and, since 1977, over 1000 fuel elements - to the US-DOE Savannah River reprocessing plant)

  6. Study of Physical Protection System at PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Ligam, A.S.; Ina, I.; Zarina Masood

    2016-01-01

    Physical protection program at PUSPATI TRIGA Reactor (RTP) which is located at Nuklear Malaysia, Bangi Complex has been strengthened and upgraded from time to time to accommodate current situation needs. However, there is always room for improvement. Hence, study have been made to look deeper into physical protection components such as delay systems, external sensors, PPS intruder alarm sensors, use of video system, personnel security or insider threats, access control operation system operation rules and security culture that may need to take into consideration. (author)

  7. Integrated management system implementation strategy for PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Phongsakorn Prak Tom; Shaharum Ramli; Mohamad Azman Che Mat Isa; Shahirah Abdul Rahman; Mohd Zaid Mohamed; Mat Zin Mat Husin; Nurfazila Husain; Mohamad Puad Abu

    2012-01-01

    Integrated Management System (IMS) designed to fulfil the requirements integrates safety, health, environmental, security, quality and economic elements. PUSPATI TRIGA Reactor (RTP) is currently implementing the Quality Assurance Program (QAP) and looking toward implementation of IMS. This paper discussed the implementation strategy of IMS for RTP. There are nine steps of IMS implementation strategy. In implementation of IMS, Gantt chart is useful project management tool in managing the project frame work. IMS is intended as a tool to enable the continuous development of safety culture and achieve higher safety levels. (author)

  8. Radiological monitoring related to the operation of PUSPATI's Triga Reactor

    International Nuclear Information System (INIS)

    Fatimah Mohamad Amin; Mohamad Yusof Mohamad Ali; Lau How Mooi; Idris Besar.

    1983-01-01

    Reactor operation is one of the main activities carried out at the Tun Ismail Atomic Research Centre (PUSPATI) which requires radiological monitoring. This paper describes the programme for radiological monitoring which is related to the operation of the 1 MW Triga MK II research reactor which was commissioned in July, 1982. This programme includes monitoring of the radiation and contamination levels of the reactor and its associated facilities and environmental monitoring of PUSPATI's site and its environs. The data presented in this paper covers the period between 1982 to 1983 which includes both the pre-operational and operational phases of the monitoring programme. (author)

  9. Physical and transportation requirements for a FLIP fueled TRIGA

    International Nuclear Information System (INIS)

    Johnson, A.G.; Ringle, J.C.; Anderson, T.V.

    1977-01-01

    Several major changes to the OSTR Physical Security Plan were required by the NRC prior to the August 1976 receipt and installation of a new core consisting entirely of FLIP fuel. The general nature of these changes will be reviewed along with several decisions we faced during their implementation. At the previous TRIGA Owners' Conference in Salt Lake City, Utah, we reported on Oregon's regulatory program for research reactor emergency response planning and physical security. The latter program was of particular interest to us in light of the projected FLIP fuel shipments. The impact of the State's program for physical security of FLIP fuel during transportation will be presented. (author)

  10. Development of the ageing management database of PUSPATI TRIGA reactor

    International Nuclear Information System (INIS)

    Ramli, Nurhayati; Tom, Phongsakorn Prak; Husain, Nurfazila; Farid, Mohd Fairus Abd; Ramli, Shaharum; Maskin, Mazleha; Adnan, Amirul Syazwan; Abidin, Nurul Husna Zainal

    2016-01-01

    Since its first criticality in 1982, PUSPATI TRIGA Reactor (RTP) has been operated for more than 30 years. As RTP become older, ageing problems have been seen to be the prominent issues. In addressing the ageing issues, an Ageing Management (AgeM) database for managing related ageing matters was systematically developed. This paper presents the development of AgeM database taking into account all RTP major Systems, Structures and Components (SSCs) and ageing mechanism of these SSCs through the system surveillance program

  11. Development of the ageing management database of PUSPATI TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ramli, Nurhayati, E-mail: nurhayati@nm.gov.my; Tom, Phongsakorn Prak; Husain, Nurfazila; Farid, Mohd Fairus Abd; Ramli, Shaharum [Reactor Technology Centre, Malaysian Nuclear Agency, MOSTI, Bangi, 43000 Kajang, Selangor (Malaysia); Maskin, Mazleha [Science Program, Faculty of Science and Technology, Universiti Kebangsaan Malaysia, Selangor (Malaysia); Adnan, Amirul Syazwan; Abidin, Nurul Husna Zainal [Faculty of Petroleum and Renewable Energy Engineering, Universiti Teknologi Malaysia (Malaysia)

    2016-01-22

    Since its first criticality in 1982, PUSPATI TRIGA Reactor (RTP) has been operated for more than 30 years. As RTP become older, ageing problems have been seen to be the prominent issues. In addressing the ageing issues, an Ageing Management (AgeM) database for managing related ageing matters was systematically developed. This paper presents the development of AgeM database taking into account all RTP major Systems, Structures and Components (SSCs) and ageing mechanism of these SSCs through the system surveillance program.

  12. Burn-up TRIGA Mark II benchmark experiment

    International Nuclear Information System (INIS)

    Persic, A.; Ravnik, M.; Zagar, T.

    1998-01-01

    Different reactor codes are used for calculations of reactor parameters. The accuracy of the programs is tested through comparison of the calculated values with the experimental results. Well-defined and accurately measured benchmarks are required. The experimental results of reactivity measurements, fuel element reactivity worth distribution and fuel-up measurements are presented in this paper. The experiments were performed with partly burnt reactor core. The experimental conditions were well defined, so that the results can be used as a burn-up benchmark test case for a TRIGA Mark II reactor calculations.(author)

  13. Operating and maintenance experience at the TRIGA Mainz reactor

    International Nuclear Information System (INIS)

    Menke, Helmut

    1982-01-01

    In January 1966 the TRIGA Mark II reactor of the University of Mainz went into operation, licensed for 100 kW steady state operation and 250 MW pulse performance but a severe disadvantage was observed in the Radiation Monitoring System, which lead to installation of a new monitoring system. An Inspection of the beam-tubes using an endoscope is performed to assess its condition after 17 years of work. The observed corrosion and other damages are reported. The results from the fuel element measurement are also reported

  14. Temperature distribution calculations in TRIGA fuel element after the pulse

    International Nuclear Information System (INIS)

    Mele, I.; Ravnik, M.

    1992-01-01

    The computer program TEMPUL for calculating radial temperature distribution in a fuel element after the pulse operation is shortly described. It is based on one-dimensional diffusion equation for heat transfer in cylindrical geometry and implicit boundary condition at the element-coolant interface, defined by empirical boiling curve, which relates the heat flux from the rod and the difference between the fuel element surface temperature and water boiling point. As an example the results of such analysis of maximal allowed pulse at TRIGA Mark II reactor in Ljubljana are presented. (author) [sl

  15. Corrosion problem in the CRENK Triga Mark II research reactor

    International Nuclear Information System (INIS)

    Kalenga, M.

    1990-01-01

    In August 1987, a routine underwater optical inspection of the aluminum tank housing the core of the CRENK Triga Mark II reactor, carried out to update safety condition of the reactor, revealed pitting corrosion attacks on the 8 mm thick aluminum tank bottom. The paper discuss the work carried out by the reactor staff to dismantle the reactor in order to allow a more precise investigation of the corrosion problem, to repair the aluminum tank bottom, and to enhance the reactor overall safety condition

  16. Installation and operation of a high-temperature surface ion source for the online coupling of TRIGA-SPEC to the TRIGA Mainz research reactor and high-precision mass measurements of transuranium nuclides at TRIGA-TRAP

    Energy Technology Data Exchange (ETDEWEB)

    Renisch, Dennis

    2016-06-23

    The TRIGA-SPEC setup is dedicated for high-precision measurements of ground-state properties of exotic nuclides, like fission products or transuranium nuclides. For the online coupling of TRIGA-SPEC to the TRIGA Mainz research reactor, fission products are transported from a target chamber close to the reactor core by a gas-jet system to an ion source, which provides the connected experiments with a radioactive ion beam for the actual measurements. The design, installation and operation of the online ion source was a major part of the work described in this thesis. In addition, investigations on the optimal conditions of the gas-jet system for a reliable ion source operation were performed and the coupling of the ion source part to the subsequent elements of the beamline was conducted. The second part of this thesis deals with high-precision mass measurements on transuranium nuclides with the Penning-trap mass spectrometer TRIGA-TRAP, which is one branch of the TRIGA-SPEC setup. These measurements contribute to a mapping of the region of the chart of nuclides around the deformed shell closure at N=152. The masses of some of the investigated nuclides are directly measured for the first time. Due to the appearance of systematic inconsistencies during the evaluation of the mass measurements, the focus of the final section of this thesis lies in the identification and correction of the sources of the observed inconsistencies.

  17. Plate type heat exchanger for Reaktor TRIGA PUSPATI

    International Nuclear Information System (INIS)

    Mohd Fairus Abdul Farid; Mohammad Suhaimi Kassim; Mohamad Puad Abu; Mohd Fazli Zakaria; Shaharum Ramli

    2008-08-01

    The PUSPATI TRIGA reactor (RTP) reached its first criticality status on 28 June 1982 with an installed capacity of 1 MW. After 26 years in operation, the ageing process has set in and many systems in the reactor need maintenance and replacement. Among these systems, the more critical one is the heat exchanger system. Currently, the shell and tube type heat exchanger is being used. It has been observed that the performance of the heat exchanger has dropped significantly over the years. Visual inspections indicate that the tubes are corroded or even to the extent of being totally blocked. With this in mind and also with the setting up of the new Nuclear Power Division, whose mission includes upgrading the present reactor to 3 MW, the heat exchanger system is essential and a critical component. Literature indicates that the use of plate type exchangers are more efficient than the current shell and tube type. This paper will look into the engineering and safety aspects of using the plate type heat exchanger to the current TRIGA PUSPATI reactor. (Author)

  18. Probabilistic safety analysis for the Triga reactor Vienna

    International Nuclear Information System (INIS)

    Boeck, H.; Kirchsteiger, C.

    1988-07-01

    Triga-type reactors are the most widely used low power research reactors with power levels up to 3 MW. Although Triga reactors are considered inherently safe, due to their unique features such as prompt negative temperature coefficient and low power density, the reactor core still contains a respectable amount of activity which could lead under very adverse circumstances to radiation exposure both of staff members and of public. Such circumstances could be external events, accidents during fuel element manipulation or a loss of coolant water with exposure of the core. Therefore, it was decided to look more closely to various accident pathways and to calculate their probability, if possible. A major drawback is the lack of statistical material because no centralized registration of failures is carried out. Therefore, in many cases values from other research reactor types or even from power reactor statistics had to be used, thus increasing the uncertainty of the results. As most undesired event or TOP-event in this analysis a radiation exposure of staff members, the public or both together was selected and the probabilities of different pathways leading to this exposure was calculated. In the present case 'radiation exposure' are dose rates or activity concentration above the international accepted limits for occupational staff or public. 20 refs., 10 figs. (Author)

  19. Derived release limits for airborne effluents at TRIGA - INR Reactor

    International Nuclear Information System (INIS)

    Toma, A.; Dulama, C.; Hirica, O.; Mihai, S.; Oprea, I.

    2008-01-01

    Beginning from fulfilling the purposes of dose limitation system recommended by ICRP, and now accepted in radiation protection, this paper presents an environmental transfer model to calculate derived release limits for airborne and gaseous radioactive effluents at TRIGA-INR, 14 MW Steady State Reactor, in function on INR-Pitesti site. The methodology consists in determination of the principal exposure pathways for different groups of population and dose calculations for each radionuclide. The characterization of radionuclides transfer to environment was made using the compartmental model. The parameter transfer concept was used to describe the distribution of radionuclides between the different compartments. Atmospheric dispersion was very carefully treated, because it is the primary mechanism of the transfer of radionuclides in the environment and it determines all exposure pathways. Calculation of the atmospheric dispersion was made using ORION-II computer code based on the Gaussian plume model which takes account of site's specific climate and relief conditions. Default values recommended by literature were used to calculate some of the parameters when specific site values were not available. After identification of all transfer parameters which characterize the most important exposure pathways, the release rate corresponding to the individual dose rate limit was calculated. This maximum release rate is the derived release limit for each radionuclide and source. In the paper, the derived release limits are calculated for noble gases, radioiodine and other airborne particulate radionuclides, which can be released on the TRIGA-INR reactor stack, and are important to radiation protection. (authors)

  20. Qualitative assessment of the value of the Ohio State University TRIGA reactor

    International Nuclear Information System (INIS)

    Binney, S.E.; Johnson, A.G.

    1989-01-01

    The Oregon State University (OSU) TRIGA Reactor (OSTR) is a major regional research, training, and service facility. The OSTR supports a wide variety of organizations at the local, state, regional, national, and international levels. Examples of usage of the OSTR are given in this paper to serve as a basis for assessing the value of the OSTR to its user organizations. It is difficult to assess the value of a facility such as the OSTR quantitatively, primarily because a dollar value cannot be assigned to many of the services that the OSTR performs, e.g., forensic analysis to assist police agencies in criminal cases. Significant qualitative statements can be made, however, to demonstrate the fact that the value of a research reactor facility such as the OSTR substantially outweighs the capital and operating costs of such a facility. Analysis of the data presented above clearly indicates that the value of the OSTR facility is overwhelmingly positive, i.e., the benefits associated with the services provided by the OSTR facility outweigh the cost of providing such services by perhaps as much as an order of magnitude

  1. Visualization of neutron flux and power distributions in TRIGA Mark II reactor as an educational tool

    International Nuclear Information System (INIS)

    Snoj, Luka; Ravnik, Matjaz; Lengar, Igor

    2008-01-01

    Modern Monte Carlo computer codes (e.g. MCNP) for neutron transport allow calculation of detailed neutron flux and power distribution in complex geometries with resolution of ∼1 mm. Moreover they enable the calculation of individual particle tracks, scattering and absorption events. With the use of advanced software for 3D visualization (e.g. Amira, Voxler, etc.) one can create and present neutron flux and power distribution in a 'user friendly' way convenient for educational purposes. One can view axial, radial or any other spatial distribution of the neutron flux and power distribution in a nuclear reactor from various perspectives and in various modalities of presentation. By visualizing the distribution of scattering and absorption events and individual particle tracks one can visualize neutron transport parameters (mean free path, diffusion length, macroscopic cross section, up-scattering, thermalization, etc.) from elementary point of view. Most of the people remember better, if they visualize the processes. Therefore the representation of the reactor and neutron transport parameters is a convenient modern educational tool for the (nuclear power plant) operators, nuclear engineers, students and specialists involved in reactor operation and design. The visualization of neutron flux and power distributions in Jozef Stefan Institute TRIGA Mark II research reactor is treated in the paper. The distributions are calculated with MCNP computer code and presented using Amira and Voxler software. The results in the form of figures are presented in the paper together with comments qualitatively explaining the figures. (authors)

  2. Comparison of standardised decommissioning costing tools on pilot Vienna TRIGA MARK-II research reactor

    International Nuclear Information System (INIS)

    Hornacek, M.; Kristofova, K.; Slugen, V.; Zachar, M.; Stummer, T.

    2017-01-01

    The main purpose of the paper is to compare decommissioning costing code CERREX (Cost Estimation for Research Reactors in Excel) with advanced calculation methodology applied in eOMEGA-RR code. CERREX code was developed in line with the IAEA recommendations for decommissioning costing of research facilities and fully implements the ISDC (International Structure for Decommissioning Costing of Nuclear Installations) structure and costing methodology. In comparison with CERREX, usually applied in preliminary costing, the code eOMEGA-RR incorporates the realistic activity and material flow during decommissioning process (e.g. decontamination, dismantling and waste management). This advanced approach enables to carry out the decommissioning planning and costing more effectively. Moreover, the user-friendly interface helps to perform wide range of sensitivity analyses. In order to meet the above mentioned objectives, the model calculation costing case for TRIGA MARK-II research reactor in Vienna was developed in both calculation codes. The whole process covered four step-by-step procedures to be implemented. At first, inventory database taking into account physical as well as radiological parameters (e.g.: contamination, dose rates, nuclide vectors, limits and conditions) was developed. At second, advanced decommissioning costing case using CERREX and eOMEGA-RR code was created. At third, sensitivity analyses to estimate the impact of changing input parameters on calculated results were performed. Finally, costing results obtained from both cost calculation codes are compared and discussed. (authors)

  3. A combined wet/dry sipping cell for investigating failed triga fuel elements

    International Nuclear Information System (INIS)

    Boeck, H.; Gallhammer, H.; Hammer, J.; Israr, M.

    1987-08-01

    A sipping cell to detect failed triga fuel has been designed and constructed at the Atominstitut. The cell allows both wet- and dry sipping of one single standard triga fuel element. In the dry sipping method the fuel element may be electrically heated up to a maximum temperature of about 300 0 C to allow the detection of temperature dependent fission product release from the fuel element. 20 figs., 1 tab. (Author)

  4. Environmental Assessment: Relocation and storage of TRIGA reg-sign reactor fuel, Hanford Site, Richland, Washington

    International Nuclear Information System (INIS)

    1995-08-01

    In order to allow the shutdown of the Hanford 308 Building in the 300 Area, it is proposed to relocate fuel assemblies (101 irradiated, three unirradiated) from the Mark I TRIGA Reactor storage pool. The irradiated fuel assemblies would be stored in casks in the Interim Storage Area in the Hanford 400 Area; the three unirradiated ones would be transferred to another TRIGA reactor. The relocation is not expected to change the offsite exposure from all Hanford Site 300 and 400 Area operations

  5. Proposal of LDR Ir-192 Production in the TRIGA Mark II Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Karimzadeh, S.; Khan, R.; Boeck, H., E-mail: Sam.karimzadeh@ati.ac.a, E-mail: Nrustam@ati.ac.a, E-mail: Boeck@ati.ac.a [Institute of Atomic and Subatomic Physics (ATI), Vienna University of Technology (TU-Vienna) Stadionallee 2, 1020-Vienna (Austria)

    2011-07-01

    The TRIGA MARK II research reactor in Vienna provides some irradiation positions with different flux distribution. In this regard, a case study is under investigation to appraise the possibility of medical radioisotope production in Vienna. For this purpose, neutron flux mapping and the axial neutron flux distribution are calculated by MCNP5 for the TRIGA Mark II core. This paper describes the feasibility of Low Dose Rate (LDR) {sup 192}Ir production in the core of the low power research reactor. (author)

  6. Assessment results of the Indonesian TRIGA SNF to be shipped to INEEL

    International Nuclear Information System (INIS)

    Jefimoff, J.; Robb, A.K.; Wendt, K.M.; Syarip, I.; Alfa, T.

    1997-01-01

    This paper describes the Training, Research, Isotope, General Atomics (TRIGA) spent nuclear fuel (SNF) examination performed by technical personnel from the Idaho National Engineering and Environmental Laboratory (INEEL) at the Bandung and Yogyakarta research reactor facilities in Indonesia. The examination was required before the SNF would be accepted for transportation to and storage at the INEEL. This paper delineates the Initial Preparations prior to the Indonesian foreign research reactor (FRR) fuel examination. The technical basis for the examination, the TRIGA SNF Acceptance Criteria, and the physical condition required for transportation, receipt and storage of the TRIGA SNF at the INEEL is explained. In addition to the initial preparations, preparation descriptions of the Work Plan For TRIGA Fuel Examination, the Underwater Examination Equipment used, and personnel Examination Team Training are included. Finally, the Fuel Examination and Results of the aluminum and stainless steel clad TRIGA fuel examination have been summarized. Lessons learned from all the activities completed to date is provided in an addendum. The initial preparations included: (1) coordination between the INEEL, FRR or Badan Tenaga Atom Nasional (BATAN), DOE-HQ, and the US State Department and Embassy; (2) incorporating Savannah River Site (SRS) FRR experience and lessons learned; (3) collecting both FRR facility and spent fuel data, and issuing a radionuclide report (Radionuclide Mass Inventory, Activity, Decay Heat, and Dose Rate Parametric Data for TRIGA Spent Nuclear Fuels) needed for transportation and fuel acceptance at the INEEL; and (4) preexamination work at the research reactor for the fuel examination

  7. Activation of TRIGA Mark II research reactor concrete shield

    International Nuclear Information System (INIS)

    Zagar, Tomaz; Ravnik, Matjaz; Bozic, Matjaz

    2002-01-01

    To determine neutron activation inside the TRIGA research reactor concrete body a special sample-holder for irradiation inside horizontal channel was developed and tested. In the sample-holder various samples can be irradiated at different concrete shielding depths. In this paper the description of the sample-holder, experiment conditions and results of long-lived activation measurements are given. Long-lived neutron-induced gamma-ray-emitting radioactive nuclides in the samples were measured with HPGe detector. The most active long-lived radioactive nuclides in ordinary concrete samples were found to be 60 Co and 152 Eu and in barytes concrete samples 60 Co, 152 Eu and 133 Ba. Measured activity density of all nuclides was found to decrease almost linearly with depth in logarithmic scale. (author)

  8. Decommissioning of the ICI TRIGA Mark I reactor

    International Nuclear Information System (INIS)

    Parry, D.R.; England, M.R.; Ward, A.; Green, D.

    2000-01-01

    This paper considers the fuel removal, transportation and subsequent decommissioning of the ICI TRIGA Mark I Reactor at Billingham, UK. BNFL Waste Management and Decommissioning carried out this work on behalf of ICI. The decommissioning methodology was considered in the four stages to be described, namely Preparatory Works, Reactor Defueling, Intermediate Level Waste Removal and Low Level Waste Removal. This paper describes the principal methodologies involved in the defueling of the reactor and subsequent decommissioning operations, highlighting in particular the design and safety case methodologies used in order to achieve a solution which was completed without incident or accident and resulted in a cumulative radiation dose to personnel of only 1.57 mSv. (author)

  9. Assessment of TRIGA pulsing reactor safety without loss of coolant

    International Nuclear Information System (INIS)

    Baers, B.; Kaall, L.

    1979-11-01

    The fuel temperature, fuel can pressure and cladding stresses in the TRIGA fuel elements at various transient states has in part been measured and semiempirically estimated in cases where no experiments could be carried out. To the end of obtaining more reliable and up-to-date data and reactor parameters, reactor power and fuel temperatures were measured both at steady and transient states, including pulsing up to 280 MW. A novel and more realistic model for the fuel temperature at pulsing was also presented and used in the assessments. Based on the present assessment with the water coolant present and a maximum excess reactivity of (Δksub(max)/β = ) 4 $, no reactivity induced transients can bring about any undue hazards to the reactor or to the surroundings. (author)

  10. FIR 1 TRIGA activity inventories for decommissioning planning

    International Nuclear Information System (INIS)

    Raety, Antti; Kotiluoto, Petri

    2016-01-01

    The objective of the study has been to estimate the residual activity in the decommissioning waste of TRIGA Mark II type research reactor FiR 1 in Finland. Neutron flux distributions were calculated with Monte Carlo code MCNP. These were used in ORIGEN-S point-depletion code to calculate the neutron induced activity of materials at different time points by modelling the irradiation history and radioactive decay. The knowledge of radioactive inventory of irradiated materials is important in the planning of the decommissioning activities and is essential for predicting the radiological impact to personnel and environment. Decommissioning waste consists mainly of ordinary concrete, aluminium, steel and graphite parts. Results include uncertainties due to assumptions on material compositions and possible diffusion of gaseous nuclides. Comparison to activity inventory estimates of two other decommissioned research reactors is also presented. (authors)

  11. Thermal hydraulic analysis of the IPR-R1 TRIGA reactor; Analise termo-hidraulica do reator TRIGA IPR-R1

    Energy Technology Data Exchange (ETDEWEB)

    Veloso, Marcelo Antonio [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Fortini, Maria Auxiliadora [Minas Gerais Univ., Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2002-07-01

    The subchannel approach, normally employed for the analysis of power reactor cores that work under forced convection, have been used for the thermal hydraulic evaluation of a TRIGA Mark I reactor, named IPR-R1, at 250 kW power level. This was accomplished by using the PANTERA-1P subchannel code, which has been conveniently adapted to the characteristics of natural convection of TRIGA reactors. The analysis of results indicates that the steady state operation of IPR-R1 at 250 kW do not imply risks to installations, workers and public. (author)

  12. TRIGA fuel burn-up calculations and its confirmation

    Energy Technology Data Exchange (ETDEWEB)

    Khan, R., E-mail: rustamzia@yahoo.co [Vienna University of Technology (TU Wien)/Atominstitute (ATI), Stadionallee 2, A-1020, Vienna (Austria); Karimzadeh, S.; Boeck, H. [Vienna University of Technology (TU Wien)/Atominstitute (ATI), Stadionallee 2, A-1020, Vienna (Austria)

    2010-05-15

    The Cesium (Cs-137) isotopic concentration due to irradiation of TRIGA Fuel Elements FE(s) is calculated and measured at the Atominstitute (ATI) of Vienna University of Technology (VUT). The Cs-137 isotope, as proved burn-up indicator, was applied to determine the burn-up of the TRIGA Mark II research reactor FE. This article presents the calculations and measurements of the Cs-137 isotope and its relevant burn-up of six selected Spent Fuel Elements SPE(s). High-resolution gamma-ray spectroscopy based non-destructive method is employed to measure spent fuel parameters. By the employment of this method, the axial distribution of Cesium-137 for six SPE(s) is measured, resulting in the axial burn-up profiles. Knowing the exact irradiation history and material isotopic inventory of an irradiated FE, six SPE(s) are selected for on-site gamma scanning using a special shielded scanning device developed at the ATI. This unique fuel inspection unit allows to scan each millimeter of the FE. For this purpose, each selected FE was transferred to the fuel inspection unit using the standard fuel transfer cask. Each FE was scanned at a scale of 1 cm of its active length and the Cs-137 activity was determined as proved burn-up indicator. The measuring system consists of a high-purity germanium detector (HPGe) together with suitable fast electronics and on-line PC data acquisition module. The absolute activity of each centimeter of the FE was measured and compared with reactor physics calculations. The ORIGEN2, a one-group depletion and radioactive decay computer code, was applied to calculate the activity of the Cs-137 and the burn-up of selected SPE. The deviation between calculations and measurements was in range from 0.82% to 12.64%.

  13. ANALISIS KONVEKSI ALAM TERAS REAKTOR TRIGA BERBAHAN BAKAR TIPE PELAT MENGGUNAKAN COOLOD-N2

    Directory of Open Access Journals (Sweden)

    Sudjatmi K A

    2015-06-01

    Full Text Available ABSTRAK Analisis Konveksi Alam Teras Reaktor Triga Berbahan Bakar Tipe Pelat MENGGUNAKAN COOLOD-N2. Rencana penghentian produksi elemen bakar jenis TRIGA oleh produsen elemen bakar reaktor TRIGA, sudah seharusnya diantisipasi oleh badan pengoperasi reaktor TRIGA untuk menggantikan elemen bakar tipe silinder tersebut dengan tipe pelat yang tersedia di pasaran. Pada penelitian ini dilakukan perhitungan untuk model teras reaktor dengan spesifikasi utama menggunakan bahan bakar U3Si2Al dengan pengayaan uranium  sebesar 19,75% dan tingkat muat 2,96 gU/cm3. Analisis dilakukan menggunakan program COOLOD-N2 yang tervalidasi pada konfigurasi teras TRIGA konversi berbahan bakar tipe pelat, yang tersusun atas 16 elemen bakar, 4 elemen kendali dan 1 fasilitas iradiasi yang terletak tepat di tengah teras. Hasil analisis menunjukkan bahwa dengan temperatur pendingin masuk ke teras sebesar 37oC, dan rasio faktor puncak daya radial ≤ 1,92 maka daya maksimum yang dapat dioperasikan pada moda operasi konveksi bebas adalah 600 kW. Karakteristik termohidrolika yang diperoleh antara lain adalah temperatur pendingin di sisi outlet, kelongsong dan meat masing-masing sebesar 82,39oC, 108,88oC, dan 109,02oC, pada ΔTONB (Temperature Onset of Nucleate Boiling =7,18oC dan nilai OFIR (Onset of flow instability ratio =1,03 Hasil yang diperoleh dari perhitungan ini diharapkan dapat dijadikan acuan untuk menentukan tingkat daya reaktor TRIGA berbahan bakar pelat. Kata kunci: TRIGA Konversi, COOLOD-N2, karakteristik termohidrolika, konveksi alam, elemen bakar tipe pelat.   ABSTRACT ANALYSIS OF NATURAL CONVECTION IN TRIGA REACTOR CORE PLATE TYPES FUELED USING COOLOD-N2. Any pretensions to stop the production of TRIGA fuel elements by TRIGA reactor fuel elements manufacturer should be anticipated by the operating agency of TRIGA reactor to replace the cylinder type fuel element with plate type fuel element that available on the market. In this study, the calculation of U3Si2

  14. 76 FR 69296 - University of Utah, University of Utah TRIGA Nuclear Reactor, Notice of Issuance of Renewed...

    Science.gov (United States)

    2011-11-08

    ... From the Federal Register Online via the Government Publishing Office NUCLEAR REGULATORY COMMISSION University of Utah, University of Utah TRIGA Nuclear Reactor, Notice of Issuance of Renewed... University of Utah (UU, the licensee), which authorizes continued operation of the UU TRIGA Nuclear Reactor...

  15. Shipment of TRIGA spent fuel to DOE's INEEL site - a status report

    International Nuclear Information System (INIS)

    Patterson, John; Viebrock, James; Shelton, Tom; Parker, Dixon

    1998-01-01

    DOE placed its transportation services contract with NAC International in April 1997 and awarded the first task to NAC for return of TRIGA fuel in July 1997. This initial shipment of TRIGA fuel, scheduled for early 1998, is reflective of many of the difficulties faced by DOE and the transportation services contractor in return of the foreign research reactor fuel to the United States: 1) First time use of the INEEL dry storage facility for receipt of research reactor fuel; 2) Safety analysis of the INEEL facility for the NAC-LWT shipping cask; 3) Cask certification for a mixed loading of high enriched and low enriched TRIGA fuels; 4) Cask loading for standard length and extended length rods (instrumented and fuel follower control rods); 5) Design and certification of a canister for degraded TRIGA fuel; 6) Initial port entry through the Naval Weapons Station in Concord, California; 7) Initial approval of the rail route for shipment from Concord to INEEL. In this presentation we describe the overall activities involved in the first TRIGA shipment, discuss the actions required to resolve the difficulties identified above, and provide a status report of the initial shipment from South Korea and Indonesia. Recommendations are presented as to actions that can be taken by the research reactor operator, by DOE, and by the transportation services agent to speed and simplify the transportation process. Actions having the potential to reduce costs to DOE and to reactor operators from high-income economies will be identified. (author)

  16. Control and safety systems for TRIGA irradiation facilities C5 and C9

    International Nuclear Information System (INIS)

    Talpalariu Cornel Talpalariu Jeni Crucean Mircea Matei Corina

    2008-01-01

    Full text: The Institute for Nuclear Research conducted research for designing and manufacturing of microprocessor equipment for some irradiation facilities operating by the TRIGA reactor. This equipment has accumulated a wide operating time allowing the conclusions referring to reliability, ergonomics, and design of the operating facilities. Based upon these studies a new program was initiated for the design and manufacturing of a modern equipment with improved reliability and flexibility performances. The system provides the user with a multitude of options, numerical and analog interfaces, keyboard and high reliability local display. The main functional components of the system are: - 8 PID full options regulating loops; - 8 safety analog channels having 4 preset trips; - watch dog restart and fault tolerant facilities; - 8 high precision analog with an input of 0 - 15 mV from thermocouple; - 8 computer controlled power supplies of 220 V, 1 kWA; - alphanumeric display and keyboard; - fault tolerant analog scanner. A real improvement of the system is the future remote control computer, a PC AT Pentium working like a system controller, real time data acquisition, and operator's adviser. This new facility allows the operator to set the trips or to control remotely all the power supply and step-by-step positioner of irradiation device. Software design for acquisition and data processing provides modern techniques for operator interfacing, representation recording and protection of test results. Software implementation keeps a special organization supported by a real time executive that is the best method to achieve the performance required. Following this objective, the software structure consists of: 1. Tasks as follows: - testing parameters setup; - data processing routines; - engineering and electrical conversion; - numerical / graphical data representation; - test results recording routines and data base management. 2. Drivers as follows: - A/I and D

  17. Generic Procedures for Response to a Nuclear or Radiological Emergency at Triga Research Reactors. Attachment 1 (2011)

    International Nuclear Information System (INIS)

    2011-01-01

    The publication provides guidance for response to emergencies at TRIGA research reactors in Threat Category II and III. It contains information on the unique behaviour of TRIGA fuel during accident conditions; it describes design characteristics of TRIGA research reactors and provides specific symptom-based emergency classification for this type of research reactor. This publication covers the determination of the appropriate emergency class and protective actions for a nuclear or radiological emergency at TRIGA research reactors. It does not cover nuclear security at TRIGA research reactors. The term 'threat category' is used in this publication as described in Ref. [6] and for the purposes of emergency preparedness and response only; this usage does not imply that any threat, in the sense of an intention and capability to cause harm, has been made in relation to facilities, activities or sources. The threat category is determined by an analysis of potential nuclear and radiological emergencies and the associated radiation hazard that could arise as a consequence of those emergencies. STRUCTURE. The attachment consists of an introduction which defines the background, objective, scope and structure, two sections covering technical aspects and appendices. Section 2 describes the characteristics of TRIGA fuel in normal and accident conditions. Section 3 contains TRIGA research reactor specific emergency classification tables for Threat Category II and III. These tables should be used instead of the corresponding emergency classification tables presented in Ref. [1] while developing the emergency response arrangements at TRIGA research reactors. The appendices present some historical overview and typical general data for TRIGA research reactor projects and the list of TRIGA installations around the world. The terms used in this document are defined in the IAEA Safety Glossary and the IAEA Code of Conduct on the Safety of Research Reactors.

  18. Computer programs for TRIGA calibration, burnup evaluation, and bookkeeping

    International Nuclear Information System (INIS)

    Nelson, George W.

    1978-01-01

    Several computer programs have been developed at the University of Arizona to assist the direction and operation of the TRIGA Reactor Laboratory. The programs fall into the following three categories: 1. Programs for calculation of burnup of each fuel element in the reactor core, for maintaining an inventory of fuel element location and fissile content at any time, and for evaluation of the reactivity effects of burnup or proposed fuel element rearrangement in the core. 2. Programs for evaluation, function fitting, and tabulation of control rod measurements. 3. Bookkeeping programs to summarize and tabulate reactor runs and irradiations according to time, energy release, purpose, responsible party, etc. These summarized data are reported in an annual operating report for the facility. The use of these programs has saved innumerable hours of repetitious work, assuring more accurate, objective results, and requiring a minimum of effort to repeat calculations when input data are modified. The programs are written in FORTRAN-IV, and have been used on a CDC-6400 computer. (author)

  19. Planned Scientific programs around the Triga Mark 2 Reactor

    International Nuclear Information System (INIS)

    Majah, M Ibn.

    2007-01-01

    Full text: Nuclear techniques have been introduced to Morocco since the sixties. After the energy crisis of 1973, Morocco decides to create the National Center for Energy Sciences and Nuclear Techniques (CNESTEN) under the supervision of the Ministry of high Education and Research, with a research commercial and support vocation. CNESTEN is in charge of promoting nuclear application, to act as technical support for the authorities and to prepare the technological basis for nuclear power option. In 1998, CNESTEN started the construction of Nuclear Research Centre. The on going activities cover many sectors : earth and environmental sciences, high energy physics, safety and security, waste management. In 2001, CNESTEN started the construction of a 2MW TRiga Mark 2 Reactor, with the possibility to increase the power to 3 MW. The construction was achieved in January 2007. The operation of the reactor is expected for April 2007. The program of the utilization of the reactor was established with th contribution of the university and with the assistance of IAEA. Some of the experimental set-up installed around the reactor have been designed. CNESTEN has developed cooperation with Nuclear research centres from other countries and is receiving visitors and trainees mainly through the IAEA [fr

  20. Characterization of the Ljubljana TRIGA thermal column neutron radiographic facility

    International Nuclear Information System (INIS)

    Nemec, T.; Rant, J.; Kristof, E.; Glumac, B.

    1995-01-01

    An extensive characterization of the neutron beam of the existing neutron radiographic facility in the thermal column of the Ljubljana Triga Mark II research reactor is in progress. Neutron beam characteristics are needed to determine the effect of various neutron and gamma radiation on the neutron radiographic image. Commercially available medical scintillator converter screens based on Gd dioxy sulphite as well as Gd metal neutron converters are used to record neutron radiographic image. Thermal, epithermal and fast neutron fluxes were measured using Au and In activation detectors and cadmium ratio is determined. Neutron beam flux profiles are measured by film densitometry and by Au activation detector wires. By exposing films shielded by boral or lead plates individual contributions of thermal, epithermal neutrons and gamma radiation are estimated by densitometric measurements. By recording images of neutron image quality indicators BPI (Beam Purity Indicator) and SI (Sensitivity Indicator) produced by Riso, standard neutron radiography image characteristic are established. In gamma dosimetric measurements thermoluminescent detectors (CaF 2 Mn) are used. (author)

  1. New burnup calculation of TRIGA IPR-R1 reactor

    International Nuclear Information System (INIS)

    Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z.

    2015-01-01

    The IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, operates since 1960.The reactor is operating for more than fifty years and has a long history of operation. Determining the current composition of the fuel is very important to calculate various parameters. The reactor burnup calculation has been performed before, however, new techniques, methods, software and increase of the processing capacity of the new computers motivates new investigations to be performed. This work presents the evolution of effective multiplication constant and the results of burnup. This new model has a more detailed geometry with the introduction of the new devices, like the control rods and the samarium discs. This increase of materials in the simulation in burnup calculation was very important for results. For these series of simulations a more recently cross section library, ENDF/B-VII, was used. To perform the calculations two Monte Carlo particle transport code were used: Serpent and MCNPX. The results obtained from two codes are presented and compared with previous studies in the literature. (author)

  2. Assessment of Power Quality Problems for TRIGA PUSPATI Reactor (RTP)

    International Nuclear Information System (INIS)

    Mohd Fazli Zakaria; Ramachandaramurthy, V.K.

    2016-01-01

    The electrical power systems are exposed to different types of power quality disturbances. Investigation and monitoring of power quality is necessary to maintain accurate operation of sensitive equipment especially for nuclear installations. This paper will discuss the power quality problems observed at the electrical sources of PUSPATI TRIGA Reactor (RTP). Assessment of power quality requires the identification of any anomalous behavior on a power system, which adversely affects the normal operation of electrical or electronic equipment. A power quality assessment involves gathering data resources; analyzing the data (with reference to power quality standards) then, if problems exist, recommendation of mitigation techniques must be considered. Field power quality data is collected by power quality recorder and analyzed with reference to power quality standards. Normally the electrical power is supplied to the RTP via two sources in order to keep a good reliability where each of them is designed to carry the full load. The assessment of power quality during reactor operation was performed for both electrical sources. There were several disturbances such as voltage harmonics and flicker that exceeded the thresholds. (author)

  3. Medical and radiobiological applications at the research reactor TRIGA Mainz

    International Nuclear Information System (INIS)

    Hampel, G.; Grunewald, C.; Kratz, J.-V.; Schmitz, T.; Schutz, C.; Werner, S.; Appelman, K.; Moss, R.; Blaickner, M.; Nawroth, T.; Otto, G.; Schmidberger, H.

    2010-01-01

    At the University of Mainz, Germany, a boron neutron capture therapy (BNCT) project has been started with the aim to expand and advance the research on the basis of the TAOrMINA protocol for the BNCT treatment of liver metastases of colorectal cancer. Irradiations take place at the TRIGA Mark II reactor. Biological and clinical research and surgery take place at the University and its hospital of Mainz. Both are situated in close vicinity to each other, which is an ideal situation for BNCT treatment, as similarly performed in Pavia, in 2001 and 2003. The application of BNCT to auto-transplanted organs requires development in the methodology, as well as regard to the irradiation facility and is part of the complex, interdisciplinary treatment process. The additional high surgical risk of auto-transplantation is only justified when a therapeutic benefit can be achieved. A BNCT protocol including explantation and conservation of the organ, neutron irradiation and re-implantation is logistically a very challenging task. Within the last years, research on all scientific, clinical and logistical aspects for the therapy has been performed. This includes work on computational modelling for the irradiation facility, tissue and blood analysis, radiation biology, dosimetry and surgery. Most recently, a clinical study on boron uptake in both healthy and tumour tissue of the liver and issues regarding dosimetry has been started, as well as a series of cell-biology experiments to obtain concrete results on the relative biological effectiveness (RBE) of ionizing radiation in liver tissue. (author)

  4. The new neutron imaging facility at TRIGA reactor in Morocco

    Energy Technology Data Exchange (ETDEWEB)

    Ouardi, A.; Alami, R.; Bensitel, A. [Centre National de l' Energie des Science et des Techniques Nucleaires, PB.1382 R.P 10001 Rabat (Morocco)

    2011-07-01

    A new neutron imaging facility is currently developed around 2 MW TRIGA MARK-II reactor at Maamora Nuclear research centre (CENM). Neutron imaging combined to X-ray or gamma radiography offers the opportunity to extend Non Destructive Testing (NDT) activities DT in Morocco to new fields of applications such as space and aircraft Moroccan industry, mining, wood industry and Archeology. The facility is planed to be completed in the end of 2011. In order to reduce the gamma-ray content in the neutron beam, the reactor tangential channel is selected. For power of 2 MW, the corresponding thermal neutron flux at the inlet of the tangential channel is around 1.10{sup 13}ncm{sup 2}/s. The facility will be based on a conical neutron collimator with a flight tube of 8m and offers three circular diaphragms with diameters of 1cm, 2 cm and 4 cm corresponding to L/D-ratio varying between 200 and 600. The holes will be housed in the primary shutter. These diaphragms' sizes allow to perform neutron radiography with high resolution (L/D = 600) and high speed (L/D= 200). Monte Carlo calculations by a fully 3D numerical code GEANT4 are used to optimize the whole neutron beam line and to reach a shorten distance between the source and detector and reduce as possible the exposure time. (author)

  5. Thermal spectra of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Macias B, L.R.; Palacios G, J.

    1998-01-01

    The diffraction phenomenon is gave in observance of the well known Bragg law in crystalline materials and this can be performance by mean of X-rays, electrons and neutrons among others, which allows to do inside the field of each one of these techniques the obtaining of measurements focussed at each one of them. For the present work, it will be mentioned only the referring to X-ray and neutron techniques. The X-ray diffraction due to its properties just it does measurements which are known in general as superficial measurements of the sample material but for the properties of the neutrons, this diffraction it explores in volumetric form the sample material. Since the neutron diffraction process depends lots of its intensity, then it is important to know the neutron source spectra that in this case is supplied by the TRIGA Mark III reactor. Within of diffraction techniques a great number of them can be found, however some of the traditional will be mentioned such as the identification of crystalline samples, phases identification and the textures measurement. At present this last technique is founded on the dot of a minimum error and the technique of phases identification performs but not compete with that which is obtained by mean of X-rays due to this last one has a major resolution. (Author)

  6. Neutronics analysis of TRIGA Mark II research reactor

    Directory of Open Access Journals (Sweden)

    Haseebur Rehman

    2018-02-01

    Full Text Available This article presents clean core criticality calculations and control rod worth calculations for TRIGA (Training, Research, Isotope production-General Atomics Mark II research reactor benchmark cores using Winfrith Improved Multi-group Scheme-D/4 (WIMS-D/4 and Program for Reactor In-core Analysis using Diffusion Equation (PRIDE codes. Cores 133 and 134 were analyzed in 2-D (r, θ and 3-D (r, θ, z, using WIMS-D/4 and PRIDE codes. Moreover, the influence of cross-section data was also studied using various libraries based on Evaluated Nuclear Data File (ENDF/B-VI.8 and VII.0, Joint Evaluated Fission and Fusion File (JEFF-3.1, Japanese Evaluated Nuclear Data Library (JENDL-3.2, and Joint Evaluated File (JEF-2.2 nuclear data. The simulation results showed that the multiplication factor calculated for all these data libraries is within 1% of the experimental results. The reactivity worth of the control rods of core 134 was also calculated with different homogenization approaches. A comparison was made with experimental and reported Monte Carlo results, and it was found that, using proper homogenization of absorber regions and surrounding fuel regions, the results obtained with PRIDE code are significantly improved.

  7. PENGARUH NILAI BAKAR TERHADAP INTEGRITAS KELONGSONG ELEMEN BAKAR TRIGA 2000

    Directory of Open Access Journals (Sweden)

    K.A. Sudjatmi

    2015-04-01

    Full Text Available Bentuk elemen bakar reaktor TRIGA Bandung adalah silinder padat yang merupakan campuran homogen paduan uranium dan zirkonium hidrida. Pada saat reaktor beroperasi, suhu elemen bakar akan bertambah, akibatnya akan menaikan tekanan gas-gas yang ada di dalam kelongsong elemen bakar. Tekanan gas yang timbul dalam kelongsong elemen bakar merupakan penjumlahan tiga komponen tekanan yaitu tekanan akibat udara yang terperangkap antara kelongsong dengan bahan bakar, tekanan oleh gas hasil fisi yang terbentuk dari elemen bakar dan tekanan yang berasal dari pemisahan hidrogen dari paduan zirkonium hidrida. Gas hasil fisi yang terbentuk oleh bahan bakar sebanding dengan besarnya fraksi bakar oleh setiap elemen bakar dalam teras reaktor. Semakin besar fraksi bakar elemen bakar, semakin besar gas gas hasil fisi yang dihasilkannya, akibatnya semakin besar tekanan di dalam kelongsong yang disebabkan oleh gas gas hasil fisi tersebut. Perhitungan jumlah gas-gas hasil fisi dalam kelongsong yang merupakan fungsi dari nilai bakar dilakukan dengan menggunakan program ORIGEN-2. Program ORIGEN-2 adalah kode komputer yang banyak digunakan untuk menghitung hasil fisi, peluruhan dan pengolahan bahan radioaktif. Tampang lintang, presentase timbulnya hasil fisi, data peluruhan, dan data lainnya yang diperlukan disediakan dalam pustaka data selama eksekusi program. Dari hasil perhitungan dapat disimpulkan bahwa tekanan gas yang diakibatkan oleh gas hasil fisi adalah 4,13 10-3 psi dan tekanan gas yang diakibatkan udara yang terjebak di dalam kelongsong adalah 56,6 psi, yang mengakibatkan tegangan pada kelongsong sebesar 2080 psi dan nilai ini jauh lebih kecil dari setengah tegangan luluh bahan kelongsong sebesar 12.000 psi pada temperatur 750 oC atau sekitar 40.000 psi pada temperatur 138 oC. Akhirnya dapat disimpulkan bahwa dilihat dari sisi nilai bakar, maka elemen bakar layak digunakan sampai mencapai nilai bakar maksimum. Kata kunci : TRIGA, nilai bakar, elemen bakar

  8. Visual beam tube inspection at the TRIGA reactor Vienna

    International Nuclear Information System (INIS)

    Boeck, H.; Musilek, A.; Villa, M.

    2006-01-01

    Of the four TRIGA beam tubes two have been visually inspected in 1985. Prior to the inspection the reactor was shut down for 3 weeks. The fuel elements around the beam tubes were removed. Stainless steel dummy elements were inserted in the fuel positions to shield the core radiation. The active part of the Fast Rabbit Tube was removed into the beam tube loading device and transferred to an interim storage: Front dose rate was ∼ 50 mSv/h. Generally the beam tube was very clean, after the last inspection about 30 years ago. A1 cm cut was observed at the beam tube front end. A rigid endoscope was used to check the beam tube's inner surface using a 90 degree deflection objective and photo- and video equipment. The direct dose rate in front of the beam tube was about 30 mSv/h. The beam tube was vacuum cleaned. A corroded shielding tank containing boric acid has leaked. A wooden collimator partially disintegrating due to extreme temperature was removed from beam tube D. Documentation of the inspection for visible defects is produced for later comparison

  9. Probabilistic safety assessment for TRIGA SSR 14 MW

    International Nuclear Information System (INIS)

    Mladin, Daniela; Preda, Marin; Mladin, Mirea; Stefan, Ion

    2006-01-01

    Probabilistic Safety Assessment is a method for risk public and personnel evaluation in a complex installation such as a nuclear reactor, a chemical installation, etc. PSA techniques are powerful tools for improvement in design, construction, operation, safety and design efficiency, for modification and management of such complex installations. International practice has shown that safety analysis both for power reactors and research reactors involves definition of postulated initiating events. These initiators depend on reactor type, nominal power, fuel characteristics, safety systems, etc. Such scenarios or accident sequences derived from these initiating events are postulated in some cases, on the basis of engineering judgement. In the others cases, they are developed from probabilistic view point, starting with initiating event frequency and continuing with safety systems failures and human errors, which may intervene in the evolution of the event. Starting from these initiating events, event trees technique allows the calculation of the frequencies of final damage states having environmental or only inside installation consequences, depending on the success or failure of the safety systems analysed separately in the fault trees. The paper presents results of the PSA model for TRIGA SSR 14 Mw reactor, more precisely the qualitative and quantitative analysis of initiating evens and event trees

  10. Reactor calculations for improving utilization of TRIGA reactor

    International Nuclear Information System (INIS)

    Ravnik, M.

    1986-01-01

    A brief review of our work on reactor calculations of 250 kW TRIGA with mixed core (standard + FLIP fuel) will be presented. The following aspects will be treated: - development of computer programs; - optimization of in-core fuel management with respect to fuel costs and irradiation channels utilization. TRIGAP programme package will be presented as an example of computer programs. It is based on 2-group 1-D diffusion approximation and besides calculations offers possibilities for operational data logging and fuel inventory book-keeping as well. It is developed primarily for the research reactor operators as a tool for analysing reactor operation and fuel management. For this reason it is arranged for a small (PC) computer. Second part will be devoted to reactor physics properties of the mixed cores. Results of depletion calculations will be presented together with measured data to confirm some general guidelines for optimal mixed core fuel management. As the results are obtained using TRIGAP program package results can be also considered as an illustration and qualification for its application. (author)

  11. Improving the TRIGA facility maintenance by predictive maintenance techniques

    International Nuclear Information System (INIS)

    Preda, M.; Sabau, C.; Barbalata, E.

    1997-01-01

    This work deals with the specific operation of equipment in radioactive environment or in conditions allowing radioactive contamination. The requirements of remote operation ensuring the operators' protection are presented. Also, the requirements of international standards issued by IAEA-Vienna are reviewed. The organizational withdraws of the maintenance activities, based on the standards and maintenance and repair directives still in force, are shown. It is emphasized the fact that this type of maintenance was adequate to a given level of technical development, characteristic for pre-computerized industry, but, at present, it is obsolete and uneconomic both in utilization and maintenance. Such a system constitutes already a burden hindering the efforts of maximizing the availability, maintenance, prolongation the service life of equipment and utilities, finally, of increasing the efficiency of complex installations. Moreover, the predictive maintenance techniques are strongly requested by the character of radioactive installations precluding the direct access in given zones (a potential risk of irradiation or radioactive contamination) of installations during operation. The results obtained by applying the predictive maintenance techniques in the operation of the double circuit irradiation loop, used in the TRIGA reactors, are presented

  12. TRIGA R a widely used fuel - Assessments, validation and application of criticality standards

    International Nuclear Information System (INIS)

    Comte, N.; Thievenaz, B.; Doucet, M.; Torlini, E.; Pin, T.

    2009-01-01

    The TRIGA R (Training, Research, Isotope production, General Atomics), reactor is the most widely used non-power nuclear reactor in the world. Sixty six TRIGA R reactors are running or under construction in 24 countries. These reactors are used in many diverse applications, including production of radioisotopes for medicine and industry, treatment of tumors, non destructive testing, basic research on the properties of matter, and for education and training. TRIGA R International, a joint venture company between General Atomics and CERCA, a wholly-owned subsidiary of AREVA NP, manufactures and sells TRIGA R fuel to research reactor. The CERCA Factory is currently the only Company in the world to manufacture this type of fuel. Criticality safety evaluations implement requirements to justify sufficient sub critical margins for fuel fabrication plants. Basic criticality data (i. e., criticality standards) are used to determine the sub critical margins for all processes involving enriched uranium. Safety analysis help to identify and update those standards to bound normal, abnormal and accidental conditions. The French criticality code package CRISTAL V1 was used to perform the calculations. After a brief description of the fuel design and the calculation scheme, this paper presents the results for the different types and media of TRIGA R fuel (powder, elements). Experimental benchmarks data were used for validation as well as comparative calculations with Monte-Carlo codes MORET 4 and MCNP. The criticality section of the safety analysis report supporting the actual TRIGA workshops was reviewed to ensure the update of the criticality standards does not harm the safety of the installations. (authors)

  13. Analysis of gamma heating at TRIGA mark reactor core Bandung using plate type fuel

    International Nuclear Information System (INIS)

    Setiyanto; Tukiran Surbakti

    2016-01-01

    In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities) and central irradiation position (CIP), especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0.87 W/g), but very low value for Lazy Susan position (lest then 0.11 W/g). Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. (author)

  14. Modernization design of neutron radiography of ITU TRIGA Mark-II reactor

    International Nuclear Information System (INIS)

    Tugrul, B.; Bilge, A.N.

    1988-01-01

    ITU TRIGA MARK-II Research and Training Reactor has a power of 250 KW and has three beam tubes. One of them is tangential beam tube used for neutron radiography. In this study, the neutron radiography set in the tangential beam tube is described with its problems for ITU TRIGA Reactor. After that modernization of the system is designed and the applicability of the direct and indirect methods is evaluated. Improving the ratio of length to diameter for the beam tube, elimination the fogging on the film and constructive design for practice and secure application of the technique is developed. (author)

  15. Design and implementation of the control system for the new console of TRIGA-3-Salazar Reactor

    International Nuclear Information System (INIS)

    Gonzalez M, J.L.

    1994-01-01

    TRIGA-3-Salazar Reactor was set in operation in 1968 and the aging of its components has cause the increasing in the maintenance. In the presence of this, it becomes necessary to replace the reactor console using new technologies, considering the incorporation of a personal computer. The aim of this work is the design and construction of the equipment interfaces as well as the digital computer program for the automation and control of the TRIGA-3-Salazar Reactor by means of a personal computer. (Author)

  16. Operation and maintenance experience at the General Atomic Company's TRIGA reactor facility at San Diego, California

    International Nuclear Information System (INIS)

    Whittemore, W.L.; Stout, W.A.; Shoptaugh, J.R.; Chesworth, R.H.

    1982-01-01

    Since the startup of the original 250 kW TRIGA Mark I reactor in 1958, General Atomic Company has accumulated nearly 24 years of operation and maintenance experience with this type of reactor. In addition to the nearly 24 years of experience gained on the Mark I, GA has operated the 1.5 MW Advanced Prototype Test Reactor (Mark F) for 22 years and operated a 2 MW below-ground TRIGA Mark III for five years. Information obtained from normal and abnormal operation are presented. (author)

  17. Integrating 3D CAD data for manufacturing and fabrication the core model of reactor TRIGA PUSPATI

    International Nuclear Information System (INIS)

    Abu Bakar Harun

    2005-01-01

    This paper describe the intrigue integration of digital 3 Dimensional Computer Aided Design (3D CAD) data manipulation for the Core Model fabrication of REAKTOR TRIGA PUSPATI and ready for mass manufacturing. 3 Dimensional CAD data from Computer Aided Design program will be used as an interpreter in the fabrication of this project. The Core Model of REAKTOR TRIGA PUSPATI will be fabricated with the aid of 3D CAD drawings and digital files. The components will be segregated and divided into 2 categories namely Conventional d Rapid Fabrication. (Author)

  18. ANALYSIS OF GAMMA HEATING AT TRIGA MARK REACTOR CORE BANDUNG USING PLATE TYPE FUEL

    Directory of Open Access Journals (Sweden)

    Setiyanto Setiyanto

    2016-10-01

    Full Text Available ABSTRACT In accordance with the discontinuation of TRIGA fuel element production by its producer, the operation of all TRIGA type reactor of at all over the word will be disturbed, as well as TRIGA reactor in Bandung. In order to support the continuous operation of Bandung TRIGA reactor, a study on utilization of fuel plate mode, as used at RSG-GAS reactor, to replace the cylindrical model has been done. Various assessments have been done, including core design calculation and its safety aspects. Based on the neutronic calculation, utilization of fuel plate shows that Bandung TRIGA reactor can be operated by 20 fuel elements only. Compared with the original core, the new reactor core configuration is smaller and it results in some empty space that can be used for in-core irradiation facilities. Due to the existing of in-core irradiation facilities, the gamma heating value became a new factor that should be evaluated for safety analysis. For this reason, the gamma heating for TRIGA Bandung reactor using fuel plate was calculated by Gamset computer code. The calculations based on linear attenuation equations, line sources and gamma propagation on space. Calculations were also done for reflector positions (Lazy Susan irradiation facilities and central irradiation position (CIP, especially for any material samples. The calculation results show that gamma heating for CIP is significantly important (0,87 W/g, but very low value for Lazy Susan position (lest then 0,11 W/g. Based on this results, it can be concluded that the utilization of CIP as irradiation facilities need to consider of gamma heating as data for safety analysis report. Keywords: gamma heating, nuclear reactor, research reactor, reactor safety.   ABSTRAK Dengan dihentikannya produksi elemen bakar reaktor jenis Triga oleh produsen, maka semua reaktor TRIGA di dunia terganggu operasinya, termasuk juga reaktor TRIGA 2000 di Bandung. Untuk mendukung pengoperasian reaktor TRIGA Bandung

  19. Twenty years of operation of Ljubljana's TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Dimic, V.

    1986-01-01

    Twenty years have now passed since the start of the TRIGA Mark II reactor in Ljubljana. The reactor was critical on May 31, 1966. The total energy produced until the end of May 1986 was 14.048 MWh or 585 MWd. For the first 14 years (until 1981) the yearly energy produced was about 600 MWh, since 1981 the yearly energy produced was 1000 MWh when a routine radioactive isotopes production started for medical use as well as other industrial applications, such as doping and irradiation with fast neutrons of silicon monocrystals, production of level indicators (irradiated cobalt wire), production of radioactive iridium for gamma-radiography, leak detection in pipes by sodium, etc. Besides these, applied research around the reactor is being conducted in the following main fields, where- many unique methods have been developed or have found their way into the local industry or hospitals: neutron radiography, neutron induced auto-radiography using solid state nuclear track detectors, nondestructive methods for assessment of nuclear burn-up, neutron dosimetry, calculation of core burn-up for the optimal in-core fuel management strategy. The solvent extraction method was developed for the everyday production of 99m Tc, which is the most widely used radionuclide in diagnostic nuclear medicine. The methods were developed for the production of the following isotopes: 18 F, 85m Kr, 24 Na, 82 Br, 64 Zn, 125 I. Neutron activation analysis represents one of the major usages for the TRIGA reactor. Basic research is being conducted in the following main fields: solid state physics (elastic and inelastic scattering of the neutrons), neutron dosimetry, neutron radiography, reactor physics and neutron activation analysis. The reactor is used very extensively as a main instrument in the Reactor Training Centre in Ljubljana where manpower training for our nuclear power plant and other organisations has been performed. Although the reactor was designed very carefully in order to be used for

  20. Automated power control system for reactor TRIGA PUSPATI

    Science.gov (United States)

    Ghazali, Anith Khairunnisa; Minhat, Mohd Sabri; Hassan, Mohd Khair

    2017-01-01

    Reactor TRIGA PUSPATI (RTP) Mark II type undergoes safe operation for more than 30 years and the only research reactor exists in Malaysia. The main safety feature of Instrumentation and Control (I&C) system design is such that any failure in the electronic, or its associated components, does not lead to an uncontrolled rate of reactivity. The existed controller using feedback approach to control the reactor power. This paper introduces proposed controllers such as Model Reference Adaptive Control (MRAC) and Proportional Integral Derivatives (PID) controller for the RTP simulation. In RTP, the most important considered parameter is the reactor power and act as nervous system. To design a controller for complex plant like RTP is quite difficult due to high cost and safety factors cause by the failure of the controller. Furthermore, to overcome these problems, a simulator can be used to replace functions the hardware and test could then be simulated using this simulator. In order to find the best controller, several controllers were proposed and the result will be analysed for study the performances of the controller. The output result will be used to find out the best RTP power controller using MATLAB/Simulink and gives result as close as the real RTP performances. Currently, the structures of RTP was design using MATLAB/Simulink tool that consist of fission chamber, controller, control rod position, height-to-worth of control rods and a RTP model. The controller will control the control rod position to make sure that the reactivity still under the limitation parameter. The results given from each controller will be analysed and validated through experiment data collected from RTP.

  1. Control console conceptual design for sheet type fuels of Triga Mark-II reactor

    International Nuclear Information System (INIS)

    Eko Priyono; Kurnia Wibowo; Anang Susanto

    2016-01-01

    The control console conceptual design for sheet type fuel of TRIGA Mark-II reactor has been made. The control console conceptual design was made with refer study result of instrument and control system which is used in BATAN'S reactor i.e TRIGA-2000 Bandung, TRIGA Yogyakarta and MPR-30 Serpong. The control console conceptual design was made by using AutoCad software. The control console conceptual design reactor for sheet type fuel of TRIGA Mark-II reactor consist of 5 segments that is 3 segments for placing the computer monitors, 1 segment for placing bargraph displays and recorders and 1 segment for placing panel meters. There are the door on front and back position at each segment for enter and out devices in the console. The control console conceptual design is also equipped by the table along in front of console for placing reactor panel control and for writing, 3 drawers for 3 keyboards. The dimension of console will refer control room size and the components will be placed on console which will be detailed in detail design if this conceptual design has been approved. (author)

  2. Measurements of thermal and fast neutron fluxes at the TRIGA reactor

    International Nuclear Information System (INIS)

    Zerdin, F.; Grabovsek, Z.; Klinc, T.; Solinc, H.

    1966-01-01

    Gold foils were placed at different positions in the TRIGA reactor core and in the experimental devices. Absolute values of the thermal neutron flux at these positions were obtained by coincidence method. Preliminary fast neutron spectrum was measured by threshold detector and by 'Li 6 sandwich' detector. A short description of the applied method and obtained measurements results are included [sl

  3. Experience, status and future of the computerized reactor instrumentation at the TRIGA reactor Vienna

    International Nuclear Information System (INIS)

    Frankl, M.; Boeck, H.; Katrik, P.; Schachner, H.

    1997-01-01

    The paper describes the 33 years old history of the instrumentation of the TRIGA reactor Vienna and focuses on the present computerized instrumentation installed in 1992. The experience of three years of operation is discussed and some of the failures are analyzed. Potential future problems both with soft- and hardware as well as with spare part supplies are analyzed. (author). 6 figs

  4. Measuring temperature coefficient of TRIGA MARK I reactor by noise analysis

    International Nuclear Information System (INIS)

    Soares, P.A.

    1975-01-01

    The transfer function of TRIGA MARK I Reactor is measured at power zero (5w) and power 118Kw, in the frequency range of 0.02 to 0.5 rd/s. The method of intercorrelation between a pseudostochasticbinary signal is used. A simple dynamic model of the reactor is developed and the coefficient of temperature is estimated [pt

  5. Neutron Field Characterization of Irradiation Locations Applied to the Slovenian TRIGA Reactor

    International Nuclear Information System (INIS)

    Barbot, Loic; Domergue, Christophe; Breaud, Stephane; Destouches, Christophe; Villard, Jean-Francois; Snoj, Luka; Stancar, Ziga; Radulovic, Vladimir; Trkov, Andrej

    2013-06-01

    This work deals with several neutron flux measurement instruments and particle transport calculations combined in a method to assess the neutron field in experimental locations in nuclear reactor core or reflector. First test of this method in the TRIGA Mark II of Slovenia led to the assessment of three energy groups neutron fluxes in central irradiation locations within reactor core. (authors)

  6. Teaching and training applications of the Oregon State University Triga Reactor

    International Nuclear Information System (INIS)

    Dodd, B.

    1990-01-01

    This paper summarizes the recent use of the Oregon State University TRIGA Reactor (OSTR) for teaching and training. In particular, data spanning the last five years are presented which cover teaching through formal university classes, theses, public information, and school programs. Training is covered by presenting data on domestic and foreign reactor operator training, health physics training, and neutron activation analysis training

  7. Meteorological observation and radiological dose assessment from the decommission of TRIGA

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Eun Han; Han, Moon Hee; Suh, Kyung Suk; Hwang, Won Tae; Choi, Young Gil [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-06-01

    The radiational effect of the decommissioning of TRIGA research reactor on the environment was estimated. Annual joint frequency of wind data was calculated by statistical analysis of wind direction, wind speed and atmospheric stability. The annual effective dose for critical group (child) was estimated as 5.06 x 10{sup -6} mSv/yr at the site boundary on which the maximum atmospheric dispersion factor is represented. The estimated effective dose is 2.02 x 10{sup -3}% of the dose limit of 0.25 mSv/yr. The maximum whole body effective dose to the public at the site boundary for two hours following the accidental releases was 8.17 x 10{sup -10} Sv and 7.22 x 10{sup -10} Sv for TRIGA I and TRIGA II, respectively. These values are 3.27 x 10{sup -7}% and 2.89 x 10{sup -7}% of the dose limit of 3 Sv. As a conclusion, it can be said that the estimated exposure dose from both normal and accident of the decommissioning process of TRIGA research reactor is very small comparing with the dose limit. 19 refs., 5 figs., 32 tabs. (Author)

  8. In-service inspection and maintenance schedule for a typical TRIGA Mark-II reactor

    International Nuclear Information System (INIS)

    Boeck, H.

    1996-05-01

    This report lists all the systems and components of the TRIGA reactor Vienna which are inspected and maintained in regular intervals. These intervals are categorized in monthly, quarterly, semi-annual and annual inspections. Further the type of inspection and the responsibility for the inspection is shown. For each component specific inspection sheets have been developed, some examples are given in the annex. (author)

  9. Neutron optics experiments at the TRIGA Mark II reactor of the Atominstitut Wien

    International Nuclear Information System (INIS)

    Jericha, E.; Badurek, G.; Baron, M.; Hasegawa, Y.; Jaekel, M.; Klepp, J.; Rofner, A.; Sponar, S.; Trinker, M.; Villa, M.; Rauch, H.

    2004-01-01

    We present the layout and characteristics of the 3 neutron optics instruments located at the beam ports of the Vienna TRIGA reactor (hosted by the Atominstitut of the Austrian Universities, Vienna University of Technology) and the most recent experiments performed thereon. (author)

  10. PERHITUNGAN SUHU ELEMEN BAKAR REAKTOR TRIGA 2000 DALAM TABUNG SIPPING TEST MENGGUNAKAN CFD

    Directory of Open Access Journals (Sweden)

    K.A. Sudjatmi

    2015-03-01

    Full Text Available Telah dihitung suhu elemen bakar pada perangkat sipping test reaktor TRIGA 2000 Bandung. Perhitungan perlu dilakukan untuk memastikan bahwa suhu elemen bakar masih dibawah atau pada batas suhu elemen bakar yang diizinkan pada saat reaktor beroperasi, sehingga dapat dipastikan bahwa pada pelaksanaan pengujian dengan menggunakan perangkat ini, suhu masih dalam batas keselamatan. Perhitungan dilakukan dengan membuat model tabung sipping test berisi elemen bakar yang dikelilingi oleh 9 buah elemen bakar, sesuai dengan posisi tabung sipping test di teras reaktor, dengan menggunakan GAMBIT. Dimensi model disesuaikan dengan dimensi tabung dan elemen bakar dalam teras reaktor TRIGA 2000 Bandung. Pengoperasian sipping test untuk tiap elemen bakar dilakukan selama 30 menit pada daya 300 kW. Perhitungan dilakukan dengan menggunakan perangkat lunak Computational Fluid Dynamics (CFD dan sebagai inputan disesuaikan dengan parameter reaktor TRIGA 2000. Simulasi dilakukan pada pengoperasian dari 30, 60, 90, 120 150, 180 sampai 210 menit. Hasil perhitungan menunjukan bahwa suhu pusat bahan bakar dalam tabung sipping test sebesar 236,06 oC, sedangkan suhu dinding elemen bakar adalah sebesar 87,58 oC. Suhu maksimum pusat bahan bakar reaktor TRIGA 2000 pada operasi normal adalah 650 oC, dan tidak diizinkan terjadinya pendidihan di dalam teras reaktor. Jadi dapat disimpulkan bahwa pengoperasian perangkat sipping test masih sangat aman karena suhu bahan bakar berada dibawah batasan suhu bahan bakar yang diizinkan pada kondisi operasi normal demikian juga suhu dinding elemen bakar masih dibawah suhu didih air. Kata kunci: sipping test, TRIGA, elemen bakar, CFD   It has been calculated the fuel element temperature in the sipping test of Bandung TRIGA 2000 reactor. The calculation needs to be done to ascertain that the fuel element temperatures are below or at the limit of the allowable temperature fuel elements during reactor operation, ensuring that the

  11. Studies on decommissioning of TRIGA reactors and site restoration technologies in the Republic of Korea

    International Nuclear Information System (INIS)

    Oh, Won-Zin; Kim, Gye-Nam; Won, Hui-Jun

    2002-01-01

    Research and development on research reactor decommissioning and environmental restoration has been carried out at KAERI since 1997 to prepare for the decommissioning of KAERI's two TRIGA-type research reactors, which had been shut down since 1995. A 3-D graphic model of the TRIGA research reactor was built using IGRIP. The dismantling process was simulated in the graphic environment to verify the feasibility of individual operations before the execution of the remote dismantling process. An under-water wall-climbing robot, moving by propeller injection, and identifying its coordinates by using a laser sensor, was developed and tested in the TRIGA reactor pool by measuring a radioactive contamination map of the reactor surface. Using MODFLOW and TRIGA site geological data, a computer simulation of the underground migration of residual radionuclides, after the TRIGA reactor decommissioning, was carried out. It was found that the underground migration rate was very slow such that, when radionuclide decay and dilution are considered, the residual radionuclides will not have a significant environmental impact. The soil decontamination R and D, using soil washing, solvent flushing and electro-decontamination technologies, was carried out to determine the best method for decontaminating the soil waste accumulated in KAERI. The decontamination results indicated that, using the soil washing method, more than 80% of the soil wastes could be decontaminated well enough to discharge them to the environment. It was also determined that the control of solution pH and temperature in the soil washing process is important for the reduction of decontamination waste. Further decontamination, using an electro-kinetic decontamination method, was considered necessary for the residual soil waste, which consisted mainly of fine soil particles. (author)

  12. Transient rod failure in a pulsing TRIGA Mark I reactor

    International Nuclear Information System (INIS)

    Draper, E.L. Jr.; Atkinson, G.D. Jr.

    1972-01-01

    Full text: On July 7, 1970 the University of Texas at Austin TRIGA Mark I Pulsing Reactor experienced a failure of the transient control rod. Although no danger to personnel or damage to the reactor other than the pulse rod occurred, the failure was promptly reported to the USAEC regional compliance office. The first indication of an abnormal situation was unusual multiplication behavior during the first start-up of the day. As usual for steady state operation, the operator removed the transient rod and began to withdraw the shim and regulating rods. After partial withdrawal, he noticed that the count rate was not increasing as rapidly as was customary. While remaining at the console,the operator had a technician make a visual inspection of the core. The technician observed the transient drive rod was swinging freely in the pool and the poison section was detached. It was concluded, based on the indications of the.reactor instrumentation and visual inspection, that the transient control rod had broken off and remained in position in the core. The regulating and shim rods were inserted and the transient rod was manually cranked to the down position. The manual manipulation of the transient rod, instead of dropping the rod by gravity, was used so that the connecting rod could be reinserted in the control rod guide tube. The reactor core was then partially unloaded so that a critical mass was not present. The transient rod drive and connecting rod were removed from the pool. The poison section was retrieved from its position in the core by welding a tap to a long rod and tapping into the top of the poison section. Visual inspection of the poison section showed that the weld joining the male threads on the poison section to the main body of the control rod had failed. The threads remained screwed in the control rod drive shaft upon separation and the poison section remained fully inserted in the core. A new control rod was fabricated by Gulf General Atomic and shipped

  13. United States Domestic Research Reactor Infrastructure - TRIGA Reactor Fuel Support

    Energy Technology Data Exchange (ETDEWEB)

    Morrell, Douglas [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

    2008-10-29

    The purpose of the United State Domestic Research Reactor Infrastructure Program is to provide fresh nuclear reactor fuel to United States universities at no, or low, cost to the university. The title of the fuel remains with the United States government and when universities are finished with the fuel, the fuel is returned to the United States government. The program is funded by the United States Department of Energy - Nuclear Energy division, managed by Department of Energy - Idaho Field Office, and contracted to the Idaho National Laboratory's Management and Operations Contractor - Battelle Energy Alliance. Program has been at Idaho since 1977 and INL subcontracts with 26 United States domestic reactor facilities (13 TRIGA facilities, 9 plate fuel facilities, 2 AGN facilities, 1 Pulstar fuel facility, 1 Critical facility). University has not shipped fuel since 1968 and as such, we have no present procedures for shipping spent fuel. In addition: floor loading rate is unknown, many interferences must be removed to allow direct access to the reactor tank, floor space in the reactor cell is very limited, pavement ends inside our fence; some of the surface is not finished. The whole approach is narrow, curving and downhill. A truck large enough to transport the cask cannot pull into the lot and then back out (nearly impossible / refused by drivers); a large capacity (100 ton), long boom crane would have to be used due to loading dock obstructions. Access to the entrance door is on a sidewalk. The campus uses it as a road for construction equipment, deliveries and security response. Large trees are on both sides of sidewalk. Spent fuel shipments have never been done, no procedures approved or in place, no approved casks, no accident or safety analysis for spent fuel loading. Any cask assembly used in this facility will have to be removed from one crane, moved on the floor and then attached to another crane to get from the staging area to the reactor room

  14. Current activities at the FiR 1 TRIGA reactor

    International Nuclear Information System (INIS)

    Salmenhaara, Seppo

    2002-01-01

    The FiR 1 -reactor, a 250 kW Triga reactor, has been in operation since 1962. The main purpose to run the reactor is now the Boron Neutron Capture Therapy (BNCT). The epithermal neutrons needed for the irradiation of brain tumor patients are produced from the fast fission neutrons by a moderator block consisting of Al+AlF 3 (FLUENTAL), which showed to be the optimum material for this purpose. Twenty-one patients have been treated since May 1999, when the license for patient treatment was granted to the responsible BNCT treatment organization. The treatment organization has a close connection to the Helsinki University Central Hospital. The BNCT work dominates the current utilization of the reactor: three days per week for BNCT purposes and only two days per week for other purposes such as the neutron activation analysis and isotope production. In the near future the back end solutions of the spent fuel management will have a very important role in our activities. The Finnish Parliament ratified in May 2001 the Decision in Principle on the final disposal facility for spent fuel in Olkiluoto, on the western coast of Finland. There is a special condition in our operating license. We have now about two years' time to achieve a binding agreement between VTT and the Nuclear Power Plant Companies about the possibility to use the final disposal facility of the Nuclear Power Plants for our spent fuel. If this will not happen, we have to make the agreement with the USDOE with the well-known time limits. At the moment it seems to be reasonable to prepare for both spent fuel management possibilities: the domestic final disposal and the return to the USA offered by USDOE. Because the cost estimates of the both possibilities are on the same order of magnitude, the future of the reactor itself will determine, which of the spent fuel policies will be obeyed. In a couple of years' time it will be seen, if the funding of the reactor and the incomes from the BNC treatments will cover

  15. SANS facility at the Pitesti 14 MW Triga reactor

    International Nuclear Information System (INIS)

    Ionita, I.; Anghel, E.; Mincu, M.; Datcu, A.; Grabcev, B.; Todireanu, S.; Constantin, F.; Shvetsov, V.; Popescu, G.

    2006-01-01

    Full text of publication follows: At the present time, an important not yet fully exploited potentiality is represented by the SANS instruments existent at lower power reactors and reactors in developing countries even if they are, generally, endowed with a simpler equipment and are characterized by the lack of infrastructure to maintain and repair high technology accessories. The application of SANS at lower power reactors and in developing countries nevertheless is possible in well selected topics where only a restricted Q range is required, when scattering power is expected to be sufficiently high or when the sample size can be increased at the expense of resolution. Examples of this type of applications are: 1) Phase separation and precipitates in material science, 2) Ultrafine grained materials (nano-crystals, ceramics), 3) Porous materials such as concretes and filter materials, 4) Conformation and entanglements of polymer-chains, 5) Aggregates of micelles in microemulsions, gels and colloids, 6) Radiation damage in steels and alloys. The need for the installation of a new SANS facility at the Triga Reactor of the Institute of Nuclear Researches in Pitesti, Romania become actual especially after the shutting down of the VVRS Reactor from Bucharest. A monochromatic neutron beam with 1.5 Angstrom ≤ λ ≤ 5 Angstrom is produced by a mechanical velocity selector with helical slots.The distance between sample and detectors plane is (5.2 m ). The sample width may be fixed between 10 mm and 20 mm. The minimum value of the scattering vector is Q min = 0.005 Angstrom -1 while the maximal value is Q max = 0.5 Angstrom -1 . The relative error is ΔQ/Q min = 0.5. The cooperation partnership between advanced research centers and the smaller ones from developing countries could be fruitful. The formers act as mentors in solving specific problems. Such a partnership was established between INR Pitesti, Romania and JINR Dubna, Russia. The first step in this cooperation

  16. Operation experience and maintenance at the TRIGA Mark II L.E.N.A. reactor

    International Nuclear Information System (INIS)

    Gngoli, F.; Berzero, A.; Lana, F.; Rosti, G.; Meloni, S.

    2008-01-01

    The TRIGA Mark II reactor of the University of Pavia was operated in the last two years on a routine basis, mostly for neutron activation analysis purposes. Moreover the reactor was completely shutdown in the first six months of this year to allow the dismantling of the NADIR experimental setup. The paper presents: - Reactor operation from July 1990 to June 1992; - Reactor users in the time period January 1990 - December 1991; - Specific activities of some radionuclides in the filling materials; - Specific activity of some radionuclides in thermal column materials. Operations related to dismantling of NADIR experimental facility are described. Finally the new thermal column configuration is presented. Starting from the end inside the reactor tank, a graphite layer (35 cm thick) was positioned, followed by a bismuth layer (10 cm thick) to reduce gamma-ray intensity. The old graphite rods were then positioned leaving in the central part, on the equatorial plane of the thermal column, a cavity whose vertical section has 40 cm width and 20 cm height. The bottom of the cavity, towards to the reactor tank, has been lined with additional layers of graphite (10 cm), bismuth (10 cm) and again graphite (1 cm). The new configuration allowed new experiments to be performed. The cavity in the central part has been created to allow the irradiation of large biological samples such as experimental animal and human livers. This is a peculiar step in a neutron capture boron therapy project to be carried out at the University of Pavia. In order to avoid an implemented 41 Ar production in the void space between shutters and the thermal column outer end, the external surface of the thermal column has been coated with boral sheets. The neutron flux profile, both thermal and epithermal, and cadmium ratio for gold are shown. The flux distribution appears to be adequate to proceed with the neutron capture boron therapy experiment. The LENA Health Physics Service has checked all phases of

  17. Immobilization of ion exchange radioactive resins of the TRIGA Mark III nuclear reactor; Inmovilizacion de resinas de intercambio ionico radiactivas del reactor nuclear Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Garcia M, H.; Emeterio H, M.; Canizal S, C. [Instituto Nacional de Investigaciones Nucleares, A.P. 18-1027, C.P. 11801 Mexico D.F. (Mexico)

    2000-07-01

    This work has the objective to develop the process and to define the agglutinating material which allows the immobilization of the ion exchange radioactive resins coming from the TRIGA Mark III nuclear reactor contaminated with Ba-133, Co-60, Cs-137, Eu-152, and Mn-54 through the behavior analysis of different immobilization agents such as: bitumens, cement and polyester resin. According to the International Standardization the archetype samples were observed with the following tests: determination of free liquid, leaching, charge resistance, biodegradation, irradiation, thermal cycle, burned resistance. Generally all the tests were satisfactorily achieved, for each agent. Therefore, the polyester resin could be considered as the main immobilizing. (Author)

  18. Application of system-process-goal approach for description of TRIGA RC-1 system

    International Nuclear Information System (INIS)

    Gadomski, Adam M.

    1986-01-01

    The new methodology of the goal oriented description of an artificial system is presented. In the SPG approach (System-Process-Goal) it is assumed that the knowledge necessary for achieving the goal is available but it is not ordered or ordered for other purposes. The aim of SPG is to give the description of the analyzed system in form of network by decomposition of goal-system relationships using uniform and mathematical formalism. The SPG approach is useful to build a reactor operator aid system. This paper presents the conception of the application of the SPG approach to the decomposition of TRIGA RC-1 dynamics and for designing of TRIGA diagnostic algorithms. (author)

  19. The research reactor TRIGA Mainz. A neutron source for versatile applications in research and education

    International Nuclear Information System (INIS)

    Eberhardt, K.; Kronenberg, A.

    2000-01-01

    Currently, four research reactors with a thermal power ranging from 0.1 to 23 MW th are in operation in Germany and one new reactor (20 MW th ) is under construction. The TRIGA Mark II reactor at the Institut fuer Kernchemie became first critical on August 3, 1965. It can be operated in the steady state mode with a maximum power of 100 kW th and in the pulse mode with a peak power of 250 MW th . A survey of the research programmes carried out at the TRIGA Mainz is given covering a wide range of applications in basic and applied science in nuclear chemistry, nuclear- and particle physics. Furthermore, the reactor is used for neutron activation analysis and for education and training of students and technical personal. (orig.) [de

  20. Startup tests for TRIGA-ACPR at the Japan Atomic Energy Research Institute

    International Nuclear Information System (INIS)

    West, G.; Whittemore, W.

    1976-01-01

    The JAERI ACPR TRIGA startup tests involved procedures somewhat different from those considered standard for TRIGA. While the approach to critical followed standard procedures, the tests involving (1) the core loading to the permitted excess reactivity, and (2) the calibration of the 11 control rods (six Reg, two Safety, three Transient rods) were somewhat unusual. The power calibration involved three techniques: reactor noise analysis, flux foil activation, and calorimetry. The pulsing tests involved the insertion of increasing amounts of excess reactivity until the predicted performance was reached with a total insertion of $4.70. For,this the peak measured fuel temperature was 850 o C, and the integrated prompt energy release was about 100 megawatt-second, both in good agreement with predictions. (author)

  1. Feasibility studies of a mixed core using standard TRIGA and FLIP fuel

    International Nuclear Information System (INIS)

    Randall, J.D.; Hardt, M.; Feltz, D.E.

    1972-01-01

    The continued burnup of the Texas A and M TRIGA reactor produced a core that was becoming too large and producing lower fluxes. Since this was affecting important experimental programs it was obvious that some action must be taken. Several alternatives were considered including continuing with standard fuel. A second alternative would be to return to MTR fuel which would mean losing the important advantages of TRIGA fuel Since the cost of a total FLIP core was above the amount that was considered to be available, it was decided to investigate the possibilities of a mixed FLIP-standard core. A safety analysis is made for those characteristics that are significantly different from the standard core

  2. Tests for removal of Co-60 and Eu-154 from irradiated graphite in the TRIGA Reactor

    International Nuclear Information System (INIS)

    Arsene, Carmen

    2009-01-01

    The irradiated graphite in Romania is mainly generated in the thermal columns of TRIGA and WWER-S research reactors (about 9 tones). It was found that the radionuclide content of the graphite irradiated in the TRIGA research reactor is mainly due to C-14 (103 Bq/g), Eu-152 (600-700 Bq/g) and Co-60 (130-150 Bq/g) and low amounts of Eu-154 and Cs-137, depending on location in the thermal column and on irradiation history. In order to minimize the waste inventory and volume in view of their final disposal, in the present paper we show the results of experiments performed for developing and optimizing methods for the chemical decontamination of the irradiated graphite. These procedures are based on strong alkaline solutions for Eu-152 and strong acid solutions for Co-60. The influence of the process parameters on the decontamination factor is investigated. (authors)

  3. Applicable regulations and development of surveillance experiments of criticality approach in the TRIGA III Mark reactor

    International Nuclear Information System (INIS)

    Gonzalez M, J.L.; Aguilar H, F.; Rivero G, T.; Sainz M, E.

    2000-01-01

    In the procedure elaborated to repair the vessel of TRIGA III Mark reactor is required to move toward two tanks of temporal storage the fuel elements which are in operation and the spent fuel elements which are in decay inside the reactor pool. The National Commission of Nuclear Safety and Safeguards (CNSNS) has requested as protection measure that it is carried out a surveillance of the criticality approach of the temporal storages. This work determines the main regulation aspects that entails an experiment of criticality approach, moreover, informing about the results obtained in the developing of this experiments. The regulation aspects are not exclusives for this work in the TRIGA Mark III reactor but they also apply toward any assembling of fissile material. (Author)

  4. Forensic INAA of bullet-lead and shotshell-pellet evidence specimens with a TRIGA reactor

    International Nuclear Information System (INIS)

    Guinn, Vincent P.

    1988-01-01

    This paper has been published earlier, in the references cited. The main purpose of this paper is to acquaint interested TRIGA reactor groups with the main features of the Forensic INAA of BL and SSP evidence specimens - and to recommend that they consider acquiring the necessary expertise and then provide such analysis services to law enforcement agencies, public defenders, and defence attorneys in their respective areas

  5. Technology development and demonstration for TRIGA research reactor decontamination, decommissioning and site restoration

    International Nuclear Information System (INIS)

    Oh, Won Zin; Jung, Ki Jung; Lee, Byung Jik

    1997-01-01

    This paper describes the introduction to research reactor decommissioning plan at KAERI, the background of technology development and demonstration, and the current status of the system decontamination technology for TRIGA reactors, concrete decontamination and dust treatment technologies, wall ranging robot and graphic simulation of dismantling processes, soil decontamination and restoration technology, recycling or reuse technologies for radioactive metallic wastes, and incineration technology demonstration for combustible wastes. 9 figs

  6. TRIGA MARK II first research reactor facility in Kingdom of Morocco

    International Nuclear Information System (INIS)

    Nacir, B.

    2010-01-01

    The research reactor facility is located at Centre d'Etudes Nucleaires de la Maamora(CENM), located approximately 25 kilometers north of the city of Rabat. This facility will enable CNESTEN, as the operating organization, to fulfil its missions for promotion of nuclear technology in Morocco, contribute to the implementation of a national nuclear power program, and assist the state in monitoring nuclear activities for protection of the public and environment. The reactor building include TRIGA Mark II research reactor with an initial power level of 2000kW (t), and equipped for a planned future upgrade to 3,000-kilowatts.The facility is the keystone structure of CENM, and contain in addition to the TRIGA research reactor, extensively equipped laboratories and all associate support systems, structures, and supply facilities with the support of the AIEA, French CEA and LLNL (USA). The CENM with its TRIGA reactor and fully equipped laboratories will give the kingdom of Morocco its first nuclear installation with extensive capabilities. These will include the production of radioisotopes for medical, industrial and environmental uses, metallurgy and chemistry, implementation of nuclear analytical techniques such as neutron activation analysis and non-destructive examination techniques, as well as carrying out basic research programs in solid state and reactor physics. The feedback from the commissioning and the implementation of the safety standards during this phase was very interesting from safety point of view. The TRIGA Mark II research reactor at CENM achieved initial criticality on May 2, 2007 at 13:30 with 71 fuel elements and culminated with the successful completion of the full power endurance testing on 6 September, 2007.

  7. Conceptual design of fuel transfer cask for Reactor TRIGA PUSPATI (RTP)

    Energy Technology Data Exchange (ETDEWEB)

    Muhamad, Shalina Sheik [Prototype and Plant Development Center, Technical Support Division, Malaysian Nuclear Agency, Bangi, 43000, Kajang, Selangor (Malaysia); Hamzah, Mohd Arif Arif B. [Prototype and Plant Development Center, Technical Support Division Malaysian Nuclear Agency, Bangi, 43000, Kajang, Selangor (Malaysia)

    2014-02-12

    Spent fuel transfer cask is used to transfer a spent fuel from the reactor tank to the spent fuel storage or for spent fuel inspection. Typically, the cask made from steel cylinders that are either welded or bolted closed. The cylinder is enclosed with additional steel, concrete, or other material to provide radiation shielding and containment of the spent fuel. This paper will discuss the Conceptual Design of fuel transfer cask for Reactor TRIGA Puspati (RTP)

  8. Return of TRIGA fuel from the Medical University of Hanover (MHH) to the United States

    International Nuclear Information System (INIS)

    Hampel, Gabriele; Klaus, Uwe; Schmidt, Thomas

    1999-01-01

    The Medical University of Hanover (MHH) returned its TRIGA fuel to the United States in the summer of 1999. This paper deals with the procedure for handling the fuel elements within and outside the reactor facility. It describes the dry loading technology, taking into account the special conditions relevant to the MHH. It also includes the time scale for both the various steps of the procedure and the entire process, as well as the main results of the radiological surveys. (author)

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

    International Nuclear Information System (INIS)

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

    1985-12-01

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

  10. Calculation of fuel element temperature TRIGA 2000 reactor in sipping test tubes using CFD

    International Nuclear Information System (INIS)

    Sudjatmi KA

    2013-01-01

    It has been calculated the fuel element temperature in the sipping test of Bandung TRIGA 2000 reactor. The calculation needs to be done to ascertain that the fuel element temperatures are below or at the limit of the allowable temperature fuel elements during reactor operation. ensuring that the implementation of the test by using this device, the temperature is still within safety limits. The calculation is done by making a model sipping test tubes containing a fuel element surrounded by 9 fuel elements. according to the position sipping test tubes in the reactor core. by using Gambit. Dimensional model adapted to the dimensions of the tube and the fuel element in the reactor core of Bandung TRIGA 2000 reactor. Sipping test Operation for each fuel element performed for 30 minutes at 300 kW power. Calculations were performed using CFD software and as input adjusted parameters of TRIGA 2000 reactor. Simulations carried out on the operation of the 30, 60, 90, 120, 150, 180 and 210 minutes. The calculation result shows that the temperature of the fuel in tubes sipping test of 236.06 °C, while the temperature of the wall is 87.58 °C. The maximum temperature in the fuel center of TRIGA 2000 reactor in normal operation is 650 °C. and the boiling is not allowed in the reactor. So it can be concluded that the operation of the sipping test device are is very safe because the fuel center temperature is below the temperature limits the allowable fuel under normal operating conditions as well as the fuel element wall temperature is below the boiling temperature of water. (author)

  11. The 10 MW multipurpose TRIGA reactor at Ongkharak Nuclear Research Center, Thailand

    International Nuclear Information System (INIS)

    Thurgood, B.E.; Razvi, J.; Whittemore, J.L.; Bhadrakom, K.

    1997-01-01

    General Atomics (GA), has been selected to lead a team of firms from the United States, Japan, Australia and Thailand to design, build and commission the Ongkharak Nuclear Research Center near Bangkok, Thailand, for the Office of Atomic Energy for Peace. The facilities to be provided comprise of: A Reactor Island, consisting of a 10 MW TRIGA reactor that takes full advantage of the inherent safety characteristics of uranium-zirconium hydride (UZrH) fuel; An Isotope Production Facility for the production of radioisotopes and radiopharmaceuticals using the TRIGA reactor; A Waste Processing and Storage Facility for the processing and storage of radioactive waste from the facility as well as other locations in Thailand. The centerpiece of the Center will be the TRIGA reactor, fueled with low-enriched UZrH fuel, cooled and moderated by light water, and reflected by beryllium and heavy water. The UZrH fueled reactor will have a rated steady state thermal power output of 10 MW, and will be capable of performing the following: Radioisotope production for medical, industrial and agricultural uses; Neutron transmutation doping of silicon; Beam experiments such as Neutron Scattering, Neutron Radiography (NR), and Prompt Gamma Neutron Activation Analysis (PGNAA); Medical therapy of patients using Boron Neutron Capture Therapy (BNCT); Applied research and technology development in the nuclear field; Training in principles of reactor operation, reactor physics, reactor experiments, etc. (author)

  12. Validation of neutron flux redistribution factors in JSI TRIGA reactor due to control rod movements

    International Nuclear Information System (INIS)

    Kaiba, Tanja; Žerovnik, Gašper; Jazbec, Anže; Štancar, Žiga; Barbot, Loïc; Fourmentel, Damien; Snoj, Luka

    2015-01-01

    For efficient utilization of research reactors, such as TRIGA Mark II reactor in Ljubljana, it is important to know neutron flux distribution in the reactor as accurately as possible. The focus of this study is on the neutron flux redistributions due to control rod movements. For analyzing neutron flux redistributions, Monte Carlo calculations of fission rate distributions with the JSI TRIGA reactor model at different control rod configurations have been performed. Sensitivity of the detector response due to control rod movement have been studied. Optimal radial and axial positions of the detector have been determined. Measurements of the axial neutron flux distribution using the CEA manufactured fission chambers have been performed. The experiments at different control rod positions were conducted and compared with the MCNP calculations for a fixed detector axial position. In the future, simultaneous on-line measurements with multiple fission chambers will be performed inside the reactor core for a more accurate on-line power monitoring system. - Highlights: • Neutron flux redistribution due to control rod movement in JSI TRIGA has been studied. • Detector response sensitivity to the control rod position has been minimized. • Optimal radial and axial detector positions have been determined

  13. Role of decommissioning plan and its progress for the PUSPATI TRIGA Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zakaria, Norasalwa, E-mail: norasalwa@nuclearmalaysia.gov.my; Mustafa, Muhammad Khairul Ariff, E-mail: norasalwa@nuclearmalaysia.gov.my; Anuar, Abul Adli, E-mail: norasalwa@nuclearmalaysia.gov.my; Idris, Hairul Nizam, E-mail: norasalwa@nuclearmalaysia.gov.my; Ba' an, Rohyiza, E-mail: norasalwa@nuclearmalaysia.gov.my [Malaysian Nuclear Agency, 43000 Kajang, Selangor (Malaysia)

    2014-02-12

    Malaysian nuclear research reactor, the PUSPATI TRIGA Reactor, reached its first criticality in 1982, and since then, it has been serving for more than 30 years for training, radioisotope production and research purposes. Realizing the age and the need for its decommissioning sometime in the future, a ground basis of assessment and an elaborative project management need to be established, covering the entire process from termination of reactor operation to the establishment of final status, documented as the Decommissioning Plan. At international level, IAEA recognizes the absence of Decommissioning Plan as one of the factors hampering progress in decommissioning of nuclear facilities in the world. Throughout the years, IAEA has taken initiatives and drawn out projects in promoting progress in decommissioning programmes, like CIDER, DACCORD and R2D2P, for which Malaysia is participating in these projects. This paper highlights the concept of Decommissioning plan and its significances to the Agency. It will also address the progress, way forward and challenges faced in developing the Decommissioning Plan for the PUSPATI TRIGA Reactor. The efforts in the establishment of this plan helps to provide continual national contribution at the international level, as well as meeting the regulatory requirement, if need be. The existing license for the operation of PUSPATI TRIGA Reactor does not impose a requirement for a decommissioning plan; however, the renewal of license may call for a decommissioning plan to be submitted for approval in future.

  14. Decontamination and decommissioning project status of the TRIGA Mark II and III in Korea

    International Nuclear Information System (INIS)

    Paik, S.T.; Park, S.K.; Chung, K.W.; Chung, U.S.; Jung, K.J.

    1999-01-01

    TRIGA Mark-II, the first research reactor in Korea, has operated since 1962, and the second one, TRIGA Mark-III since 1972. Both of them had their operation phased out in 1995 due to their lives and operation of the new research reactor, HANARO (High-flux Advanced Neutron Application Reactor) at the Korea Atomic Energy Institute (KAERI) in Taejon. Decontamination and decommissioning (D and D) project of TRIGA Mark-II and Mark-III was started in January 1997 and will be completed in December 2002. The first year of the project, work was performed in preparation of the decommissioning plan, start of the environmental impact assessment and setup licensing procedure and documentation for the project with cooperation of Korea Institute of Nuclear Safety (KINS). Hyundai Engineering Company (HEC) is the main contractor to do design and licensing documentation for the D and D of both reactors. British Nuclear Fuels plc (BNFL) is the technical assisting partner of HEC. The decommissioning plan document was submitted to the Ministry of Since and Technology (MOST) for the decommissioning license in December 1998, and it expecting to be issued a license in mid 1999. The goal of this project is to release the reactor site and buildings as an unrestricted area. This paper summarizes current status and future plan for the D and D project. (author)

  15. Role of decommissioning plan and its progress for the PUSPATI TRIGA Reactor

    Science.gov (United States)

    Zakaria, Norasalwa; Mustafa, Muhammad Khairul Ariff; Anuar, Abul Adli; Idris, Hairul Nizam; Ba'an, Rohyiza

    2014-02-01

    Malaysian nuclear research reactor, the PUSPATI TRIGA Reactor, reached its first criticality in 1982, and since then, it has been serving for more than 30 years for training, radioisotope production and research purposes. Realizing the age and the need for its decommissioning sometime in the future, a ground basis of assessment and an elaborative project management need to be established, covering the entire process from termination of reactor operation to the establishment of final status, documented as the Decommissioning Plan. At international level, IAEA recognizes the absence of Decommissioning Plan as one of the factors hampering progress in decommissioning of nuclear facilities in the world. Throughout the years, IAEA has taken initiatives and drawn out projects in promoting progress in decommissioning programmes, like CIDER, DACCORD and R2D2P, for which Malaysia is participating in these projects. This paper highlights the concept of Decommissioning plan and its significances to the Agency. It will also address the progress, way forward and challenges faced in developing the Decommissioning Plan for the PUSPATI TRIGA Reactor. The efforts in the establishment of this plan helps to provide continual national contribution at the international level, as well as meeting the regulatory requirement, if need be. The existing license for the operation of PUSPATI TRIGA Reactor does not impose a requirement for a decommissioning plan; however, the renewal of license may call for a decommissioning plan to be submitted for approval in future.

  16. Burn up calculations and validation by gamma scanning of a TRIGA HEU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Khan, R. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan).; Karimzadeh, S.; Boeck, H.; Villa, M.; Stummer, T. [Vienna Univ. of Technology (Austria). Atominstitut

    2013-03-15

    The TRIGA Mark II research reactor operated by Atominstitut (Vienna/Austria) is one of the few TRIGA reactors, which still utilizes several High Enriched Uranium (HEU) Zirconium-Hydride (U-Zr-H) fuel elements. Its current core is a completely mixed core with 3 different types of fuel elements including one HEU type with 70 % enrichment and a stainless steel cladding. The present paper calculates the burn up of the FLIP (Fuel Lifetime Improvement Program) fuel using the burn up code ORIGEN2 and validates the theoretical results by high resolution gamma spectrometry using a unique fuel scanning device (FSD) developed at the Atominstitut especially for TRIGA fuel. For this purpose a FLIP fuel element was removed from the reactor core and stored in the research reactor pool for an appropriate cooling period. The fuel element was then transferred into the fuel scanning device to determine the Cesium-137 isotope distribution along the axis of the fuel element. The comparison between theoretical predictions and experimental results is the highlight of the present paper. (orig.)

  17. Neutron flux characterisation of the Pavia TRIGA Mark II research reactor for radiobiological and microdosimetric applications.

    Science.gov (United States)

    Alloni, D; Prata, M; Salvini, A; Ottolenghi, A

    2015-09-01

    Nowadays the Pavia TRIGA reactor is available for national and international collaboration in various research fields. The TRIGA Mark II nuclear research reactor of the Pavia University offers different in- and out-core neutron irradiation channels, each characterised by different neutron spectra. In the last two years a campaign of measurements and simulations has been performed in order to guarantee a better characterisation of these different fluxes and to meet the demands of irradiations that require precise information on these spectra in particular for radiobiological and microdosimetric studies. Experimental data on neutron fluxes have been collected analysing and measuring the gamma activity induced in thin target foils of different materials irradiated in different TRIGA experimental channels. The data on the induced gamma activities have been processed with the SAND II deconvolution code and finally compared with the spectra obtained with Monte Carlo simulations. The comparison between simulated and measured spectra showed a good agreement allowing a more precise characterisation of the neutron spectra and a validation of the adopted method. © The Author 2015. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  18. Fission products detection in irradiated TRIGA fuel by means of gamma spectroscopy and MCNP calculation.

    Science.gov (United States)

    Cagnazzo, M; Borio di Tigliole, A; Böck, H; Villa, M

    2018-05-01

    Aim of this work was the detection of fission products activity distribution along the axial dimension of irradiated fuel elements (FEs) at the TRIGA Mark II research reactor of the Technische Universität (TU) Wien. The activity distribution was measured by means of a customized fuel gamma scanning device, which includes a vertical lifting system to move the fuel rod along its vertical axis. For each investigated FE, a gamma spectrum measurement was performed along the vertical axis, with steps of 1 cm, in order to determine the axial distribution of the fission products. After the fuel elements underwent a relatively short cooling down period, different fission products were detected. The activity concentration was determined by calibrating the gamma detector with a standard calibration source of known activity and by MCNP6 simulations for the evaluation of self-absorption and geometric effects. Given the specific TRIGA fuel composition, a correction procedure is developed and used in this work for the measurement of the fission product Zr 95 . This measurement campaign is part of a more extended project aiming at the modelling of the TU Wien TRIGA reactor by means of different calculation codes (MCNP6, Serpent): the experimental results presented in this paper will be subsequently used for the benchmark of the models developed with the calculation codes. Copyright © 2018 Elsevier Ltd. All rights reserved.

  19. Determining Coolant Flow Rate Distribution In The Fuel-Modified TRIGA Plate Reactor

    Science.gov (United States)

    Puji Hastuti, Endiah; Widodo, Surip; Darwis Isnaini, M.; Geni Rina, S.; Syaiful, B.

    2018-02-01

    TRIGA 2000 reactor in Bandung is planned to have the fuel element replaced, from cylindrical uranium and zirconium-hydride (U-ZrH) alloy to U3Si2-Al plate type of low enriched uranium of 19.75% with uranium density of 2.96 gU/cm3, while the reactor power is maintained at 2 MW. This change is planned to anticipate the discontinuity of TRIGA fuel element production. The selection of this plate-type fuel element is supported by the fact that such fuel type has been produced in Indonesia and used in MPR-30 safely since 2000. The core configuration of plate-type-fuelled TRIGA reactor requires coolant flow rate through each fuel element channel in order to meet its safety function. This paper is aimed to describe the results of coolant flow rate distribution in the TRIGA core that meets the safety function at normal operation condition, physical test, shutdown, and at initial event of loss of coolant flow due power supply interruption. The design analysis to determine coolant flow rate in this paper employs CAUDVAP and COOLODN computation code. The designed coolant flow rate that meets the safety criteria of departure from nucleate boiling ratio (DNBR), onset of flow instability ratio (OFIR), and ΔΤ onset of nucleate boiling (ONB), indicates that the minimum flow rate required to cool the plate-type fuelled TRIGA core at 2 MW is 80 kg/s. Therefore, it can be concluded that the operating limitation condition (OLC) for the minimum flow rate is 80 kg/s; the 72 kg/s is to cool the active core; while the minimum flow rate for coolant flow rate drop is limited to 68 kg/s with the coolant inlet temperature 35°C. This thermohydraulic design also provides cooling for 4 positions irradiation position (IP) utilization and 1 central irradiation position (CIP) with end fitting inner diameter (ID) of 10 mm and 20 mm, respectively.

  20. Graphical user interfaces for McClellan Nuclear Radiation Center

    International Nuclear Information System (INIS)

    Brown-VanHoozer, S.A.; Power, M.; Forsmann, H.

    1998-01-01

    The control console of the TRIGA reactor at McClellan's Nuclear Radiation Center (MNRC) is in the process of being replaced because of spurious scrams, outdated software, and obsolete parts. The intent of the new control console is to eliminate the existing problems by installing a UNIX-based computer system with industry-standard interface software and by incorporating human factors during all stages of the graphical user interface (GUI) development and control console design. This paper gives a brief description of some of the guidelines used in developing the MNRC's GUIs as continuous, real-time displays

  1. Neutronics modeling of TRIGA reactor at the University of Utah using agent, KENO6 and MCNP5 codes

    International Nuclear Information System (INIS)

    Yang, X.; Xiao, S.; Choe, D.; Jevremovic, T.

    2010-01-01

    The TRIGA reactor at the University of Utah is modelled in 2D using the AGENT state-of-the-art methodology based on the Method of Characteristics (MOC) and R-function theory supporting detailed reactor analysis of reactor geometries of any type. The TRIGA reactor is also modelled using KENO6 and MCNP5 for comparison. The spatial flux and reaction rates distribution are visualized by AGENT graphics support. All methodologies are in use in to study the effect of different fuel configurations in developing practical educational exercises for students studying reactor physics. At the University of Utah we train graduate and undergraduate students in obtaining the Nuclear Regulatory Commission license in operating the TRIGA reactor. The computational models as developed are in support of these extensive training classes and in helping students visualize the reactor core characteristics in regard to neutron transport under various operational conditions. Additionally, the TRIGA reactor is under the consideration for power uprate; this fleet of computational tools once benchmarked against real measurements will provide us with validated 3D simulation models for simulating operating conditions of TRIGA. (author)

  2. Development of Virtual Simulation, Data Logging and Supervisory Control System (DLSC) for Laboratory Scale of Mini Reactor TRIGA PUSPATI (RTP)

    International Nuclear Information System (INIS)

    Mohd Sabri Minhat; Izhar Abu Hussin; Zamrul Faizal Omar

    2014-01-01

    The Reactor TRIGA PUSPATI (RTP) type TRIGA Mark II was installed in the year 1982. The scope for this research paper is to produce a detail design for development of Virtual Simulation, Data Logging and Supervisory Control System (DLSC) for Laboratory Scale of Mini Reactor TRIGA PUSPATI (RTP). The purpose of this project is to develop a Laboratory Scale of Mini Reactor TRIGA PUSPATI (RTP) which consists of cooling system, Data Logging and Supervisory Control system (DLSC), and Operator Workstation (OWS) for education and training in reactor engineering. Initial research, Automated Controller System for Controlling Reactivity (ACSCR) was developed in Science Fund project (03-03-01-SF0159) entitled Development of Automated Controller System for Controlling Reactivity by using Field Programmable Gate Array (FGGA) in Research Reactor Application and Radiation Facility. The ACSCR is developed to control the movement of the control rods by new algorithm movement method. The Data Acquisition and Control System (DACS) of ACSCR is setup and installed by using FPGA/Real-Time Controller. The software development in this ACSCR is programmed using LabVIEW which consist of simulation and control program. This mini reactor project will extended and integrate the developed ACSCR include DACS with cooling system and OWS. The development of this Laboratory Scale of Mini Reactor TRIGA PUSPATI (RTP) is part of facility for centre of excellence program in reactor technology. (author)

  3. Determination of the energy spectrum of the neutrons in the central thimble of the reactor core TRIGA Mark III

    International Nuclear Information System (INIS)

    Parra M, M. A.

    2014-01-01

    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 197 , Ni 58 , In 115 , Mg 24 , Al 27 , Fe 58 , Co 59 and Cu 63 , they were selected to cover the full range the energies (10 -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 research projects. (Author)

  4. The history and perspective of Romania-USA cooperation in the field of technologic transfer of TRIGA reactor concept

    International Nuclear Information System (INIS)

    Ciocaanescu, M.; Ionescu, M.

    1996-01-01

    The cooperation between Romania and the USA in the field of technologic transfer of nuclear research reactor technology began with the steady state 14 MW t TRIGA reactor, installed at INR Pitesti, Romania. It is the first in the range of TRIGA reactors proposed as a materials testing reactor. The first criticality was reached in November 19, 1979 and first operation at 14 MW t level was in February 1980. The paper will present the short history of this cooperation and the perspective for a new cooperation for building a Nuclear Heating Plant using the TRIGA reactor concept for demonstration purpose. The energy crisis is a world-wide problem which affects each country in different ways because the resources and the consumption are unfairly distributed. World-wide research points out that the fossil fuel sources are not to be considered the main energy sources for the long term as they are limited

  5. Analysis concerning the perspective of Romania-USA technological cooperation with a view to performing TRIGA reactor project

    International Nuclear Information System (INIS)

    Ciocanescu, M.; Ionescu, M.; Constantin, L.

    1998-01-01

    The co-operation between Romania and the USA in the field of technologic transfer of nuclear research reactor technology began with the steady state 14 MW, TRIGA reactor, installed at INR Pitesti, Romania. It is the first in the range of TRIGA reactors proposed as a materials testing reactor. The first criticality was reached in November 19, 1979 and first operation at 14 MW, level was in February 1980. The paper will present the short history of this co-operation and the perspective for a new co-operation for building a Nuclear Heating Plant using the TRIGA reactor concept for demonstration purpose. The energy crisis is a world-wide problem which affects each country in different ways because the resources and the consumption are unfairly distributed. World-wide research points out that the fossil fuel sources are not to be considered the main energy sources for the long term as they are limited. (author)

  6. Simulator of the punctual kinetics of a TRIGA Mark III nuclear reactor with diffuse control of power in a visual environment

    International Nuclear Information System (INIS)

    Perez M, C.

    2004-01-01

    The development of a software that simulates the punctual kinetics of a nuclear research reactor model TRIGA Mark III, generating the answers of the reactor low different algorithms of control of power is presented. The user requires a graphic interface that allows him easily interacting with the pretender. To achieve the proposed objective, first the system was modeled in open knot, not using a mathematical model of the consistent reactor in a system of ordinary differential equations lineal. For their solution in real time the numeric method of Runge-Kutta-Fehlberg was used. As second phase, it was modeled to the system in closed knot, using for it an algorithm of control of the power based on fuzzy logic. Taking into account the graphic characteristics detailed in the requirements of the system (chapter 4), it was chosen to develop the pretender the language of Visual programming Basic 6.0. The program fulfills the time of execution satisfactorily, assisting to the necessity of visualizing the behavior in real time of the reactor, and it responds from an effective way to the petitions of changes of power on the part of the user. (Author)

  7. Assessment results of the South Korea TRIGA SNF to be shipped to INEEL

    International Nuclear Information System (INIS)

    Cole, C.M.; Dirk, W.J.; Cottam, R.E.; Paik, S.T.

    1997-01-01

    This paper describes the Training, Research, Isotope, General Atomics (TRIGA) spent nuclear fuel (SNF) examination at the Seoul and the Taejon Research Reactor Facilities in South Korea. The examination was required before the SNF would be accepted for transportation and storage at the INEEL. The results of the aluminum and stainless steel clad TRIGA fuel examination have been summarized. A description of the examination team training, the examination work plan and examination equipment is also included. This paper also explains the technical basis for the examination and physical condition criteria used to determine what, if any, additional packaging would be required for transportation and for the receipt and storage of the fuel at the INEEL. This paper delineates the preparation activities prior to the fuel examinations and includes (1) collecting spent fuel data; (2) preparatory work by the Korean Atomic Energy Research Institute (KAERI) for fuel examination: (3) preparation of a radionuclide report, Radionuclide Mass Inventory, Activity, Decay Heat, and Dose Rate Parametric Data for TRIGA Spent Nuclear Fuels needed to provide input data for transportation and fuel acceptance at the Idaho National Engineering and Environmental Laboratory (INEEL); (4) gathering FRR Facility data; and (5) coordination between the INEEL and KAERI. Included, are the unanticipated conditions encountered in the unloading of fuel from the dry storage casks in Taejon in preparation for examination, a description of the damaged condition of the fuel removed from the casks, and the apparent cause of the damages. Lessons learned from all the activities are also addressed. A brief description of the preparatory work for the shipment of the spent fuel from Korea to INEEL is included

  8. Neutronics analysis of the initial core of the TRIGA Mark II reactor

    Energy Technology Data Exchange (ETDEWEB)

    Khan, R., E-mail: rustamzia@yahoo.co [Vienna University of Technology, Atominstitute (ATI), Stadion allee 2, A-1020, Vienna (Austria); Stummer, T.; Boeck, H.; Villa, M. [Vienna University of Technology, Atominstitute (ATI), Stadion allee 2, A-1020, Vienna (Austria)

    2011-05-15

    Highlights: The TRIGA Mark II Vienna is modeled employing MCNP5. The model is confirmed through three different experiments. Initial critical, reactivity distribution and flux mapping experiment. - Abstract: The Atominstitute (ATI) of Vienna University of Technology (VUT) operates a TRIGA Mark II research reactor since March 1962. Its initial criticality was achieved on 7th March 1962 when 57th Fuel Element (FE) was loaded to the core. This paper describes the development of the MCNP model of the TRIGA reactor and its validation through three different experiments i.e. initial criticality, reactivity distribution and a thermal flux mapping experiment in the reactor core. All these experiments were performed on the initial core configuration. The MCNP model includes all necessary core components i.e. FE, Graphite Element GE, neutron Source Element (SE), Central IRradiation channel (CIR) etc. Outside the core, this model simulates the annular grooved graphite reflector, the thermal and thermalizing column, four beam tubes and the reactor water tank up to 100 cm in radial and +60 and -60 cm in axial direction. Each grid position at its exact location is modeled. This model employs the ENDF/B-VI data library except for the Sm-isotopes which are taken from JEFF 3.1 because ENDF/B-VI lacks samarium (Sm) cross sections. For the first experiment, the model predicts an effective multiplication factor ({kappa}{sub eff}) of 1.00183 with an estimated standard deviation 0.00031 which is very close to the experimental value 1.00114. The second experiment measures the reactivity values of four FE and one GE. In comparison to the MCNP results, the percent difference ranges from 4 to 22. The third experiment verifies the model at a local level with the radial and axial thermal flux density distribution in the core. Though the trends are similar, the MCNP model overestimates the radial thermal flux density in the core and underestimates these results at the core periphery.

  9. The contribution of a small triga university research reactor to nuclear research on an international level

    International Nuclear Information System (INIS)

    Villa, M.; Boeck, H.; Weber, H.W.

    2001-01-01

    The paper focuses especially on the important results in neutron- and solid state physics and the co-operation between the low power TRIGA reactor with high flux neutron sources in Europe such as the Institute Laue-Langevin (ILL) in Grenoble, the Paul Scherrer Institut (PSI) in Villigen, the Rutherford Appleton Laboratory (RAL) in Didcot and the Research Center Juelich. Experiments are set up for test purposes at the TRIGA reactor and then transferred to the powerful neutron sources. Different new perfect silicon channel-cut and interferometer crystals are prepared and then tested at the Bonse-Hart camera, which is a double crystal (or triple axis) diffractometer and at the interferometer set-up. Historically, the first verification of neutron interferometry at a perfect crystal device has been achieved at the 250 kW TRIGA-reactor in Vienna in the year 1974. Also the co-operation with the PSI and the TU Munich in the field of neutron radiography and neutron tomography and VESTA, an experiment for storing cold neutrons with a wavelength of 6.27 A, installed at the pulsed neutron source ISIS at RAL will be mentioned. The second topic treated in this paper shows the international co-operation in the field of superconductors. This research work is carried out under two European TMR-Network programs. The third topic in this paper focuses on the co-operation in the field of safeguard. Several projects have been carried out during the past years in co-operation with the IAEA such as establishing a gamma spectrum reference catalogue for CdZnTe detectors and tests of safeguard video cameras under neutron irradiation. Further an integrated safeguard surveillance network composed of a video camera, a gamma monitor and a neutron monitor is under development. (orig.)

  10. Neutron flux measurement in the central channel (XC-1) of TRIGA 14 MW LEU core

    International Nuclear Information System (INIS)

    BARBOS, D.; BUSUIOC, P.; ROTH, Cs.; PAUNOIU, C.

    2008-01-01

    The TRIGA 14 MW reactor, operated by Institute for Nuclear Research Pitesti, Romania, is a pool type reactor, and has a rectangular shape which holds fuel bundles and is surrounded with beryllium reflectors. Each fuel bundle is composed of 25 nuclear fuel rods. The TRIGA 14 MW reactor was commissioned 28 years ago with HEU fuel rods. The conversion was gradually achieved, starting in February 1992 and completed in March 2006. The full conversion of the 14 MW TRIGA Research Reactor was completed in May 2006 and each step of the conversion was achieved by removal of HEU fuel, replaced by LEU fuel, accompanied by a large set of theoretical evaluation and physical measurements intended to confirm the performances of gradual conversion. After the core full conversion, a program of measurements and comparisons with previous results of core physics and measurements is underway, allowing data acquisition for normal operation, demonstration of safety and economics of the converted core. Neutron flux spectrum measurements in the XC in the XC-1 water 1 water-filled channel were performed using multi multi-foil activation techniques. The neutron spectra and flux are obtained by unfolding from measured reaction rates using SAND II computer code. The integral neutron flux value for LEU core is greater of 13% than for the standard HEU core. Also thermal neutron flux value for converted LEU core is smaller by 0.38% than for the standard HEU core. These differences appear because the foil activation detectors have been irradiated using a pneumatic rabbit having a diameter of 32 mm, whereas foil irradiations in standard HEU core has been performed with a pneumatic rabbit having a diameter of 14 mm, and therefore the neutron spectra in LEU core is less thermalized and the weight of fast neutron is greater

  11. Operational experience with the TRIGA reactor of the University of Pavia

    International Nuclear Information System (INIS)

    Borio di Tigliole, A.; Alloni, D.; Cagnazzo, M.; Coniglio, M.; Lana, F.; Losi, A.; Magrotti, G.; Manera, S.; Marchetti, F.; Pappalardo, P.; Prata, M.; Salvini, A.; Scian, G.; Vinciguerra, G.

    2008-01-01

    The TRIGA Mark II research reactor of the University of Pavia is in operation since 1965. The annual operational time at nominal power (250 kW) is in the range of 300 - 400 hours depending upon the time schedule of some experiments and research activities. The reactor is mainly used for NAA activities and BNCT research. Few tens of hours per year are dedicated also to electronic devices irradiation and student training courses. Few homemade upgrading of the reactor were realized in the past two years: components of the secondary/tertiary cooling circuit were substituted and a new radiation area monitoring system was installed. Also the Instrumentation and Control (I and C) system was almost completely refurbished. The presentation describes the major extraordinary maintenance activities implemented and the status of main reactor systems: - The I and C System: complete substitution, channel-by-channel without changing the operating and safety logics; - Tertiary and secondary water-cooling circuits: complete substitution of the tertiary water-cooling circuit and partial substitution of the components of the secondary water-cooling circuit; - Reactor Building Air Filtering and Ventilation System: installation of a computerized air filtering and ventilation system; - Radiation Area Monitoring System: new system based on a commercial micro-computer and an home-made software developed on Lab-View platform. The system is made of a network of different instruments coupled, trough a serial bus line RS232, with a data acquisition station; - Fuel Elements: at the moment, the core is made of 48 Aluminium clad and 34 SST clad TRIGA fuel elements controlled periodically for their elongation and/or bowing. All components and systems undergo ordinary maintenance according to the Technical Prescriptions and to the 'Good Practice Procedures'. In summary, the TRIGA reactor of the University of Pavia shows a very good technical state and, at the moment, there are no political or

  12. Argon-41 production and evolution at the Oregon State University TRIGA Reactor (OSTR)

    International Nuclear Information System (INIS)

    Anellis, L.G.; Johnson, A.G.; Higginbotham, J.F.

    1988-01-01

    In this study, argon-41 concentrations were measured at various locations within the reactor facility to assess the accuracy of models used to predict argon-41 evolution from the reactor tank, and to determine the relationship between argon gas evolution from the tank and subsequent argon-41 concentrations throughout the reactor room. In particular, argon-41 was measured directly above the reactor tank with the reactor tank lids closed, at other accessible locations on the reactor top with the tank lids both closed and open, and at several locations on the first floor of the reactor room. These measured concentrations were then compared to values calculated using a modified argon-41 production and evolution model for TRIGA reactor tanks and ventilation values applicable to the OSTR facility. The modified model was based in part on earlier TRIGA models for argon-41 production and release, but added features which improved the agreement between predicted and measured values. The approximate dose equivalent rate due to the presence of argon-41 in reactor room air was calculated for several different locations inside the OSTR facility. These dose rates were determined using the argon-41 concentration measured at each specific location, and were subsequently converted to a predicted quarterly dose equivalent for each location based on the reactor's operating history. The predicted quarterly dose equivalent values were then compared to quarterly doses measured by film badges deployed as dose-integrating area radiation monitors at the locations of interest. The results indicate that the modified production and evolution model is able to predict argon-41 concentrations to within a factor of ten when compared to the measured data. Quarterly dose equivalents calculated from the measured argon-41 concentrations and the reactor's operating history seemed consistent with results obtained from the integrating area radiation monitors. Given the argon-41 concentrations measured

  13. Assessment results of the South Korea TRIGA SNF to be shipped to INEEL

    International Nuclear Information System (INIS)

    Cole, Charles M.; Dirk, Willam J.; Cottam, Russel E.; Paik, Sam T.

    1997-01-01

    This paper describes the Training, Research, Isotope, General Atomics (TRIGA) spent nuclear fuel (SNF) examination at the Seoul and the Taejon Research Reactor Facilities in South Korea. The examination was required before the SNF would be accepted for transportation and storage at the INEEL. The results of the aluminum and stainless steel clad TRIGA fuel examination have been summarized. A description of the examination team training, the examination work plan and examination equipment is also included. This paper also explains the technical basis for the examination and physical condition criteria used to determine what, if any, additional packaging (canning) would be required for transportation and for the receipt and storage of the fuel at the INEEL. This paper delineates the preparation activities prior to the fuel examinations and includes (1) collecting spent fuel data; (2) preparatory work by the Korean Atomic Energy Research Institute (KAERI) for fuel examination: (3) preparation of a radionuclide report, 'Radionuclide Mass Inventory, Activity, Decay Heat, and Dose Rate Parametric Data for TRIGA Spent Nuclear Fuels' needed to provide input data for transportation and fuel acceptance at the Idaho National Engineering and Environmental Laboratory (INEEL); (4) gathering FRR Facility data; (5) preparation of Appendix A; (6) and coordination between the INEEL and KAERI. Included, are the unanticipated conditions encountered in the unloading of fuel from the dry storage casks in Taejon in preparation for examination, a description of the damaged condition of the fuel removed from the casks, and the apparent cause of the damages. Lessons learned from all the activities are also addressed. A brief description of the preparatory work for the shipment of the spent fuel from Korea to INEEL is included. (author)

  14. The contribution of a small TRIGA university research reactor to nuclear research on an international level

    International Nuclear Information System (INIS)

    Villa, M.; Bastuerk, M.; Boeck, H.

    2002-01-01

    The paper focuses especially on the important results in neutron- and solid state physics and the co-operation between the low power TRIGA reactor with high flux neutron sources in Europe such as the Institute Laue-Langevin (ILL) in Grenoble, the Paul Scherrer Institut (PSI) in Villigen, the Rutherford Appleton Laboratory (RAL) in Didcot and the Research Center Juelich. Experiments are set up for test purposes at the TRIGA reactor and then transferred to the powerful neutron sources. Different new perfect silicon channel-cut and interferometer crystals are prepared and then tested at the Bonse-Hart camera, which is a double crystal (or triple axis) diffractometer and at the interferometer set-up. Historically, the first verification of neutron interferometry at a perfect crystal device has been achieved at the 250 kW TRIGA-reactor in Vienna in the year 1974. Also the co-operation with the PSI and the TU Munich in the field of neutron radiography and neutron tomography and VESTA, an experiment for storing cold neutrons with a wavelength of 6.27A, installed at the pulsed neutron source ISIS at RAL are mentioned. The second topic in this paper focuses on the co-operation in the field of safeguard. Several projects have been carried out during the past years in co-operation with the IAEA such as establishing a gamma spectrum reference catalogue for CdZnTe detectors and tests of safeguard video cameras under neutron irradiation. Further an integrated safeguard surveillance network composed of a video camera, a gamma monitor and a neutron monitor is under development

  15. Simulator of the punctual kinetics of a TRIGA Mark III reactor with power diffuse control in a visual environment; Simulador de la cinetica puntual de un reactor nuclear TRIGA Mark III con control difuso de potencia en un ambiente visual

    Energy Technology Data Exchange (ETDEWEB)

    Perez M, C

    2004-07-01

    The development of a software is presented that simulates the punctual kinetics of a nuclear reactor of investigation model TRIGA Mark III, generating the answers of the reactor low different algorithms of control of power. The user requires a graphic interface that allows him easily interacting with the simulator. To achieve the proposed objective, first the system was modeled in open loop, not using a mathematical model of the consistent reactor in a system of linear ordinary differential equations. For their solution in real time the numeric method of Runge-Kutta-Fehlberg was used. As second phase, it was modeled to the system in closed loop, using for it an algorithm of control of the power based on fuzzy logic. This software has as purpose to help the investigator in the control area who will be able to prove different algorithms for the control of the power of the reactor. This is achieved using the code source in language C, C++, Visual Basic, with which a file is generated. DLL and it is inserted in the simulator. Then they will be able to visualize the results as if their controller had installed in the reactor, analyzing the behavior of all his variables that will be stored in files, for his later study. The easiness of proving these control algorithms in the reactor without necessity to make it physically has important consequences as the saving in the expense of fuel, the not generation of radioactive waste and the most important thing, one doesn't run any risk. The simulator can be used how many times it is necessary until the total purification of the algorithm. This program is the base for following investigation processes, enlarging the capacities and options of the same one. The program fulfills the time of execution satisfactorily, assisting to the necessity of visualizing the behavior in real time of the reactor, and it responds from an effective way to the petitions of changes of power on the part of the user. (Author)

  16. Whole core neutronics modeling of a TRIGA reactor using integral transport theory

    International Nuclear Information System (INIS)

    Schwinkendorf, K.N.; Toffer, H.

    1990-01-01

    An innovative analysis approach for performing whole core reactor physics calculations for TRIGA reactors has been employed recently at the Westinghouse Hanford Company. A deterministic transport theory model with sufficient geometric complexity to evaluate asymmetric loading patterns was used. Calculations of this complexity have been performed in the past using Monte Carlo simulation, such as the MCNP code. However, the Monte Carlo calculations are more difficult to prepare and require more computer time. On the Hanford Site CRAY XMP-18 computer, the new methods required less than one-third of the central processing unit time per calculation as compared to an MCNP calculation using 100,000 neutron histories

  17. Calculation of fundamental parameters for the dynamical study of TRIGA-3-Salazar reactor (Mixed reactor core)

    International Nuclear Information System (INIS)

    Viais J, J.

    1994-01-01

    Kinetic parameters for dynamic study of two different configurations, 8 and 9, both with standard fuel, 20% enrichment and Flip (Fuel Life Improvement Program with 70% enrichment) fuel, for TRIGA Mark-III reactor from Mexico Nuclear Center, are obtained. A calculation method using both WIMS-D4 and DTF-IV and DAC1 was established, to decide which of those two configurations has the best safety and operational conditions. Validation of this methodology is done by calculate those parameters for a reactor core with new standard fuel. Configuration 9 is recommended to be use. (Author)

  18. Operations of a TRIGA reactor at a small private liberal arts college

    International Nuclear Information System (INIS)

    Church, L.B.

    1978-01-01

    A small private liberal arts college is not a very representative place to have a TRIGA reactor. Reed is a wholly undergraduate institution with a strong emphasis in the traditional liberal arts and fundamental sciences. Many of the larger state universities provide an excellence in nuclear science which is often presented to students in a somewhat distant manner. By providing a reactor that was immediately accessible to undergraduate students it has been realized that the excitement attendant with nuclear science would be available to them in an immediate hands-on manner

  19. Mechanism design for the control rods conduction of TRIGA Mark III reactor in the NINR

    International Nuclear Information System (INIS)

    Franco C, A.

    1997-01-01

    This work presents in the first chapter a general studio about the reactor and the importance of control rods in the reactor , the mechaniucal design attending to requisitions that are imposed for conditions of operation of the reactor are present in the second chapter, the narrow relation that exists with the new control console and the mechanism is developed in the thired chapter, this relation from a point of view of an assembly of components is presents in fourth chapter, finally reaches and perspectives of mechanism forming part of project of the automation of reactor TRIGA MARK III, are present in the fifth chapter. (Author)

  20. Collimator and shielding design for boron neutron capture therapy (BNCT) facility at TRIGA MARK II reactor

    International Nuclear Information System (INIS)

    Mohd Rafi Mohd Solleh; Abdul Aziz Tajuddin; Abdul Aziz Mohamed; Eid Mahmoud Eid Abdel Munem; Mohamad Hairie Rabir; Julia Abdul Karim; Yoshiaki, Kiyanagi

    2011-01-01

    The geometry of reactor core, thermal column, collimator and shielding system for BNCT application of TRIGA MARK II Reactor were simulated with MCNP5 code. Neutron particle lethargy and dose were calculated with MCNPX code. Neutron flux in a sample located at the end of collimator after normalized to measured value (Eid Mahmoud Eid Abdel Munem, 2007) at 1 MW power was 1.06 x 10 8 n/ cm 2 / s. According to IAEA (2001) flux of 1.00 x 10 9 n/ cm 2 / s requires three hours of treatment. Few modifications were needed to get higher flux. (Author)

  1. Neuro-diffuse algorithm for neutronic power identification of TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Rojas R, E.; Benitez R, J. S.; Segovia de los Rios, J. A.; Rivero G, T.

    2009-10-01

    In this work are presented the results of design and implementation of an algorithm based on diffuse logic systems and neural networks like method of neutronic power identification of TRIGA Mark III reactor. This algorithm uses the punctual kinetics equation as data generator of training, a cost function and a learning stage based on the descending gradient algorithm allow to optimize the parameters of membership functions of a diffuse system. Also, a series of criteria like part of the initial conditions of training algorithm are established. These criteria according to the carried out simulations show a quick convergence of neutronic power estimated from the first iterations. (Author)

  2. Performance of a new small-angle neutron scattering instrument at the Malaysian TRIGA reactor

    International Nuclear Information System (INIS)

    Sufi, M.A.M.; Radiman, S.; Wiedenmann, A.; Mortensen, K.

    1997-01-01

    The set-up and alignment of a new small-angle neutron scattering (SANS) instrument, installed at the 1 MW light-water-moderated MINT TRIGA research reactor, are described. The wavelength distribution and the flux at the sample position have been determined. First neutron scattering measurements were made on two reference samples with strong scattering power; the results prove that the SANS signal is well reproduced on the instrument when samples of typical size are used, despite the high level of the background of fast and epithermal neutrons. (orig.)

  3. Experience with effluent release from the Omaha V. A. Hospital TRIGA reactor

    International Nuclear Information System (INIS)

    Blotcky, A.J.

    1974-01-01

    The effluent release from experiments is controlled by limiting the size of each sample irradiated so that if it was accidentally completely volatized into the closed room, the radioactive concentration would not exceed the permitted limits. The possible releases of Ar-41 and N-16 from the reactor are also considered. The experimentally determined levels of radiation around the Omaha facility are shown. From the data and calculations it was concluded that the levels of effluent release from the Omaha TRIGA are very small

  4. The development of quality assurance program in Reactor TRIGA PUSPATI (RTP)

    International Nuclear Information System (INIS)

    Rosli Darmawan; Mohd Rizal Mamat; Mohamad Zaid Mohamad; Mohd Ridzuan Abdul Mutalib

    2007-01-01

    One of the trivial issues in the operation of Nuclear Reactor is the safety of the system. Worldwide publicity on a few nuclear accidents as well as the notorious Hiroshima and Nagasaki bombing has always bring about general public fear on anything related to nuclear. IAEA has always emphasized on the assurance of nuclear safety for all nuclear installations and activities. According to the IAEA safety guides, all research reactors are required to implement quality assurance programs to ensure the conduct of operations are in accordance with the safety standards required. This paper discusses the activities carried out toward the establishment of Quality Assurance Program for Reaktor TRIGA PUSPATI (RTP). (Author)

  5. Evaluation of WIMS-D/4 nuclear data library used on TRIGA reactor calculation

    International Nuclear Information System (INIS)

    Chen Wei; Xie Zhongsheng; Jiang Xinbiao; Chen Da

    1997-01-01

    The 69 groups constants of H in ZrH, 166 Er and 167 Er generated by NJOY and GASKET codes are inserted into WIMS nuclear data library WIMS-CNDC and WIMS-NINT libraries used on RTIGA reactor calculation are obtained. In order to check WIMS-CNDC and WIMS-NINT libraries, the scattering cross-section is compared with that in WIMS-IJS library. The group constant, K ∞ and temperature coefficient are calculated by using WIMS-CNDC, WIMS-NINT and WIMS-IJS. The results show the both libraries are suitable for calculation of TRIGA reactor

  6. Evaluation Of Fire Safety And Protection At PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Ahmad Nabil Ab Rahim; Alfred Sanggau Ligam; Nurhayati Ramli; Mohd Fazli Zakaria; Naim Syauqi Hamzah; Phongsakorn Prak; Mohammad Suhaimi Kassim; Zarina Masood

    2014-01-01

    Fire hazard is one of many risks that can affect the safety operation of PUSPATI TRIGA Reactor. Reactor building in Malaysian Nuclear Agency was built in 1980s and the fire system has been introduced since then. The evaluation of the fire safety system at this time is important to ensure the efficiency of fire prevention, fighting and mitigation task that probably occurs. This evaluation involves with the fire fighting system and equipment, integrity of the system from the perspective of management and equipment, fire fighting procedure and fire fighting response team. (author)

  7. Large Object Irradiation Facility In The Tangential Channel Of The JSI TRIGA Reactor

    CERN Document Server

    Radulovic, Vladimir; Kaiba, Tanja; Kavsek, Darko; Cindro, Vladimir; Mikuz, Marko; Snoj, Luka

    2017-01-01

    This paper presents the design and installation of a new irradiation device in the Tangential Channel of the JSI TRIGA reactor in Ljubljana, Slovenia. The purpose of the device is to enable on-line irradiation testing of electronic components considerably larger in size (of lateral dimensions of at least 12 cm) than currently possible in the irradiation channels located in the reactor core, in a relatively high neutron flux (exceeding 10^12 n cm^-2 s^-1) and to provide adequate neutron and gamma radiation shielding.

  8. Neutronics analysis of the proposed 25-MW leu TRIGA Multipurpose Research Reactor

    International Nuclear Information System (INIS)

    Nurdin, M.; Bretscher, M.M.; Snelgrove, J.L.

    1982-01-01

    More than two years ago the government of Indonesia announced plans to purchase a research reactor for the Puspiptek Research Center in Serpong Indonesia to be used for isotope production, materials testing, neutron physics measurements, and reactor operator training. Reactors using low-enriched uranium (LEU) plate-type and rod-type fuel elements were considered. This paper deals with the neutronic evaluation of the rod-type 25-MW LEU TRIGA Multipurpose Research Reactor (MPRR) proposed by the General Atomic Company of the United States of America

  9. The Boron Neutron Capture Therapy (BNCT) Project at the TRIGA Reactor in Mainz, Germany

    DEFF Research Database (Denmark)

    Hampel, G.; Grunewald, C.; Schütz, C.

    2011-01-01

    The thermal column of the TRIGA reactor in Mainz is being used very effectively for medical and biological applications. The BNCT (boron neutron capture therapy) project at the University of Mainz is focussed on the treatment of liver tumours, similar to the work performed at Pavia (Italy) a few...... have also been initiated to investigate radiobiological effects of radiation generated during BNCT. For both experiments and treatment, a reliable dosimetry system is necessary. From work elsewhere, the use of alanine detectors appear to be an appropriate dosimetry technique....

  10. Immobilization of ion exchange radioactive resins of the TRIGA Mark III nuclear reactor

    International Nuclear Information System (INIS)

    Garcia M, H.; Emeterio H, M.; Canizal S, C.

    1999-01-01

    This work has the objective to develop the process and to define the agglutinating material which allows the immobilization of the ion exchange radioactive resins coming from the TRIGA Mark III nuclear reactor contaminated with Ba-133, Co-60, Cs-137, Eu-152, and Mn-54 through the behavior analysis of different immobilization agents such as: bitumens, cement and polyester resin. According to the International Standardization the archetype samples were observed with the following tests: determination of free liquid, leaching, charge resistance, biodegradation, irradiation, thermal cycle, burned resistance. Generally all the tests were satisfactorily achieved, for each agent. Therefore, the polyester resin could be considered as the main immobilizing. (Author)

  11. On-site emergency intervention plan for nuclear accident situation at SCN-Pitesti TRIGA Reactor

    International Nuclear Information System (INIS)

    Margeanu, S.; Oprea, I.

    2008-01-01

    A 14 MW TRIGA research reactor is operated on the Institute for Nuclear Research site. In the event of a nuclear accident or radiological emergency that may affect the public the effectiveness of protective actions depends on the adequacy of intervention plans prepared in advance. Considerable planning is necessary to reduce to manageable levels the types of decisions leading to effective responses to protect the public in such an event. The essential structures of our on-site, off-site and county emergency intervention plan and the correlation between emergency intervention plans are presented. (authors)

  12. Remodeling of the instrumentation and control system of TRIGA MARK II reactor

    International Nuclear Information System (INIS)

    Koh, B.J.; Shim, B.S.; Han, S.J.; Yun, W.Y.; Kwon, Y.J.; Kim, J.K.

    1982-01-01

    It was required to remodel the instrumentation and control system of TRIGA MARK II reactor installed in 1962 due to the difficulty in acquiring spare parts, and aging of the equipments. The instrumentation and control system were designed and manufactured in order to normalize the operation of reactor by using semiconductors and integrated circuits instead of the old-fashioned tube-type instruments for the first time in our country. In consideration of human factors engineering and the physical standards of the Korean, the control console was also designed. (Author)

  13. Time Evolution of Selected Actinides in TRIGA MARK-II Fuel

    International Nuclear Information System (INIS)

    Usang, M.D.; Naim Shauqi Hamzah; Mohamad Hairie Rabir

    2011-01-01

    Study is made on the evolution of several actinides capable of undergoing fission or breeding available on the Malaysian Nuclear Agency (MNA) TRIGA MARK-II fuel. Population distribution of burned fuel in the MNA reactor is determined with a model developed using WIMS. This model simulates fuel conditions in the hottest position in the reactor, thus the location where most of the burn up occurs. Theoretical basis of these nuclide time evolution are explored and compared with the population obtained from our models. Good agreements are found for the theoretical time evolution and the population of Uranium-235, Uranium-236, Uranium-238 and Plutonium-239. (author)

  14. An Overview of Ageing Management Programme for PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Syahirah Abdul Rahman; Mohamad Azman Che Mat Isa; Mohd Zaid Mohamed

    2011-01-01

    The PUSPATI TRIGA reactor (RTP) at Malaysian Nuclear Agency which has been operating for 29 years now faces increasingly serious aging problems. Many components are obsolete whereas genuine parts are no longer in the market. Currently, the aging problem is addressed through periodic maintenance on all systems, structures and components (SSC). As a holistic measure, the Aging Management Program (AMP) was formulated to solve the problems from the grassroots. This paper describes the first stage of the AMP which identifies the strengths and capabilities. This includes identifying the types of aging, responsible parties and relationship between aging problems and safety of RTP. (author)

  15. Calculations and selection of a TRIGA core for the Nuclear Reactor IAN-R1

    International Nuclear Information System (INIS)

    Castiblanco, L.A.; Sarta, J.A.

    1997-01-01

    The Reactor Group used the code WIMS reduced to five groups of energy, together with the code CITATION, and evaluated four configurations for a core, according to the grid actually installed. The four configurations were taken from the two proposals presented to the Instituto de Ciencias Nucleares y Energias Alternativas by General Atomics Company. In this paper, the Authors selected the best configuration according to the performance of flux distribution and excess reactivity, for a TRIGA core to be installed in the Nuclear Reactor IAN-R1

  16. Database of Samples Irradiated at Reactor TRIGA PUSPATI (RTP)

    International Nuclear Information System (INIS)

    Muhd Husamuddin Abdul Khalil; Mohd Amin Sharifuldin Salleh; Julia Abdul Karim

    2011-01-01

    Evaluation has been made to data of irradiated samples for the type of sample requested for activation at RTP. Sample types are grouped with percentage of total throughputs to rule out the weight percent of every respective group. The database consists of radionuclide inventory of short and long half-life and high activity radionuclides such as Br and Au have been identified and that database has been constructed using a user-friendly Microsoft Access. Through this, trend of gamma exposure will easily be evaluated at experimental facilities and could ensure radiological effect towards safety and health is limited per Radiation Protection (Basic Safety Standard) Regulation 1988. This database places an important parameter to improve management system in acquiring information of the samples irradiated at RTP and will enhance the safety assurance and reliability of the experimental design basis. (author)

  17. Simulator of the punctual kinetics of a TRIGA Mark III reactor with power diffuse control in a visual environment

    International Nuclear Information System (INIS)

    Perez M, C.

    2004-01-01

    The development of a software is presented that simulates the punctual kinetics of a nuclear reactor of investigation model TRIGA Mark III, generating the answers of the reactor low different algorithms of control of power. The user requires a graphic interface that allows him easily interacting with the simulator. To achieve the proposed objective, first the system was modeled in open loop, not using a mathematical model of the consistent reactor in a system of linear ordinary differential equations. For their solution in real time the numeric method of Runge-Kutta-Fehlberg was used. As second phase, it was modeled to the system in closed loop, using for it an algorithm of control of the power based on fuzzy logic. This software has as purpose to help the investigator in the control area who will be able to prove different algorithms for the control of the power of the reactor. This is achieved using the code source in language C, C++, Visual Basic, with which a file is generated. DLL and it is inserted in the simulator. Then they will be able to visualize the results as if their controller had installed in the reactor, analyzing the behavior of all his variables that will be stored in files, for his later study. The easiness of proving these control algorithms in the reactor without necessity to make it physically has important consequences as the saving in the expense of fuel, the not generation of radioactive waste and the most important thing, one doesn't run any risk. The simulator can be used how many times it is necessary until the total purification of the algorithm. This program is the base for following investigation processes, enlarging the capacities and options of the same one. The program fulfills the time of execution satisfactorily, assisting to the necessity of visualizing the behavior in real time of the reactor, and it responds from an effective way to the petitions of changes of power on the part of the user. (Author)

  18. 78 FR 5840 - Notice of License Termination for University of Illinois Advanced TRIGA Reactor, License No. R-115

    Science.gov (United States)

    2013-01-28

    ... University of Illinois Advanced TRIGA Reactor, License No. R-115 The U.S. Nuclear Regulatory Commission (NRC) is noticing the termination of Facility Operating License No. R-115, for the University of Illinois... Operating License No. R-115 is terminated. The above referenced documents may be examined, and/or copied for...

  19. 77 FR 7613 - Dow Chemical Company; Dow Chemical TRIGA Research Reactor; Facility Operating License No. R-108

    Science.gov (United States)

    2012-02-13

    ... Chemical TRIGA Research Reactor; Facility Operating License No. R-108 AGENCY: Nuclear Regulatory Commission... Facility Operating License No. R-108 (``Application''), which currently authorizes the Dow Chemical Company... License No. R-108 for the DTRR. The application contains SUNSI. Based on its initial review of the...

  20. Verification of Monte Carlo calculations of the neutron flux in typical irradiation channels of the TRIGA reactor, Ljubljana

    NARCIS (Netherlands)

    Jacimovic, R; Maucec, M; Trkov, A

    2003-01-01

    An experimental verification of Monte Carlo neutron flux calculations in typical irradiation channels in the TRIGA Mark II reactor at the Jozef Stefan Institute is presented. It was found that the flux, as well as its spectral characteristics, depends rather strongly on the position of the

  1. Conceptual design of a clinical BNCT beam in an adjacent dry cell of the Jozef Stefan Institute TRIGA reactor

    NARCIS (Netherlands)

    Maucec, M

    2000-01-01

    The MCNP4B Monte Carlo transport code is used in a feasibility study of the epithermal neutron boron neutron capture therapy facility in the thermalizing column of the 250-kW TRIGA Mark II reactor at the Jozef Stefan Institute (JSI). To boost the epithermal neutron flux at the reference irradiation

  2. Successful completion of a time sensitive MTR and TRIGA Indonesian shipment

    International Nuclear Information System (INIS)

    Anne, Catherine; Patterson, John; Messick, Chuck

    2005-01-01

    Early this year, a shipment of 109 MTR fuel assemblies was received at the Department of Energy's Savannah River Site from the BATAN reactor in Serpong, Indonesia and another of 181 TRIGA fuel assemblies was received at the Idaho National Laboratory from the two BATAN Indonesian TRIGA reactors in Bandung and Yogyakarta, Indonesia. These were the first Other-Than- High-Income Countries shipments under the FRR program since the Spring 2001. The Global Threat Reduction Initiative announced by Secretary Abraham will require expeditious scheduling and extreme sensitivity to shipment security. The subject shipments demonstrated exceptional performance in both respects. Indonesian terrorist acts and 9/11 impacted the security requirements for the spent nuclear fuel shipments. Internal Indonesian security issues and an upcoming Indonesian election led to a request to perform the shipment with a very short schedule. Preliminary site assessments were performed in November 2003. The DOE awarded a task order to NAC for shipment performance just before Christmas 2003. The casks departed the US in January and the fuel elements were delivered at the DOE sites by the end of April 2004. The paper will present how the team completed a successful shipment in a timely manner. (author)

  3. Modeling a TRIGA Mark II reactor using the Attila three-dimensional deterministic transport code

    International Nuclear Information System (INIS)

    Keller, S.T.; Palmer, T.S.; Wareing, T.A.

    2005-01-01

    A benchmark model of a TRIGA reactor constructed using materials and dimensions similar to existing TRIGA reactors was analyzed using MCNP and the recently developed deterministic transport code Attila TM . The benchmark reactor requires no MCNP modeling approximations, yet is sufficiently complex to validate the new modeling techniques. Geometric properties of the benchmark reactor are specified for use by Attila TM with CAD software. Materials are treated individually in MCNP. Materials used in Attila TM that are clad are homogenized. Attila TM uses multigroup energy discretization. Two cross section libraries were constructed for comparison. A 16 group library collapsed from the SCALE 4.4.a 238 group library provided better results than a seven group library calculated with WIMS-ANL. Values of the k-effective eigenvalue and scalar flux as a function of location and energy were calculated by the two codes. The calculated values for k-effective and spatially averaged neutron flux were found to be in good agreement. Flux distribution by space and energy also agreed well. Attila TM results could be improved with increased spatial and angular resolution and revised energy group structure. (authors)

  4. Assessment of a RELAP5 model for the IPR-R1 TRIGA research reactor

    International Nuclear Information System (INIS)

    Reis, Patricia A.L.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Mesquita, Amir Z.; Soares, Humberto V.

    2010-01-01

    RELAP5 code was developed at the Idaho National Environmental and Engineering Laboratory and it is widely used for thermal hydraulic studies of commercial nuclear power plants and, currently, it has been also applied for thermal hydraulic analysis of nuclear research systems with good predictions. This work is a contribution to the assessment of RELAP5/3.3 code for research reactors analysis. It presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 and 100 kW. The reactor is located at the Nuclear Technology Development Centre (CDTN), Brazil. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of code-to-data validation. The RELAP5 results were also compared with calculation performed using the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The use of a cross flow model has been essential to improve results in the transient condition respect to preceding investigations.

  5. Thermal power evaluation of the TRIGA nuclear reactor at CDTN in operations of long duration

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Zacarias; Ferreira, Andrea Vidal; Rezende, Hugo Cesar, E-mail: amir@cdtn.b, E-mail: avf@cdtn.b, E-mail: hcr@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2010-07-01

    The standard operations of nuclear research reactor IPR-R1 TRIGA located at CDTN (Belo Horizonte) usually have duration of not more than 8h. However in 2009 two operations for samples irradiations lasted about 12 hours each at a power of 100 kW. These long lasting operations started in the evening and most of them were carried out at night, when there are only small fluctuations in atmosphere temperature. Therefore the conditions were ideal for evaluating the thermal balance of the power dissipated by the reactor core through the forced cooling system. Heat balance is the standard methodology for power calibration of the IPR-R1 reactor. As in any reactor operation, the main operating parameters were monitored and stored by the Data Acquisition System developed for the reactor. These data have been used for the analysis and calculation of the evolution of several neutronic and thermalhydraulic parameters involved in the reactor operation. This paper analyzes the two long lasting operations of the IPR-R1 TRIGA and compares the recorded results for the power dissipated through the primary cooling loop with the results of the power calibration conducted in March 2009. The results corresponded to those of the thermal power calibration within the uncertainty of this methodology, indicating system stability over a period of six months. (author)

  6. Estimation of fast neutron fluence in steel specimens type Laguna Verde in TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Galicia A, J.; Francois L, J. L.; Aguilar H, F.

    2015-09-01

    The main purpose of this work is to obtain the fluence of fast neutrons recorded within four specimens of carbon steel, similar to the material having the vessels of the BWR reactors of the nuclear power plant of Laguna Verde when subjected to neutron flux in a experimental facility of the TRIGA Mark III reactor, calculating an irradiation time to age the material so accelerated. For the calculation of the neutron flux in the specimens was used the Monte Carlo code MCNP5. In an initial stage, three sheets of natural molybdenum and molybdenum trioxide (MoO 3 ) were incorporated into a model developed of the TRIGA reactor operating at 1 M Wth, to calculate the resulting activity by setting a certain time of irradiation. The results obtained were compared with experimentally measured activities in these same materials to validate the calculated neutron flux in the model used. Subsequently, the fast neutron flux received by the steel specimens to incorporate them in the experimental facility E-16 of the reactor core model operating at nominal maximum power in steady-state was calculated, already from these calculations the irradiation time required was obtained for values of the neutron flux in the range of 10 18 n/cm 2 , which is estimated for the case of Laguna Verde after 32 years of effective operation at maximum power. (Author)

  7. Fundamental approach to TRIGA steady-state thermal-hydraulic CHF analysis

    International Nuclear Information System (INIS)

    Feldman, E.E.

    2008-01-01

    Methods are investigated for predicting the power at which critical heat flux (CHF) occurs in TRIGA reactors that rely on natural convection for primary flow. For a representative TRIGA reactor, two sets of functions are created. For the first set, the General Atomics STAT code and the more widely-used RELAP5-3D code are each employed to obtain reactor flow rate as a function of power. For the second set, the Bernath correlation, the 2006 Groeneveld table, the Hall and Mudawar outlet correlation, and each of the four PG-CHF correlations for rod bundles are used to predict the power at which CHF occurs as a function of channel flow rate. The two sets of functions are combined to yield predictions of the power at which CHF occurs in the reactor. A combination of the RELAP5-3D code and the 2006 Groeneveld table predicts 67% more CHF power than does a combination of the STAT code and the Bernath correlation. (author)

  8. Optimization and analysis of the effects of physical parameters in a TRIGA-ADSR

    Energy Technology Data Exchange (ETDEWEB)

    Hassanzadeh, M. [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of)

    2017-07-15

    In the current research, sensitivity analysis of the accelerator and fissionable and non-fissionable spallation targets containing U-238, Pb, LBE and W materials, Y{sub n/p} (spallation neutron yield), E{sub p} (proton energy), G (energy gain), φ* (importance of neutron source), K{sub eff} (effective multiplication coefficient), K{sub s} (source multiplication coefficient) and I{sub p} (accelerator current) for two cases of K{sub s} including: 0.91 and 0.97 in an Accelerator Driven Subcritical Reactor TRIGA (TRIGA-ADSR) were studied. These neutronic factors were computed by MCNPX code. The obtained results of this study show that for the Y{sub n/p} and G parameters, there is an optimum energy from 800 to 1000 MeV. Furthermore, according to the results, if this reactor would be operated close to criticality, the effect of reactivity insertion on the core power is raised. Also, the optimum value of K{sub s} = 0.97 was chosen as adequate multiplication coefficient in this research because of appropriate margins. Lastly, the results of these investigations show that analysis of sensitivity and specificity of the accelerator and spallation target factors is required to optimize the neutronic plan of ADSR.

  9. 3 MW TRIGA Research Reactor facility of BAEC and its Utilization

    International Nuclear Information System (INIS)

    Molla, N.I.; Bhuiyan, S.I.; Wadud Mondal, M.A.; Ahmed, F.U.; Islam, M.N.; Hossain, S.M.; Ahmed, K.; Zulquarnain, A.; Abedin, Z.

    1999-01-01

    The paper briefly describes the Utilisation of 3 MW TRIGA Research Reactor of BAEC for neutron beam research, neutron activation analysis are isotope production. It includes the installation of the triple axis neutron spectrometer at the radial piercing beam port and a neutron radiography set-up at the tangential beam port and their uses for material analysis and condensed matter research and material testing. Nuclear and magnetic structures of some ferrites have been studied in powder diffraction method in the double axis mode. SANS technique with double crystal diffraction known as Bonse and Hart's method has been adopted in an experiment with alumina sample. The neutron radiography set-up and its use in the detection of corrosion in alumina have been reported. Determination of arsenic concentration in drinking water from tube well via Instrumental Neutron Activation Analysis and production of radioiodine-131 by dry distillation method are presented. Our experience on the removal of N-16 decay tank because of the leakage of coolant and bringing the research reactor back to operational by-passing the decay tank have been focussed. A possible reconfiguration of the existing TRIGA core, without exceeding the safety margins, providing additional irradiation channel and upgrading the neutron flux for increased radioisotope production has been attempted. Cross section library ENDF/B-VI and JENDL3.2, code NJOY94.10, WIMSD package, 3-D code CITATION, PARET and Monte Carlo code MCNP4B2 have been employed to achieve the objective. (author)

  10. Applied research performed and in progress by using a TRIGA nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Moauro, A.; Madaro, M.

    1985-10-01

    The TRIGA reactor of ENEA Casaccia has been used in different applied research fields among which particularly applications of neutron activation analysis (NAA) are reported. The results are referred to the determination of about 30 microelements in marine suspended particulate matter and sediments. A study was also performed on the impurities of the filters used for environmental studies and an intercomparison was carried out on two sediments distributed by the Joint Research Center, Ispra, Euratom. Many applications of NAA have been performed in forensics on request of Italian Courts for determination of gunshot residues, particularly on paraffin gloves and clothes of person suspected to have handled and/or fired a gun; also the firing distance was determined in some cases. The uptake of cobalt and zinc has been studied in pot and the tolerance index was established in cooperation with the Istituto di Cerealicoltura Roma. The rare earth concentrations normalized with respect to the chondritic rocks can be utilized for petrogenetic studies and some determinations have been performed both by thermal and epithermal NAA on some rock samples. A program of utilization of the TRIGA reactor in the field of the preparation of a Ir-191m generator has been provided for applications in the pediatric angiography. For this purpose a cooperation with the Clinical Physiology Institute of National Research Council of Pisa has been established.

  11. Aspects of intellectual property related to the TRIGA reactor in Romania

    International Nuclear Information System (INIS)

    Chirita, Ion

    2008-01-01

    Full text: A TRIGA - type research reactor has been operating in Pitesti since 1979. In Romania, the first research reactor - of the WWR-C type - has been operating since 1957. Both these reactors have contributed to the formation of well - trained specialists, whose works constitute an important intellectual and industrial property. Institute for Nuclear Research (formerly INT, then INPR) is the holder of several published patents, such as: Procedure for decontamination of water and primary circuits of irradiation devices; Reconditioning of ion exchangers; Nozzle for flow water gaugers; Oscillating electromagnetic pump; Facility for determining nuclear fuel burnup; Portable monitor for contamination measurements; Cable joints with biological protection; Anti-seismic and thermal connection; Automatic facility for nuclear fuel irradiation testing; Method for determining power distribution specific for research rector fuel elements; Tight end-fittings; Cooling damage facility, etc. Many of these have been applied or can be applied to reactors of the TRIGA family or are already installed or under installation to research reactors of other types. (authors)

  12. Measurements of neutron flux distributions in the core of the Ljubljana TRIGA Mark II Reactor

    International Nuclear Information System (INIS)

    Rant, J.; Ravnik, M.; Mele, I.; Dimic, V.

    2008-01-01

    Recently the Ljubljana TRIGA Mark II Reactor has been refurbished and upgraded to pulsed operation. To verify the core design calculations using TRIGAP and PULSTR1 codes and to obtain necessary data for future irradiation and neutron beam experiments, an extensive experimental program of neutron flux mapping and neutron field characterization was carried out. Using the existing neutron measuring thimbles complete axial and radial distributions in two radial directions were determined for two different core configurations. For one core configuration the measurements were also carried out in the pulsed mode. For flux distributions thin Cu (relative measurements) and diluted Au wires (absolute values) were used. For each radial position the cadmium ratio was determined in two axial levels. The core configuration was rather uniform, well defined (fresh fuel of a single type, including fuelled followers) and compact (no irradiation channels or gaps), offering unique opportunity to test the computer codes for TRIGA reactor calculations. The neutron flux measuring procedures and techniques are described and the experimental results are presented. The agreement between the predicted and measured power peaking factors are within the error limits of the measurements (<±5%) and calculations (±10%). Power peaking occurs in the B ring, and in the A ring (centre) there is a significant flux depression. (authors)

  13. Gamma spectrometry inspection of TRIGA MARK II fuel using caesium isotopes

    Energy Technology Data Exchange (ETDEWEB)

    Karimzadeh, S., E-mail: sam.karimzadeh@ati.ac.a [Vienna University of Technology, Institute of Atomic and Subatomic Physics (ATI), Stadionallee 2, A-1020 Vienna (Austria); Khan, R.; Boeck, H. [Vienna University of Technology, Institute of Atomic and Subatomic Physics (ATI), Stadionallee 2, A-1020 Vienna (Austria)

    2011-01-15

    Research highlights: Cs isotopes are the best choices for the burn up determination of spent fuel. Gamma spectrometer calibration using MCNP5. Cs-ratio can be applied by relative calibration method. - Abstract: Gamma spectrometry is one of the common methods to inspect the spent fuel from research reactors. This method has been applied to in-pool measurements of the Spent Fuel Elements (SPEs) of the TRIGA Mark II research reactor. Due to mixed nature of the reactor core and complicated irradiation history of the fuel elements (FEs), the gamma spectrometry of the FE establishes improvements in the calculation and measurement of the SPE. In order to inspect the TRIGA SPE from dry storage and cooled fuel from the reactor pool, the selected spend fuels are scanned and measured using the fuel-scanning machine. Gamma spectrometry is performed by HPGe detector for spend fuel inspection and determination of the {sup 137}Cs activity and {sup 134}Cs/{sup 137}Cs ratio. In this work, the steps of the detector calibration and the use of the Monte Carlo radiation transport code (MCNP5) have been described. In addition, the fuel-scanning machine and the gamma spectrometer are modelled by MCNP5 to simulate the gamma transport from fuel to detector. It also simulate the gamma spectrometer calibration for the burn up determination of the spend fuel. The results from MCNP5 simulation are applied to spectroscopic measurements and compared with the theoretical predictions of the neutronics code ORIGEN2 in this research work.

  14. Study on Reactor Performance of Online Power Monitoring in PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Sabri Minhat

    2014-01-01

    The Reactor TRIGA PUSPATI (RTP) at Malaysia Nuclear Agency is a TRIGA Mark II type reactor and pool type cooled by natural circulation of light water. This paper describe on reactor performance of online power monitoring based on various parameter of reactor such as log power, linear power, period, Fuel and coolant temperature and reactivity parameter with using neutronic and other instrumentation system of reactor. Methodology of online power estimation and monitoring is to evaluate and analysis of reactor power which is important of reactor safety and control. Neutronic instrumentation system will use to estimate power measurement, differential of log and linear power and period during reactor operation .This study also focus on noise fluctuation from fission chamber during reactor operation .This work will present result of online power monitoring from RTP which indicated the safety parameter identification and initiate safety action on crossing the threshold set point trip. Conclude that optimization of online power monitoring will improved the reactor control and safety parameter of reactor during operation. (author)

  15. Validation of neutron flux redistribution factors in JSI TRIGA reactor due to control rod movements.

    Science.gov (United States)

    Kaiba, Tanja; Žerovnik, Gašper; Jazbec, Anže; Štancar, Žiga; Barbot, Loïc; Fourmentel, Damien; Snoj, Luka

    2015-10-01

    For efficient utilization of research reactors, such as TRIGA Mark II reactor in Ljubljana, it is important to know neutron flux distribution in the reactor as accurately as possible. The focus of this study is on the neutron flux redistributions due to control rod movements. For analyzing neutron flux redistributions, Monte Carlo calculations of fission rate distributions with the JSI TRIGA reactor model at different control rod configurations have been performed. Sensitivity of the detector response due to control rod movement have been studied. Optimal radial and axial positions of the detector have been determined. Measurements of the axial neutron flux distribution using the CEA manufactured fission chambers have been performed. The experiments at different control rod positions were conducted and compared with the MCNP calculations for a fixed detector axial position. In the future, simultaneous on-line measurements with multiple fission chambers will be performed inside the reactor core for a more accurate on-line power monitoring system. Copyright © 2015 Elsevier Ltd. All rights reserved.

  16. Fuel burnup analysis of the TRIGA Mark II reactor at the University of Pavia

    International Nuclear Information System (INIS)

    Chiesa, Davide; Clemenza, Massimiliano; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica; Alloni, Daniele; Magrotti, Giovanni; Manera, Sergio; Prata, Michele; Salvini, Andrea; Cammi, Antonio; Zanetti, Matteo; Sartori, Alberto

    2016-01-01

    Highlights: • A fuel evolution model for a TRIGA Mark II reactor has been developed. • Reproduction of nearly 50 years of reactor operation. • The model was used to predict the best reactor reconfiguration. • Reactor life was extended without adding fresh fuel elements. - Abstract: A time evolution model was developed to study fuel burnup for the TRIGA Mark II reactor at the University of Pavia. The results were used to predict the effects of a complete core reconfiguration and the accuracy of this prediction was tested experimentally. We used the Monte Carlo code MCNP5 to reproduce system neutronics in different operating conditions and to analyze neutron fluxes in the reactor core. The software that took care of time evolution, completely designed in-house, used the neutron fluxes obtained by MCNP5 to evaluate fuel consumption. This software was developed specifically to keep into account some features that differentiate low power experimental reactors from those used for power production, such as the daily ON/OFF cycle and the long fuel lifetime. These effects can not be neglected to properly account for neutron poison accumulation. We evaluated the effect of 48 years of reactor operation and predicted a possible new configuration for the reactor core: the objective was to remove some of the fuel elements from the core and to obtain a substantial increase in the Core Excess reactivity value. The evaluation of fuel burnup and the reconfiguration results are presented in this paper.

  17. Sensitivity Analysis of the TRIGA IPR-R1 Reactor Models Using the MCNP Code

    Directory of Open Access Journals (Sweden)

    C. A. M. Silva

    2014-01-01

    Full Text Available In the process of verification and validation of code modelling, the sensitivity analysis including systematic variations in code input variables must be used to help identifying the relevant parameters necessary for a determined type of analysis. The aim of this work is to identify how much the code results are affected by two different types of the TRIGA IPR-R1 reactor modelling processes performed using the MCNP (Monte Carlo N-Particle Transport code. The sensitivity analyses included small differences of the core and the rods dimensions and different levels of model detailing. Four models were simulated and neutronic parameters such as effective multiplication factor (keff, reactivity (ρ, and thermal and total neutron flux in central thimble in some different conditions of the reactor operation were analysed. The simulated models presented good agreement between them, as well as in comparison with available experimental data. In this way, the sensitivity analyses demonstrated that simulations of the TRIGA IPR-R1 reactor can be performed using any one of the four investigated MCNP models to obtain the referenced neutronic parameters.

  18. Safety analysis and optimization of the core fuel reloading for the Moroccan TRIGA Mark-II reactor

    International Nuclear Information System (INIS)

    Nacir, B.; Boulaich, Y.; Chakir, E.; El Bardouni, T.; El Bakkari, B.; El Younoussi, C.

    2014-01-01

    Highlights: • Additional fresh fuel elements must be added to the reactor core. • TRIGA reactor could safely operate around 2 MW power with 12% fuel elements. • Thermal–hydraulic parameters were calculated and the safety margins are respected. • The 12% fuel elements will have no influence on the safety of the reactor. - Abstract: The Moroccan TRIGA MARK II reactor core is loaded with 8.5% in weight of uranium standard fuel elements. Additional fresh fuel elements must periodically be added to the core in order to remedy the observed low power and to return to the initial reactivity excess at the End Of Cycle. 12%-uranium fuel elements are available to relatively improve the short fuel lifetime associated with standard TRIGA elements. These elements have the same dimensions as standards elements, but with different uranium weight. The objective in this study is to demonstrate that the Moroccan TRIGA reactor could safely operate, around 2 MW power, with new configurations containing these 12% fuel elements. For this purpose, different safety related thermal–hydraulic parameters have been calculated in order to ensure that the safety margins are largely respected. Therefore, the PARET model for this TRIGA reactor that was previously developed and combined with the MCNP transport code in order to calculate the 3-D temperature distribution in the core and all the most important parameters like the axial distribution of DNBR (Departure from Nucleate Boiling Ratio) across the hottest channel. The most important conclusion is that the 12% fuel elements utilization will have no influence on the safety of the reactor while working around 2 MW power especially for configurations based on insertions in C and D-rings

  19. Design, construction and implementation of two redundant circuits of the actuation logic of the protection system of the new control console of TRIGA Mark III reactor of ININ

    International Nuclear Information System (INIS)

    Celestino M, E.

    2016-01-01

    The Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico has a nuclear reactor type TRIGA Mark III, which was put into operation in 1968. The reactor is used for staff training, radioisotope production, and for research projects of different areas. Over time and due to advances constantly has the electronics industry, maintenance of electronic systems is complicated because basically sometimes components that are no longer manufactured or no longer exist in the market, making it necessary to create projects required modernization. This is the case of the TRIGA reactor of ININ, so the Department of Automation and Instrumentation ININ is undertaking a new project to update the reactor control console. Systems that make up a nuclear reactor protection system (Ps) is relevant, since it is responsible for generating the necessary steps to shut down the reactor to an event of uncertainty which could affect the operators or the installation own actions. As part of the renovation project, this study design is presented to update the Logic of Action (La) of the Ps, whose final design must meet the requirements or specifications set by users and or regulations applicable to nuclear research reactors. One of the requirements established for the proposed new design La, is that it must be implemented with components and devices manufactured with latest technologies, and readily available on the market. The design which is operating currently uses TTL logic whose components are no longer available in the market, so for the new design you decide to use programmable circuits, and specifically, the CPLDs called (by the acronym Complex Programmable Logic Device). These CPLDs are electronic devices that solve complex logic equations and meeting the requirements of functionality and modernity for the new design of the La. In this work the criteria used for the selection of the CPLDs considering the availability and ease of software and hardware to use, and the design and

  20. University of Illinois nuclear pumped laser program. [experiments with a TRIGA pulsed reactor with a broad pulse and a low peak flux

    Science.gov (United States)

    Miley, G. H.

    1979-01-01

    The development of nuclear pumped lasers with improved efficiency, energy storage capability, and UF6 volume pumping is reviewed. Results of nuclear pumped laser experiments using a TRIGA-type pulsed reactor are outlined.

  1. Feasibility study of the university of Utah TRIGA reactor power upgrade - Part I: Neutronics-based study in respect to control rod system requirements and design

    Directory of Open Access Journals (Sweden)

    Ćutić Avdo

    2013-01-01

    Full Text Available We present a summary of extensive studies in determining the highest achievable power level of the current University of Utah TRIGA core configuration in respect to control rod requirements. Although the currently licensed University of Utah TRIGA power of 100 kW provides an excellent setting for a wide range of experiments, we investigate the possibility of increasing the power with the existing fuel elements and core structure. Thus, we have developed numerical models in combination with experimental procedures so as to assess the potential maximum University of Utah TRIGA power with the currently available control rod system and have created feasibility studies for assessing new core configurations that could provide higher core power levels. For the maximum determined power of a new University of Utah TRIGA core arrangement, a new control rod system was proposed.

  2. Understanding users

    DEFF Research Database (Denmark)

    Johannsen, Carl Gustav Viggo

    2014-01-01

    Segmentation of users can help libraries in the process of understanding user similarities and differences. Segmentation can also form the basis for selecting segments of target users and for developing tailored services for specific target segments. Several approaches and techniques have been...... segmentation project using computer-generated clusters. Compared to traditional marketing texts, this article also tries to identify user segments or images or metaphors by the library profession itself....

  3. Design, construction, and demonstration of a neutron beamline and a neutron imaging facility at a Mark-I TRIGA reactor

    Science.gov (United States)

    Craft, Aaron E.

    The fleet of research and training reactors is aging, and no new research reactors are planned in the United States. Thus, there is a need to expand the capabilities of existing reactors to meet users' needs. While many research reactors have beam port facilities, the original design of the United States Geological Survey TRIGA Reactor (GSTR) did not include beam ports. The MInes NEutron Radiography (MINER) facility developed by this thesis and installed at the GSTR provides new capabilities for both researchers and students at the Colorado School of Mines. The facility consists of a number of components, including a neutron beamline and beamstop, an optical table, an experimental enclosure and associated interlocks, a computer control system, a multi-channel plate imaging detector, and the associated electronics. The neutron beam source location, determined through Monte Carlo modeling, provides the best mixture of high neutron flux, high thermal neutron content, and low gamma radiation content. A Monte Carlo n-Particle (MCNP) model of the neutron beam provides researchers with a tool for designing experiments before placing objects in the neutron beam. Experimental multi-foil activation results, compared to calculated multi-foil activation results, verify the model. The MCNP model predicts a neutron beamline flux of 2.2*106 +/- 6.4*105 n/cm2-s based on a source particle rate determined from the foil activation experiments when the reactor is operating at a power of 950 kWt with the beam shutter fully open. The average cadmium ratio of the beamline is 7.4, and the L/D of the neutron beam is approximately 200+/-10. Radiographs of a sensitivity indicator taken using both the digital detector and the transfer foil method provide one demonstration of the radiographic capabilities of the new facility. Calibration fuel pins manufactured using copper and stainless steel surrogate fuel pellets provide additional specimens for demonstration of the new facility and offer a

  4. Relative fission product yield determination in the USGS TRIGA Mark I reactor

    Science.gov (United States)

    Koehl, Michael A.

    Fission product yield data sets are one of the most important and fundamental compilations of basic information in the nuclear industry. This data has a wide range of applications which include nuclear fuel burnup and nonproliferation safeguards. Relative fission yields constitute a major fraction of the reported yield data and reduce the number of required absolute measurements. Radiochemical separations of fission products reduce interferences, facilitate the measurement of low level radionuclides, and are instrumental in the analysis of low-yielding symmetrical fission products. It is especially useful in the measurement of the valley nuclides and those on the extreme wings of the mass yield curve, including lanthanides, where absolute yields have high errors. This overall project was conducted in three stages: characterization of the neutron flux in irradiation positions within the U.S. Geological Survey TRIGA Mark I Reactor (GSTR), determining the mass attenuation coefficients of precipitates used in radiochemical separations, and measuring the relative fission products in the GSTR. Using the Westcott convention, the Westcott flux, modified spectral index, neutron temperature, and gold-based cadmium ratios were determined for various sampling positions in the USGS TRIGA Mark I reactor. The differential neutron energy spectrum measurement was obtained using the computer iterative code SAND-II-SNL. The mass attenuation coefficients for molecular precipitates were determined through experiment and compared to results using the EGS5 Monte Carlo computer code. Difficulties associated with sufficient production of fission product isotopes in research reactors limits the ability to complete a direct, experimental assessment of mass attenuation coefficients for these isotopes. Experimental attenuation coefficients of radioisotopes produced through neutron activation agree well with the EGS5 calculated results. This suggests mass attenuation coefficients of molecular

  5. Experimental study of the temperature distribution in the TRIGA IPR-R1 Brazilian research reactor

    International Nuclear Information System (INIS)

    Mesquita, Amir Zacarias

    2005-01-01

    The TRIGA-IPR-R1 Research Nuclear Reactor has completed 44 years in operation in November 2004. Its initial nominal thermal power was 30 kW. In 1979 its power was increased to 100 kW by adding new fuel elements to the reactor. Recently some more fuel elements were added to the core increasing the power to 250 kW. The TRIGA-IPR-R1 is a pool type reactor with a natural circulation core cooling system. Although the large number of experiments had been carried out with this reactor, mainly on neutron activation analysis, there is not many data on its thermal-hydraulics processes, whether experimental or theoretical. So a number of experiments were carried out with the measurement of the temperature inside the fuel element, in the reactor core and along the reactor pool. During these experiments the reactor was set in many different power levels. These experiments are part of the CDTN/CNEN research program, and have the main objective of commissioning the TRIGA-IPR-R1 reactor for routine operation at 250 kW. This work presents the experimental and theoretical analyses to determine the temperature distribution in the reactor. A methodology for the calibration and monitoring the reactor thermal power was also developed. This methodology allowed adding others power measuring channels to the reactor by using thermal processes. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were also experimentally valued. lt was also presented a correlation for the gap conductance between the fuel and the cladding. The experimental results were compared with theoretical calculations and with data obtained from technical literature. A data acquisition and processing system and a software were developed to help the investigation. This system allows on line monitoring and registration of the main reactor operational parameters. The experiments have given better comprehension of the reactor thermal-fluid dynamics and helped to develop numerical

  6. Activation calculation of steel of the control rods of TRIGA Mark III reactor; Calculo de activacion del acero de las barras de control del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Garcia M, T.; Cruz G, H. S.; Ruiz C, M. A.; Angeles C, A., E-mail: teodoro.garcia@inin.gob.mx [ININ, Carretera Mexico-Toluca sn, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2014-10-15

    In the pool of TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ), there are control rods that were removed from the core, and which are currently on shelves of decay. These rods were part of the reactor core when only had fuel standard (from 1968-1989). To conduct a proper activation analysis of the rods, is very important to have well-characterized the materials which are built, elemental composition of the same ones, the atomic densities and weight fractions of the elements that constitute them. To determine the neutron activation of the control rods MCNP5 code was used, this code allows us to have well characterized the radionuclides inventory that were formed during irradiation of the control rods. This work is limited to determining the activation of the steel that is part of the shielding of the control rods, the nuclear fuel that is in the fuel follower does not include. The calculation model of the code will be validated with experimental measurements and calculating the activity of fission products of the fuel follower which will take place at the end of 2014. (Author)

  7. Immobilization of Ion Exchange radioactive resins of the TRIGA Mark III Nuclear Reactor; Inmovilizacion de resinas de intercambio ionico radiactivas del reactor nuclear TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Martinez, H

    1999-07-01

    In the last decades many countries in the world have taken interest in the use, availability, and final disposal of dangerous wastes in the environment, within these, those dangerous wastes that contain radioactive material. That is why studies have been made on materials used as immobilization agent of radioactive waste that may guarantee its storage for long periods of time under drastic conditions of humidity, temperature change and biodegradation. In mexico, the development of different applications of radioactive material in the industry, medicine and investigation, have generated radioactive waste, sealed and open sources, whose require a special technological development for its management and final disposal. The present work has as a finality to develop the process and define the agglutinating material, bitumen, cement and polyester resin that permits immobilization of resins of Ionic Exchange contaminated by Barium 153, Cesium 137, Europium 152, Cobalt 60 and Manganese 54 generated from the nuclear reactor TRIGA Mark III. Ionic interchange contaminated resin must be immobilized and is analysed under different established tests by the Mexican Official Standard NOM-019-NUCL-1995 {sup L}ow level radioactive wastes package requirements for its near-surface final disposal. Immobilization of ionic interchange contaminated resins must count with the International Standards applicable in this process; in these standards, the following test must be taken in prototype examples: Free-standing water, leachability, compressive strength, biodegradation, radiation stability, thermal stability and burning rate. (Author)

  8. Reliability data collection and processing for Romanian TRIGA-SSR 14MW

    International Nuclear Information System (INIS)

    Mladin, Daniela; Mladin, Mirea; Cristea, Dumitru

    2002-01-01

    The use of site specific reliability data for PSA use is highly recommended because it enhances the accurateness and credibility of the risk analysis. In order to obtain the database statistic items for the reactor components it is necessary to: Develop a brief reactor operation history; Identify the components which can be monitored and their specific failure modes: Clearly define the component boundary; Select and run through the sources of recorded information related to failures and operational details; Process the data; The paper presents how these steps are completed for obtaining failure rates and confidence intervals limits (95% and 5%) for a series of Romanian TRIGA components such as: pumps, motors, valves, compressors, fans, etc. The identification of component boundary and failure rates modes is performed according to the IAEA guides for research reactors database. (author)

  9. Chemical control on the TRIGA IPR-R1 reactor primary cooling system water

    Energy Technology Data Exchange (ETDEWEB)

    Auler, Lucia M.L.A.; Menezes, Maria Angela de B.C.; Oliveira, Paulo Fernando; Franco, Milton Batista; Maretti Junior, Fausto [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)]. E-mails: aulerlm@cdtn.br; menezes@cdtn.br; pfo@cdtn.br; francom@cdtn.br; fmj@cdtn.br

    2007-07-01

    The TRIGA MARK I IPR-R1 reactor located at CDTN/CNEN has been in operation and contributed to research and with services to society since 1960. It has been used in several activities such as nuclear power plant operation, graduate and post-graduate training courses, isotope production, and as an analytical irradiation tool of different types of samples. Among the several reactor structural and operational safety requirements is the chemical quality control of the primary circuit cooling water. This work proposes a water sampling plan and presents the results obtained in a period previous to this plan. Several anions and the presence of metals were determined by Ionic Exchange Chromatography, by Atomic Absorption Spectrophotometer, and by ICP-OES, all techniques available at CDTN/CNEN. The values for pH and conductivity present small deviation. (author)

  10. The characteristic assessment of spent ion exchange resin from PUSPATI TRIGA REACTOR (RTP) for immobilization process

    Energy Technology Data Exchange (ETDEWEB)

    Wahida, Nurul [School of Applied Physics, Universiti Kebangsaan Malaysia, 43600 Bangi, Selangor, Malaysia and Malaysian Nuclear Agency, Bangi 43000 Kajang, Selangor (Malaysia); Yasir, Muhamad Samudi; Majid, Amran Ab; Irwan, M. N. [School of Applied Physics, Universiti Kebangsaan Malaysia, 43600 Bangi, Selangor (Malaysia); Wahab, Mohd Abd; Marzukee, Nik; Paulus, Wilfred; Phillip, Esther; Thanaletchumy [Malaysian Nuclear Agency, Bangi 43000 Kajang, Selangor (Malaysia)

    2014-09-03

    In this paper, spent ion exchange resin generated from PUSPATI TRIGA reactor (RTP) in Malaysian Nuclear Agency were characterized based on the water content, radionuclide content and radionuclide leachability. The result revealed that the water content in the spent resin is 48%. Gamma spectrometry analysis indicated the presence of {sup 134}Cs, {sup 137}Cs, {sup 152}Eu, {sup 54}Mn, {sup 58}Co, {sup 60}Co and {sup 65}Zn. The leachability test shows a small concentrations (<1 Bq/l) of {sup 152}Eu and {sup 134}Cs were leached out from the spent resin while {sup 60}Co activity concentrations slightly exceeded the limit generally used for industrial wastewater i.e. 1 Bq/l. Characterization of spent ion exchange resin sampled from RTP show that this characterization is important as a basis to immobilize this radioactive waste using geopolymer technology.

  11. Principle of human system interface (HSI) design for new reactor console of PUSPATI TRIGA Reactor (RTP)

    International Nuclear Information System (INIS)

    Zareen Khan Abdul Jalil Khan; Ridzuan Abdul Mutalib; Mohd Idris Taib; Mohd Khairulezwan Abdul Manan; Nurfarhana Ayuni Joha; Mohd Sabri Minhat; Izhar Abu Hussin

    2013-01-01

    Full-text: This paper will describe the principle of human system interface design for new reactor console in control room at TRIGA reactor facility. In order to support these human system interface challenges in digital reactor console. Software-based instrumentation and control (I and C) system for new reactor console could lead to new human machine integration. The proposed of Human System Interface (HSI) which included the large display panels which shows reactor status, compact and computer-based workstations for monitoring, control and protection function. The proposed Human System Interface (HIS) has been evaluated using various human factor engineering. It can be concluded that the Human System Interface (HIS) is designed as to address the safety related computer controlled system. (author)

  12. Experimental measurement of the refrigerant temperature of the TRIGA Mark III reactor of the ININ

    International Nuclear Information System (INIS)

    Gallardo S, L.F.; Alonso V, G.

    1991-08-01

    With the object of knowing the axial temperature profile of the refrigerant in the core of the TRIGA Mark III reactor of the ININ, the temperatures of this, at the enter, in the center and the exit of the core were measured, in the positions: west 2, north 2 and south 1. This was made by means of the thermo pars introduction mounted in aluminum guides, connected to a measurer of digital temperature, whose resolution is of ± 0.1 C. The measurements showed a bigger heating of the refrigerant in the superior half of the core, that which suggests that the axial profile of temperature of the reactor is not symmetrical with respect to the center or that those temperature measurements in the center are not correct. (Author)

  13. Design and Implementation of a Fuzzy Controller for a TRIGA Mark III Reactor

    Directory of Open Access Journals (Sweden)

    Tonatiuh Rivero-Gutiérrez

    2012-01-01

    Full Text Available The design and testing of a fuzzy rule based controller to regulate the power of a TRIGA Mark III research nuclear reactor are presented. The design does not require the current exact parameters of the point kinetic equations of the reactor. Instead, from a qualitative analysis of the actions taken by the operators during the reactor’s operation, a set of control rules is derived. The rules cover the operation of the reactor from low levels of about dozens of watts up to its full power level of one megawatt. The controller is able to increase power from different initial values to a wide range of desired levels, maintaining constant levels for long periods of time. The controller’s output is the external reactivity, which is further converted to a control rod incremental movement. The fuzzy controller is implemented on the reactor’s digital operating console, and the results of a series of experiments are discussed.

  14. Ten-year utilization of the Oregon State University TRIGA Reactor (OSTR)

    International Nuclear Information System (INIS)

    Ringle, John C.; Anderson, Terrance V.; Johnson, Arthur G.

    1978-01-01

    The Oregon State University TRIGA Reactor (OSTR) has been used heavily throughout the past ten years to accommodate exclusively university research, teaching, and training efforts. Averages for the past nine years show that the OSTR use time has been as follows: 14% for academic and special training courses; 44% for OSU research projects; 6% for non-OSU research projects; 2% for demonstrations for tours; and 34% for reactor maintenance, calibrations, inspections, etc. The OSTR has operated an average of 25.4 hours per week during this nine-year period. Each year, about 20 academic courses and 30 different research projects use the OSTR. Visitors to the facility average about 1,500 per year. No commercial radiations or services have been performed at the OSTR during this period. Special operator training courses are given at the OSTR at the rate of at least one per year. (author)

  15. TRIGASIM: A computer program to simulate a TRIGA Mark I Reactor

    International Nuclear Information System (INIS)

    Ruby, Lawrence

    1992-01-01

    A Fortran-77 computer program has been written which simulates the operation of a TRIGA Mark I Reactor. The 'operator' has options at 1-second intervals, of raising rods, lowering rods, maintaining rods steady, dropping a rod, or scramming the reactor. Results are printed to the screen, and to 2 output files - a tabular record and a logarithmic plot of the power. The Point Kinetic Equations are programmed with 6 delayed groups, quasi-static power feedback, and forward differencing. A pulsing option is available, with simulation which employs the Fuchs Model. A pulse-tail model has been devised to simulate behavior for a few minutes following a pulse. Both graphic and tabular output are also available for the pulses. (author)

  16. Calculation of neutron fluxes in biological shield of the TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Bozic, M.; Zagar, T.; Ravnik, M.

    2001-01-01

    The complete calculation of neutron fluxes in biological shield and verification with experimental results is presented. Calculated results are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Experimental results used for comparison are available from irradiation experiment with selected type of concrete and other materials in irradiation channel 4 in TRIGA Mark II reactor. These experimental results were used as a benchmark. Homogeneous type of problem (without inserted irradiation channel) and problem with asymmetry (inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. Deviation from material data set up as original parameters is also considered (first of all presence of water in concrete and density of concrete) for type of concrete in biological shield and for selected type of concrete in irradiation channel. BUGLE-96 (47 neutron energy groups) library is used. Excellent agreement between calculated and experimental results for reaction rate is received.(author)

  17. Non-destructive material investigation with thermal neutrons at the TRIGA Mark II reactor in Vienna

    International Nuclear Information System (INIS)

    Bastuerk, M.; Boeck, H.; Zamani, B.; Zawisky, M.; Rauch, H.

    2004-01-01

    Neutron tomography providing 3D information about interior of an object is a very efficient tool to visualize inner defects of the materials, non-destructively. In this study, some applications of neutron tomography in different fields such as geology, aerospace, civil engineering and archaeology were presented. Distribution of minerals in pumice and rock samples, visualization of inner defects within a new developed titan aluminum turbine blade, and distribution of silica gel as an important impregnating agent in construction and restoration of buildings were investigated. The measurements of tomography projections taken in the 0 to 180 o angle were performed with a thermal neutron flux of 10 5 at the TRIGA Mark II research reactor in Vienna, and the common filtered back projection method was used for the 3D image reconstruction. (author)

  18. Piping Flexibility Analysis of the Primary Cooling System of TRIGA 2000 Bandung Reactor due to Earthquake

    Directory of Open Access Journals (Sweden)

    H.P. Rahardjo

    2011-08-01

    Full Text Available Earthquakes in a nuclear installation can overload a piping system which is not flexible enough. These loads can be forces, moments and stresses working on the pipes or equipments. If the load is too large and exceed the allowable limits, the piping and equipment can be damaged and lead to overall system operation failure. The load received by piping systems can be reduced by making adequate piping flexibility, so all the loads can be transmitted homogenously throughout the pipe without load concentration at certain point. In this research the analysis of piping stress has been conducted to determine the size of loads that occured in the piping of primary cooling system of TRIGA 2000 Reactor, Bandung if an earthquake happened in the reactor site. The analysis was performed using Caesar II software-based finite element method. The ASME code B31.1 arranging the design of piping systems for power generating system (Power Piping Code was used as reference analysis method. Modeling of piping systems was based on the cooling piping that has already been installed and the existing data reported in Safety Analysis Reports (SARs of TRIGA 2000 reactor, Bandung. The quake considered in this analysis is the earthquake that occurred due to the Lembang fault, since it has the Peak Ground Acceleration (PGA in the Bandung TRIGA 2000 reactor site. The analysis results showed that in the static condition for sustain and expansion loads, the stress fraction in all piping lines does not exceed the allowable limit. However, during operation moment, in dynamic condition, the primary cooling system is less flexible at sustain load, ekspansi load, and combination load and the stress fraction have reached 95,5%. Therefore a pipeline modification (rerouting is needed to make pipe stress does not exceed the allowable stress. The pipeline modification was carried out by applied a gap of 3 mm in the X direction of the support at node 25 and eliminate the support at the node

  19. Conceptual design of experimental LFR fuel element for testing in TRIGA reactor, ACPR zone

    International Nuclear Information System (INIS)

    Nastase, D.; Olteanu, G.; Ioan, M.; Pauna, E.

    2013-01-01

    In the pulsed area of the TRIGA reactor (ACPR zone), the irradiation tests called ''rapid insertions of reactivity on different types of nuclear fuel elements'' are usually realized. During these tests, in the fuel element high powers for a relatively short period of time (about few milliseconds) are generated. The generated heat in fuel pellets raise their central temperature to values over 100 deg C. The conceptual design of an experimental fuel element proposed to be developed and presented in this paper must fulfill a couple of requirements, as follows: to ensure full compatibility with irradiation device sample holder (compatibility is achieved through reduced length of the fuel stack pellets - this way assures a flow flattening on the entire length of the fuel element); to be compatible with the project of irradiated fuel bundle in Lead cooled Fast Reactors (LFR). (authors)

  20. Personnel, Area And Environmental Surveillance Practices At The PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Ligam, A.S.; Zarina Masood; Mohammad Suhaimi Kassim; Ismail Sulaiman; Mohd Fazli Zakaria; Ahmad Nabil Abdul Rahim

    2013-01-01

    Personnel, area and environmental surveillance at research reactors are important to ensure that the worker, public and environment do not receive radiation doses exceeding the prescribed national limits. A surveillance programme has been in place ever since the PUSPATI TRIGA Reactor (RTP) first started its operation in 1982. The results of the surveillance have to be reported to the national regulatory body as well as the International Atomic Energy Agency. This paper will discuss the findings and improvements of the surveillance programme over the past few years. It can be concluded that workers, public and the environment does not receive doses in excess of the prescribed limits due to the operation of the RTP.(author)

  1. Seven years of operation of the U. S. geological survey TRIGA reactor

    International Nuclear Information System (INIS)

    Kraker, Pat

    1976-01-01

    February 1976 marks 7 years of operation of the U. S. Geological Survey TRIGA Reactor (GSTR) facility. In these 7 years we have generated more than 5800 MWH's of thermal energy and irradiated more than 47,000 samples for experimenters from the Survey, universities, and other Governmental agencies. Several mechanical and electrical components have required attention. Changes to the technical specifications have included one minor wording change involving the evacuation alarm, a reevaluation of the measurement of argon-41 concentrations, a revision concerning transient-rod maintenance, and a reduction in the frequency of fuel-element measurements. To improve physical security we have increased building security, installed an intrusion alarm, and, most recently, expanded the boundaries of the facility within the building to provide better control access. There also have been major changes to our operating procedures and the initiation of a reactor-operator requalification program. (author)

  2. Microfungal Activity Test Of The Triga Mark II Reactor Tank Isolation to Aluminium corrosion

    International Nuclear Information System (INIS)

    Wahid, Rosmiarty A.; Umar, Lukman; Yestiani, Yani

    2000-01-01

    In our pres ious study some pure species of micro fungal have been isolated from cooling water and samples taken from surrounding tank wall of TRIGA Mark II Reaktor. This study was conducted to determine their activities to aluminium 6061-T using modified method of Hortative (1962) in the speed of corrosion transmission process. Each isolate was inoculated into mineral nutrient solution. Changes of ph and reduced weight of Aluminium specimen weight between the experimental and the control groups, and the amounts were proportional to to the length of investigation times. The highest degree of the corrosion speed is given by Penicillium simplicissimum inoculant 2,95.10 - 6, followed by Paecilomyceus carneus 2,61.10 6 , Penicillium canescens 2,59.10 - 6 and the control 2,11.10 6 respectively

  3. Improvements to the TRIGA Mark II instrumentation and the direct digital control by microprocessors

    International Nuclear Information System (INIS)

    Tomsic, M.; Taves, R.; Mrcun, I.; Kavsek, D.; Znidaric, B.

    1978-01-01

    Two tendencies have been present in the maintenance of the TRIGA instrumentation: one was to renew only those parts that were deteriorating with age, thus ensuring the continuation of the satisfactory service in the original scope; the other was aimed at adding new features and possibly at changing the whole concept of the reactor control and instrumentation. Though the activities along both lines were not best coordinated at all times, the presently emerging result may be highly satisfactory. Besides the well maintained instrumentation in the original scope and concept, a digital data logging and control system is being installed, based on microprocessors, which should offer new level of flexibility and convenience to the operators and experimenters, without compromising either safety of reliability of the overall instrumentation and control system

  4. Experimental research in neutron physic and thermal-hydraulic at the CDTN Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Souza, Rose Mary G.P.; Ferreira, Andrea V.; Pinto, Antonio J.; Costa, Antonio C.L.; Rezende, Hugo C., E-mail: amir@cdtn.b, E-mail: souzarm@cdtn.b, E-mail: avf@cdtn.b, E-mail: ajp@cdtn.b, E-mail: aclc@cdtn.b, E-mail: hcr@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The IPR-R1 TRIGA (Training, Research, Isotopes production, General Atomics) at Nuclear Technology Development Center (CDTN) is a pool type reactor cooled by natural circulation of light water and an open surface. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world and characterized by inherent safety. The IPR-R1 is the only Brazilian nuclear research reactor available and able to perform experiments in which interaction between neutronic and thermal-hydraulic areas occurs. The IPR-R1 has started up on November 11th, 1960. At that time the maximum thermal power was 30 kW. The present forced cooling system was built in the 70th and the power was upgraded to 100 kW. Recently the core configuration and instrumentation was upgraded again to 250 kW at steady state, and is awaiting the license of CNEN to operate definitely at this new power. This paper describes the experimental research project carried out in the IPR-R1 reactor that has as objective evaluate the behaviour of the reactor operational parameters, and mainly to investigate the influence of temperature on the neutronic variables. The research was supported by Research Support Foundation of the State of Minas Gerais (FAPEMIG) and Brazilian Council for Scientific and Technological Development (CNPq). The research project meets the recommendations of the IAEA, for safety, modernization and development of strategic plan for research reactors utilization. This work is in line with the strategic objectives of Brazil, which aims to design and construct the Brazilian Multipurpose research Reactor (RMB). (author)

  5. Moroccan TRIGA nuclear reactor, an important tool for the development of research, education and training

    International Nuclear Information System (INIS)

    Caoui, A.

    2011-01-01

    Full text: The construction of the Nuclear Research Center of Maamora (NRCM) will enable to the National Center for Nuclear Energy, Sciences and Techniques (CNESTEN) to fulfill its missions for promotion of nuclear techniques in socioeconomic fields, act as technical support for the authorities, and contribute to the introduction of nuclear power for electricity generation considered in the new energy strategy as alternative option for the period 2020-2030. The CNESTEN has commisioned its nuclear research reactor Triga Mark II of 2000 KW on 2007 for wich the operating authorization was delivered on 2009. This research reactor is the keystone structure of the NRCM, its existing and planed utilization include: production of radioisotopes for medical use, neutron activation analysis, non-destructive examination techniques, neutron scattering, reactor physics research and training. In term of human ressources development, CNESTEN is more focusing on education and training for wich an international training Center is under development. The TRIGA research reactor will be an important component of this center. In order to promote the utilization of the reserch reactor in socio-economical sectors at national level, CNESTEN organizea meetings, schools and conferences around each of the reactor applications, and offers the opportunity to researchers, students, socio-economic operators to know more about reactor utilization within scientific visits, courses and training programs. At the international level, CNESTEN strengthens its international partenership. The regional and international cooperation with IAEA, AFRA and bilateral parteners (USA, France), constitutes the platform for capacity building in different areas of CNESTEN RIGA research reactor utilization

  6. A digital data acquisition and display system for ITU TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Can, B.; Omuz, S.

    2008-01-01

    Full text: In this study, a digital data acquisition and display system realized for ITU TRIGA Mark-II Reactor is described. This system is realized in order to help the reactor operator and to increase reactor console capacity. The system consists of two main units, which are host computers and RTI-815F, analog devices, data acquisition card. RTI-815F is multi-function analog/digital input/output board that plugs into one of the available long expansion slots in the IBM-PC, PC/XT, PC/AT, or equivalent personal computers. It has 16 analog input channels for single-ended input signals or 8 analog input channels for differential input signals. But its channel capacity can be increased to 32 input channels for single-ended input signals or 16 input channels for differential input signals. RTI-815F board contains 2 analog output channels, 8 digital input channels and 8 digital output channels. In the ITD TRIGA Mark-II Reactor, 6 fuel temperature channels, 3 water temperature channels, 3 control rod position channels and 4 power channels are chosen as analog input signals for RTI-815F. Its digital outputs are assigned to cooling tower fan, primary and secondary pump reactor scram, control rod rundown. During operation, data are automatically archived to disk and displayed on screen. The channel selection time and sampling time can be adjusted. The simulated movement and position of control rods in the reactor core can be noted and displayed. The changes of power, fuel temperature and water temperature can be displayed on the screen as a graphic. In this system both period and reactivity are calculated and displayed on the screen. (authors)

  7. High Temperature Fuel Cladding Chemical Interactions Between TRIGA Fuels and 304 Stainless Steel

    Energy Technology Data Exchange (ETDEWEB)

    Perez, Emmanuel [Idaho National Lab. (INL), Idaho Falls, ID (United States); Keiser, Jr., Dennis D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Forsmann, Bryan [Boise State Univ., ID (United States); Janney, Dawn E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Henley, Jody [Idaho National Lab. (INL), Idaho Falls, ID (United States); Woolstenhulme, Eric C. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-02-01

    High-temperature fuel-cladding chemical interactions (FCCI) between TRIGA (Training, Research, Isotopes, General Atomics) fuel elements and the 304 stainless steel (304SS) are of interest to develop an understanding of the fuel behavior during transient reactor scenarios. TRIGA fuels are composed of uranium (U) particles dispersed in a zirconium-hydride (Zr-H) matrix. In reactor, the fuel is encased in 304-stainless-steel (304SS) or Incoloy 800 clad tubes. At high temperatures, the fuel can readily interact with the cladding, resulting in FCCI. A number of FCCI can take place in this system. Interactions can be expected between the cladding and the Zr-H matrix, and/or between the cladding and the U-particles. Other interactions may be expected between the Zr-H matrix and the U-particles. Furthermore, the fuel contains erbium-oxide (Er-O) additions. Interactions can also be expected between the Er-O, the cladding, the Zr-H and the U-particles. The overall result is that very complex interactions may take place as a result of fuel and cladding exposures to high temperatures. This report discusses the characterization of the baseline fuel microstructure in the as-received state (prior to exposure to high temperature), characterization of the fuel after annealing at 950C for 24 hours and the results from diffusion couple experiments carries out at 1000C for 5 and 24 hours. Characterization was carried out via scanning electron microscopy (SEM) and transmission electron microscopy (TEM) with sample preparation via focused ion beam in situ-liftout-technique.

  8. Archaeometric studies by using neutron radiography in ITU TRIGA Mark-II reactor

    International Nuclear Information System (INIS)

    Tugrul, A. Beril

    2008-01-01

    Archaeometric many studies have been done by using neutron radiography in ITU TRIGA Mark-II Training and Research Reactor for over 15 years. Tangential beam tube has been arranged for using neutron radiography. Generally, transfer technique has been preferred with using dysprosium screen, but indium screen also is used. Some studies are described which are all on the Anatolian artefacts. The first study from 13th century AD deals with Seljukian period from south-east Anatolia. It investigated a plate from Great Mosque door in Cizre. With means of the neutron radiography painting traces are investigated on the plates. Organic dye traces are noticed on some of plates, which have generally animal figures. Other studies from Urartu period at the first millennium B.C, investigates artefacts found at the vicinity of Van on east Anatolia. An important one is a sword that was found in a grave. It has some corrosion defects. The neutron radiography was applied and shown that wooden parts are there. Other studies referred to samples from the Ikiztepe Excavation site on north Anatolia. Many artefacts were examined by neutron radiography. Some of them evidenced animal parts are recognised as covering parts. An interesting result was obtained to a sword and its sheath that were corroded together. After the neutron radiography applications, it was noticed that there are a cloth between the sword and its sheath. Hence, it was the cause of corrosion of the artefact. By using neutron radiography, many interesting and detailed results were observed by means of the neutron beam from the ITU TRIGA Mark-II Training and Research Reactor. Some of them could not be evidenced by means of any other technique

  9. Applicable regulations and development of surveillance experiments of criticality approach in the TRIGA III Mark reactor; Normativa aplicable y desarrollo de experimentos de vigilancia de aproximacion a criticidad en el reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J.L.; Aguilar H, F.; Rivero G, T.; Sainz M, E. [Instituto nacional de Investigaciones Nucleares, Departamento de Automatizacion, A.P. 18-1027, Col. Escandon, 11801 Mexico D.F. (Mexico)

    2000-07-01

    In the procedure elaborated to repair the vessel of TRIGA III Mark reactor is required to move toward two tanks of temporal storage the fuel elements which are in operation and the spent fuel elements which are in decay inside the reactor pool. The National Commission of Nuclear Safety and Safeguards (CNSNS) has requested as protection measure that it is carried out a surveillance of the criticality approach of the temporal storages. This work determines the main regulation aspects that entails an experiment of criticality approach, moreover, informing about the results obtained in the developing of this experiments. The regulation aspects are not exclusives for this work in the TRIGA Mark III reactor but they also apply toward any assembling of fissile material. (Author)

  10. User 2020

    DEFF Research Database (Denmark)

    Porras, Jari; Heikkinen, Kari; Kinnula, Marianne

    2014-01-01

    determined by the era in which they were born. This is due to the fact that digital natives, born in an already “fully” digitalized world with a plethora of ICT services, have a much closer relationship to these solutions than generations before them. This has also shaped the users perspectives and had......The User 2020 vision is of the changing needs and habits of a user in the future digital world. In order to understand the needs of the future users, we need to look at how users and technology have changed during recent years. The different generations of users are products of their own time...... this evolutionary process. The basis of this Outlook lies in studies of user generations. Although it’s controversial to do so, users have been divided into generations based on their ability and willingness to use ICT solutions. Whether the users are digital ‘tourists’, ‘immigrants’ or ‘natives’ is mainly...

  11. Preparation for shipment of spent TRIGA fuel elements from the research reactor of the Medical University of Hannover

    International Nuclear Information System (INIS)

    Hampel, Gabriele; Cordes, Harro; Ebbinghaus, Kurt; Haferkamp, Dirk

    1998-01-01

    In the early seventies a research reactor of type TRIGA Mark I was installed in the Department of Nuclear Medicine at the Medical University of Hannover (MHH) for the production of isotopes with short decay times for medical use. Since new production methods have been developed, the reactor has become obsolete and the MHH decided to decommission it. Probably in the second quarter of 1999 all 76 spent TRIGA fuel elements will be shipped to Idaho National Engineering and Environmental Laboratory (INEEL), USA, in one cask of type GNS 16. Due to technical reasons within the MHH a special Mobile Transfer System, which is being developed by the company Noell-KRC, will be used for reloading the fuel elements and transferring them from the reactor to the cask GNS 16. A description of the main components of this system as well as the process for transferring the fuel elements follows. (author)

  12. Preparation and planning for the replacement of the Oregon State University TRIGA reactor rotary specimen rack assembly

    International Nuclear Information System (INIS)

    Anderson, T.V.; Dodd, B.; Johnson, A.G.; Carpenter, W.T.

    1984-01-01

    Recently there have been a number of indications that the rotating rack may be approaching the end of its useful life. In order to benefit from the experience of other reactors who have removed and replaced their rotating racks, General Atomic (GA) was contacted and previous TRIGA Conference proceedings were scanned. It was determined that a number of facilities, had experienced difficulties with their lazy susans and eventually had to replace them. However, most of the written descriptions of this project were not sufficiently detailed to be of great use. The purpose of this paper is to identify some of the more important questions related to the replacement of our rotating rack assembly and OSU's currently proposed solutions, with a view towards soliciting ideas from other members of the TRIGA reactor community

  13. Neutronic parameters characterization of the TRIGA IPR-R1 using scale 6.0 (KENO VI)

    Energy Technology Data Exchange (ETDEWEB)

    Faria, Victor; Miro, Rafael; Verdu, Gumersindo; Barrachina, Teresa [Institute for Industrial, Radiophysical and Environmental Safety (ISIRYM), Universitat Politecnica de València (Spain); Silva, Clarysson A. Mello da; Pereira, Claubia [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Dalle, Hugo Moura [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    KENO-VI is a Monte Carlo based transport code used to obtain the criticality of a nuclear system. A model built using this code in the SCALE6.0 software system was developed for the characterization of neutronic parameters of the IPR-R1 TRIGA research reactor. A comparison with experimental values and those calculated with a MCNP code model could be then attained with the purpose to validate this methodology. (author)

  14. Experimental determination of the total isothermal reactivity feedback coefficient for the University of Arizona TRIGA research reactor

    International Nuclear Information System (INIS)

    Spriggs, Gregory D.; Nelson, George W.

    1976-01-01

    An experiment was performed to measure the total isothermal (or bath) feedback coefficient of reactivity for the University of Arizona TRIGA Research Reactor (UARR). It was found that the bath coefficient was temperature-dependent and may be represented by the expression α iso .2634 x 10 -2 + .3428 x 10 -3 T - 2.471 x 10 -5 T 2 + 3.476 x 10 -7 T 3 for the temperature range of 7 C to 43 C. (author)

  15. Collinear laser spectroscopy of calcium and tin at TRIGA-LASER and ISOLDE; Kollineare Laserspektroskopie an Calcium und Zinn an TRIGA-LASER und ISOLDE

    Energy Technology Data Exchange (ETDEWEB)

    Gorges, Christian

    2017-05-10

    From the optical spectra of the ions and atoms we can extract the spin, the change in mean square charge radius, the magnetic dipole moments, and the electric quadrupole moments. To investigate these properties, collinear laser spectroscopy is a particular appropriate method because it is universally applicable, very efficient and allows to investigate short-lived particles with lifetimes in the range of milliseconds and production rates of a few hundred particles per second. Within the scope of this work, a control system (TRITON) was developed for the collinear laser spectroscopy experiment LASPEC at FAIR, which allows distributed access to the various control elements and is therefore very flexible. It was developed and tested on the LASPEC prototype, the TRIGA-LASER experiment at the research reactor TRIGA Mainz. Using TRITON, the isotope shifts of the stable calcium isotopes {sup 40,42,44,48}Ca have been measured in the 4s{sub 1/2}→4p{sub 3/2} transition with an accuracy, which exceeded the previous literature values by about one order of magnitude. These data were the basis for the precise determination of the nuclear charge radii of the calcium isotopes {sup 49-52}Ca from collinear laser spectroscopy installation COLLAPS at ISOLDE/CERN. Furthermore, developments for spectroscopy on {sup 53,54}Ca at COLLAPS have been carried out, where TRITON was also used. The production rates for {sup 54}Ca are about one ion per second and optical detection is no longer possible. Instead, the technique of optical pumping with state selective charge exchange and β-detection of single ions was implemented at COLLAPS and successfully demonstrated at {sup 52}Ca. With these applications, the performance and operational readiness of the LASPEC control system for FAIR was successfully demonstrated. In the second part of the work, spectroscopy on the tin isotopes {sup 109,112-134}Sn has been performed at the COLLAPS experiment. Several electromagnetic moments close to the N=82

  16. Modernization of the facilities of the TRIGA Mark III reactor of ININ; Modernizacion de las instalaciones del reactor TRIGA Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Mendez T, D.; Flores C, J., E-mail: dario.mendez@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2016-09-15

    The TRIGA Mark III reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) has been in operation since 1968 under strict maintenance and component replacement programs, which has allowed its safe operation during this time. Under this scheme, the reactor was operating under suitable conditions, taking into account the different requests for operation that were received for the samples irradiation for the radioisotopes production such as the Sm-153, personnel training, basic research, archaeology and environmental studies and nuclear chemistry of the elements. However, a modernization program of its components and laboratories was required, in order to improve safety in the operation of the same and to increase its use in the analysis of samples by neutron activation and in the training of personnel. This program known as Modernization Program of the Reactor Facilities, was proposed alongside the project to replace high-enrichment fuels with low-enrichment fuels at the end of 2011 and early 2012. The central aspects of this program are described in this work, grouped into generic topics that include instrumentation and control, the radiological monitoring system of the area, the cooling system, the ventilation system, the neutron activation analysis laboratory, the manufacture of graphite elements, inspection submersible system of the pool, temporary storage system for irradiated fuels, traveling crane, Reactor support laboratories and technical meetings, courses and seminars for reactor personnel and associated groups. It also describes some of the most relevant components required for each system and the progress that is made in each one of them. As a fundamental result of the implementation of this Modernization Program of the Reactor Facilities, there has been a substantial improvement in the performance of the systems and components of its facilities, in the reliability of its operation and in the safety of the same. (Author)

  17. Three-dimensional modeling and virtual TRIGA reconfigure for specialized training; Modelado 3D y TRIGA virtual reconfigurable para entrenamiento especializado

    Energy Technology Data Exchange (ETDEWEB)

    Plata M, A. C.; Morales S, J. B.; Flores, M. [Facultad de Ingenieria, Division de Estudios de Posgrado, Campus Morelos, UNAM, Paseo Cuauhnahuac 8532, Col. Progreso, 62550 Jiutepec, Morelos (Mexico)], e-mail: yoyuclof@hotmail.com

    2009-10-15

    The news products that have been realized for the training virtual room which is developing in the Engineering Faculty of National Autonomous University of Mexico are presented. These improvements are mainly in modeling of virtual reality of the reactor building, as well as internal parts of reactor. It was modified the dynamic modeling of control rods of reaction in chain and included new elements to reactor. which exist not necessarily in all the TRIGA, but that, for educational purposes are highly useful. Such is the case of addition of valves, pumps, tanks, injection lines of light or borated water, as well as a heat exchanger, with it can recycle only pool water from side to other, or to extract energy toward a secondary controller from the operator console. The models of heat decay were included, of subcooled and nucleated boiling of coolant-moderator in the core, the dynamics of xenon and samarium. These last with independent multipliers of simulation time to allow variations very fast that real time. All these additions modify the coolant-moderator characteristics and consequently the answer of simulator. The controls are separated in: an operator console (student) very similar to the real systems, another of instructor that has additional access to parameters not directly measurement in the facilities but that allow to modify the system to illustrate another not easily possible effects in the real system. The traveling crane is also modeled and is controlled in a third console from where can to replacement to reactor as well as to add or to replacement: intakes and discharges of coolant circulators, measuring instruments, reflectors and neutron sources. The dynamic models have been tested in SCILAB and SCICOS. At present is working in the integration of the dynamic simulator and the virtual reality mainly with the design requirement of allowing functions of increased reality. (Author)

  18. Adaptive fuzzy control of neutron power of the TRIGA Mark III reactor; Control difuso adaptable de la potencia neutronica del reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Rojas R, E.

    2014-07-01

    The design and implementation of an identification and control scheme of the TRIGA Mark III research nuclear reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico is presented in this thesis work. The identification of the reactor dynamics is carried out using fuzzy logic based systems, in which a learning process permits the adjustment of the membership function parameters by means of techniques based on neural networks and bio-inspired algorithms. The resulting identification system is a useful tool that allows the emulation of the reactor power behavior when different types of insertions of reactivity are applied into the core. The identification of the power can also be used for the tuning of the parameters of a control system. On the other hand, the regulation of the reactor power is carried out by means of an adaptive and stable fuzzy control scheme. The control law is derived using the input-output linearization technique, which permits the introduction of a desired power profile for the plant to follow asymptotically. This characteristic is suitable for managing the ascent of power from an initial level n{sub o} up to a predetermined final level n{sub f}. During the increase of power, a constraint related to the rate of change in power is considered by the control scheme, thus minimizing the occurrence of a safety reactor shutdown due to a low reactor period value. Furthermore, the theory of stability in the sense of Lyapunov is used to obtain a supervisory control law which maintains the power error within a tolerance region, thus guaranteeing the stability of the power of the closed loop system. (Author)

  19. Thermal hydraulic analysis of the IPR-R1 TRIGA research reactor using a RELAP5 model

    International Nuclear Information System (INIS)

    Costa, Antonella L.; Reis, Patricia Amelia L.; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Mesquita, Amir Z.; Soares, Humberto V.

    2010-01-01

    The RELAP5 code is widely used for thermal hydraulic studies of commercial nuclear power plants. Current investigations and code adaptations have demonstrated that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research reactors with good predictions. Therefore, as a contribution to the assessment of RELAP5/MOD3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed using a RELAP5 model to simulate the IPR-R1 TRIGA research reactor at 50 kilowatts (kW) of power operation. The reactor is located in the Nuclear Technology Development Center (CDTN), Brazil. It is a 250 kW, light water moderated and cooled, graphite-reflected, open pool type research reactor. The development and the assessment of a RELAP5 model for the IPR-R1 TRIGA are presented. Experimental data were considered in the process of the RELAP5 model validation. The RELAP5 results were also compared with calculated data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual steady-state reactor behavior in good agreement with the available data.

  20. Man Machine Interface Analysis For The Equipment And Instrumentation In MCR Of The TRIGA Mark II Modified Reactor

    International Nuclear Information System (INIS)

    Itjeu K; Piping S; Suharyo W; Kussigit S; Darlis; Bambang S; Sasongko Y

    2000-01-01

    Procedure and function of installation devices have been reviewed for the equipment and instrumentation in Main Control Room of TRIGA MARK II Modified Reactor. Analysis reviews carried out from the operation data/performances of the equipment and instrumentation of TRIGA II modified reactor and then evaluated. Instrumentation system and reactor control of TRIGA II is the reactor control system controlled by microprocessor installed at TM-II reactor to change analog system. These system consists of 4 parts that in such are control system console (CSC), data acquisition control unit (DAC), power monitor channel and temperature monitor. Beside that the system provided by supporting devices, for instances screen monitor, relay panel, analog and recorder includes analog input terminals and relay. Based on analysis reviewed from the procedures and function of installation devices at the obtaining system have been found that the display equipment enable to analyze reactor safety, enable to show operation condition directly and to perform information target to the each functions. Control equipment are consistent to the role of colour codes, shape and size, and also control classification fit to the reactor safety requirements priority. Alarm devices enable to transfer information promptly from the abnormal accidents and includes its anticipation. The early warning system (alarm, screen features) will show texts warning or warning signs if the reactor system is in the abnormal condition and then the operators will do proper needed immediately

  1. Design, construction and implementation of two redundant circuits of the actuation logic of the protection system of the new control console of TRIGA Mark III reactor of ININ; Diseno, construccion e implementacion de dos circuitos redundantes de la logica de actuacion del sistema de proteccion de la nueva consola de control del reactor Triga Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Celestino M, E.

    2016-07-01

    The Instituto Nacional de Investigaciones Nucleares (ININ) in Mexico has a nuclear reactor type TRIGA Mark III, which was put into operation in 1968. The reactor is used for staff training, radioisotope production, and for research projects of different areas. Over time and due to advances constantly has the electronics industry, maintenance of electronic systems is complicated because basically sometimes components that are no longer manufactured or no longer exist in the market, making it necessary to create projects required modernization. This is the case of the TRIGA reactor of ININ, so the Department of Automation and Instrumentation ININ is undertaking a new project to update the reactor control console. Systems that make up a nuclear reactor protection system (Ps) is relevant, since it is responsible for generating the necessary steps to shut down the reactor to an event of uncertainty which could affect the operators or the installation own actions. As part of the renovation project, this study design is presented to update the Logic of Action (La) of the Ps, whose final design must meet the requirements or specifications set by users and or regulations applicable to nuclear research reactors. One of the requirements established for the proposed new design La, is that it must be implemented with components and devices manufactured with latest technologies, and readily available on the market. The design which is operating currently uses TTL logic whose components are no longer available in the market, so for the new design you decide to use programmable circuits, and specifically, the CPLDs called (by the acronym Complex Programmable Logic Device). These CPLDs are electronic devices that solve complex logic equations and meeting the requirements of functionality and modernity for the new design of the La. In this work the criteria used for the selection of the CPLDs considering the availability and ease of software and hardware to use, and the design and

  2. The study of time-dependent neutronics parameters of the 2MW TRIGA Mark II Moroccan research reactor using BUCAL1 computer code

    International Nuclear Information System (INIS)

    Bakkari, B. El; Nacir, B.; El Younoussi, C.; Boulaich, Y.; Riyach, I.; Otmani, S.; Marcih, I.; Elbadri, H.; El Bardouni, T; Merroun, O.; Boukhal, H.; Zoubair, M.; Htet, A.; Chakir, M.

    2010-01-01

    The 2-MW TRIGA MARK II research reactor at Centre National de l'Energie, des Sciences et des Techniques Nucleaires (CNESTEN) achieved initial criticality on May 2, 2007 with 71 fuel elements. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower and training and production of radioisotopes for their use in agriculture, industry and medicine. This work aims to study the time-dependent neutronics parameters of the TRIGA reactor for elaborating and planning of an in-core fuel management strategy to maximize the utilization of the TRIGA fluxes, using a new elaborated burnup computer code called 'BUCAL1'. The code can be used to aid in analysis, prediction, and optimization of fuel burnup performance in a nuclear reactor. It was developed to incorporate the neutron absorption tally/reaction information generated directly by MCNP5 code in the calculation of fissioned or neutron-transmuted isotopes for multi-fueled regions. The use of Monte Carlo method and punctual cross section data characterizing the MCNP code allows an accurate simulation of neutron life cycle in the reactor, and the integration of data on the entire energy spectrum, thus a more accurate estimation of results than deterministic code can do. Also, for the purpose of this study, a full-model of the TRIGA reactor was developed using the MCNP5 code. The validation of the MCNP model of the TRIGA reactor was made by benchmarking the reactivity experiments. (author)

  3. Performance evaluation of two CANDU fuel elements tested in the TRIGA reactor

    International Nuclear Information System (INIS)

    Nita, Raluca Florentina; Uta, Octavian; Parvan, Marcel; Mincu, Marin

    2010-01-01

    Nuclear Research Institute at Pitesti has a set of facilities, which allow the testing, manipulation and examination of nuclear fuel and structure materials irradiated in CANDU reactors from Cernavoda NPP. These facilities consist of TRIGA materials testing reactor and Post-Irradiation Examination Laboratory (LEPI). The purpose of this work is to describe the post-irradiation examination, of two experimental CANDU fuel elements (EC1 and EC2). The fuel elements were mounted into a pattern port, one in extension of the other in a measuring test for the central temperature evolution. The results of post-irradiation examination are obtained from: Visual inspection and photography of the outer appearance of sheath; Profilometry (diameter, bending, ovalization) and length measuring; Determination of axial and radial distribution of the fission products activity by gamma scanning; Measurement of pressure, volume and isotopic composition of fission gas; Microstructural characterization by metallographic and ceramographic analyzes; Isotopic composition and burn-up determination. The post-irradiation examination results are used, on one hand, to confirm the security, reliability and performance of the irradiated fuel, and on the other hand, for further development of CANDU fuel. (authors)

  4. Research in the fields of radiochemistry and activation analysis using the LENA TRIGA nuclear plant

    International Nuclear Information System (INIS)

    Maxia, V.; Meloni, S.; Stella, R.; Brandone, A.

    1972-01-01

    In the past two years most of the research effort at the Radiochemistry Laboratory and National Research Council Centre for Radiochemistry and Activation Analysis has been devoted to research and development of activation analysis using the TRIGA Mark II LENA reactor of the University of Pavia. Pile neutrons have been used both in steady state and pulse mode, in the determination of oxygen in non ferrous materials. Neutron activation has been applied to the determination of some atmospheric particulate pollutants. Another field, in which activation analysis has been applied, is forensic research. Methods have been developed for the determination of antimony, barium and copper in gunpowder residues. By using inorganic materials such as molybdenum dibromide, zinc ferrocyanide and cadmium metal in granular form it was possible to set up simple chemical procedures in the activation analysis of trace amounts of noble metals in metallic matrices (high purity nickel and copper), geochemical materials (rocks and meteorites) and biological materials (orchard leaves). Neutron activation analysis was also used to investigate on the extraction of the platinum group metals from iodide and thiocyanate solutions at low concentrations

  5. Testing the applicability of the k 0-NAA method at the MINT's TRIGA MARK II reactor

    International Nuclear Information System (INIS)

    Siong, Wee Boon; Dung, Ho Manh; Wood, Ab. Khalik; Salim, Nazaratul Ashifa Abd.; Elias, Md. Suhaimi

    2006-01-01

    The Analytical Chemistry Laboratory at MINT is using the NAA technique since 1980s and is the only laboratory in Malaysia equipped with a research reactor, namely the TRIGA MARK II. Throughout the years the development of NAA technique has been very encouraging and was made applicable to a wide range of samples. At present, the k 0 method has become the preferred standardization method of NAA (k 0 -NAA) due to its multi-elemental analysis capability without using standards. Additionally, the k 0 method describes NAA in physically and mathematically understandable definitions and is very suitable for computer evaluation. Eventually, the k 0 -NAA method has been adopted by MINT in 2003, in collaboration with the Nuclear Research Institute (NRI), Vietnam. The reactor neutron parameters (α and f) for the pneumatic transfer system and for the rotary rack at various locations, as well as the detector efficiencies were determined. After calibration of the reactor and the detectors, the implemented k 0 method was validated by analyzing some certified reference materials (including IAEA Soil 7, NIST 1633a, NIST 1632c, NIST 1646a and IAEA 140/TM). The analysis results of the CRMs showed an average u score well below the threshold value of 2 with a precision of better than ±10% for most of the elemental concentrations obtained, validating herewith the introduction of the k 0 -NAA method at the MINT

  6. The future of the IPR-R1 TRIGA MARK I reactor after 48 years operation

    International Nuclear Information System (INIS)

    Maretti, Fausto Junior; Sette Camara, Luiz Otavio I.; Oliveira, Paulo Fernando

    2008-01-01

    The TRIGA Mark I IPR-R1 Reactor operates in the Nuclear Technology Development Center/ Brazilian Committion for Nuclear Energy (CDTN/CNEN), originally Institute of Radioactive Researches, in Belo Horizonte, Minas Gerais, since November 6, 1960. Initially it operated for isotope production for different uses, being later used in wide scale for another purposes as analyses for activation with neutrons and training of nuclear power plants operators. Dozens of degree theses were also developed with the use of the reactor. Along the years, several improvements were introduced in the reactor and its auxiliary systems, with the purpose to provide better use of the facilities and with the objective to increase the safety in the operation. The reactor is ready right now to operate at 250 kW, and for sure the nuclear applications programmed will be improved. The Operation Manual and the Safety Analysis report were already modified, as well as the Emergency Plan and the relative procedures to the same. After the tests at the end of 2008, the reactor will already be operating in the new power. This work presents a description of the several accomplishments of the last years and comments about the possibility of new uses for the reactor in the several areas of nuclear applications and some of the experiments and tests results during the upgrading program. (authors)

  7. Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z., E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN/RJ), Rio de Janeiro, RJ (Brazil); Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L., E-mail: claubia@nuclear.ufmg.b, E-mail: dora@nuclear.ufmg.b [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2011-07-01

    The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

  8. Operational parameters study of IPR-R1 TRIGA research reactor using virtual instruments

    International Nuclear Information System (INIS)

    Pinto, Antonio Juscelino; Mesquita, Amir Zacarias; Lameiras, Fernando Soares

    2013-01-01

    The instrumentation of nuclear reactors is designed with the principle of reliability, redundancy and diversification of control systems. Reliable monitoring of the parameters involved in the chain reaction is of great importance regarding efficiency and operational safety of the installation. The main goal of the simulation system in this proposed paper is to provide the study and improvement in understanding how these operational variables are interrelated and their behavior especially those related to neutronic and thermohydraulics. The work will be developed using the software LabVIEW ® (Laboratory Virtual Instruments Engineering Workbench). The program will enable the study of the variables involved in the operation of the installation throughout its operating range, for instance, a few mW up to 250 kW. The IPR-R1 TRIGA is a research nuclear reactor placed in open pool and cooled by light water with natural circulation. It is located at the Nuclear Technology Development Center (CDTN), in Belo Horizonte Brazil. The developing system employs the modern concept of virtual instruments (VIs), using microprocessors and visual interface on video monitors. LabVIEW ® breaks the paradigm of text-based programming language, for programming based on icons. The system will enable the use of this reactor in training and personnel training in the nuclear field. The work follows the recommendations of the International Atomic Energy Agency (IAEA), which has encouraged its members to develop strategic plans in order to use their research reactors. (author)

  9. Study on Operator Actions during the Occurrences of Undesirable Events in PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    Tom, P.P.; Nurul Husna Zainal Abidin; Lanyau, T.A.; Zaredah Hashim

    2016-01-01

    Due to the recent Fukushima accident, the potential risks at one and only nuclear research reactor in the country, which is the PUSPATI TRIGA Reactor (RTP), has increasingly gain concerns and an attempt on the development of Level 1 Probabilistic Safety Assessment (PSA) for this reactor has been commenced. The preliminary scope of the PSA is to analyse the risk of core degradation during normal daily operation due to the random component failure and human error. SPAR-H and THERP method is used for quantifying human error probability (HEP). However, the scopes of this study only cover the qualitative parts that use interview/questionnaire method. The objectives of the questionnaire are to identify the main action for RTP operators when any undesired incident occurs during full power operation that might be caused by random component failures. From the questionnaires that have been conducted, the respondents consisted of 4 licensed operators and 9 trainee operators. All licensed operators have experience of operating reactor for more than 15 years while the trainee operator have been operate the reactor with experience of less than 10 years. Generally, in the event of an abnormal condition involving the reactor, an operator whether a licensed operator or the trainee does not have to ask permission in advance from the top individuals to carry out scram. This is to prevent the situation becoming increasingly severe if the reactor is still operating. With complete training and knowledge derived from the management, an operator can act efficiently in any emergency case. (author)

  10. IPR-RI TRIGA MARK I reactor and the neutron activation analysis at CDTN/CNEN

    International Nuclear Information System (INIS)

    Menezes, Maria Angela de B.C.; Kastner, Geraldo F.; Amaral, Angela M.; Souza, Wagner de; Maretti, Fausto Junior; Leal, Alexandre S.

    2008-01-01

    The IPR-R1 TRIGA Mark I research reactor started up in 1960. It is located at Centro de Desenvolvimento da Tecnologia Nuclear (Nuclear Technology Development Centre) / Comissao Nacional de Energia Nuclear (Brazilian Commission for Nuclear Energy), CDTN/CNEN. Join to the reactor, the Laboratory for Neutron Activation Analysis has been developing its activities since 1960. The activities of the Laboratory comprise the delayed fission neutron activation analysis, instrumental (comparative and parametric methods) and radiochemical / chemical methods. These methods are responsible for relevant percentage of CDTN's analysis demand, meeting the clients' analytical needs and researches developed by the Laboratory, by CDTN and by other institutions. Over the years the work has been linked to the goals of the country and the institutions. Nowadays several elements - Ag, Al, Au, As, Ba, Br, Ca, Cd, Ce, Cl, Co, Cr, Cs, Cu, Dy, Eu, Fe, Ga, Hf, Hg, Ho, K, La, Mg, Mn, Mo, Na, Nd, Rb, Sb, Sc, Se, Sm, Sr, Ta, Tb, Th, Ti, U, V, W, Yb, Zn and Zr - are determined in several matrices and range of concentrations. In Brazil, CDTN is the only Institute that fully masters the instrumental neutron activation analysis k0-method determining short, medium and long half-life radionuclides using its own nuclear reactor. The good performance of the reactor is pointed out in a table with experimental and certified values for Certified Reference Materials. (authors)

  11. Neutron detection of the Triga Mark III reactor, using nuclear track methodology

    Energy Technology Data Exchange (ETDEWEB)

    Espinosa, G., E-mail: espinosa@fisica.unam.mx; Golzarri, J. I. [Instituto de Física, Universidad Nacional Autónoma de México Circuito de la Investigación Científica, Ciudad Universitaria. México, DF (Mexico); Raya-Arredondo, R.; Cruz-Galindo, S. [Instituto Nacional de Investigaciones Nucleares (Mexico); Sajo-Bohus, L. [Universidad Simón Bolivar, Laboratorio de Física Nuclear, Caracas (Venezuela, Bolivarian Republic of)

    2015-07-23

    Nuclear Track Methodology (NTM), based on the neutron-proton interaction is one often employed alternative for neutron detection. In this paper we apply NTM to determine the Triga Mark III reactor operating power and neutron flux. The facility nuclear core, loaded with 85 Highly Enriched Uranium as fuel with control rods in a demineralized water pool, provide a neutron flux around 2 × 10{sup 12} n cm{sup −2} s{sup −1}, at the irradiation channel TO-2. The neutron field is measured at this channel, using Landauer{sup ®} PADC as neutron detection material, covered by 3 mm Plexiglas{sup ®} as converter. After exposure, plastic detectors were chemically etched to make observable the formed latent tracks induced by proton recoils. The track density was determined by a custom made Digital Image Analysis System. The resulting average nuclear track density shows a direct proportionality response for reactor power in the range 0.1-7 kW. We indicate several advantages of the technique including the possibility to calibrate the neutron flux density measured at low reactor power.

  12. Estimation development cost, study case: Quality Management System Reactor TRIGA Mark III

    International Nuclear Information System (INIS)

    Antúnez Barbosa, Tereso Antonio; Valdovinos Rosas, Rosa María; Marcial Romero, José Raymundo; Ramos Corchado, Marco Antonio; Edgar Herrera Arriaga

    2016-01-01

    The process of estimating costs in software engineering is not a simple task, it must be addressed carefully to obtain an efficient strategy to solve problems associated with the effort, cost and time of activities that are performed in the development of an information system project. In this context the main goal for both developers and customers is the cost, since developers are worry about the effort pay-load and customers are worry about the product pay-load. However, in other fields the cost of goods depends on the activity or process that is performed, thereby deduce that the main cost of the final product of a development project software project is undoubtedly its size. In this paper a comparative study of common models for estimating costs are developed. These models are used today in order to create a structured analysis to provide the necessary information about cost, time and effort for making decisions in a software development project. Finally the models are applied to a case study, which is a system called Monitorizacion Automatica del Sistema de Gestion de Calidad del Reactor TRIGA Mark III. (author)

  13. Adaptive control method for core power control in TRIGA Mark II reactor

    Science.gov (United States)

    Sabri Minhat, Mohd; Selamat, Hazlina; Subha, Nurul Adilla Mohd

    2018-01-01

    The 1MWth Reactor TRIGA PUSPATI (RTP) Mark II type has undergone more than 35 years of operation. The existing core power control uses feedback control algorithm (FCA). It is challenging to keep the core power stable at the desired value within acceptable error bands to meet the safety demand of RTP due to the sensitivity of nuclear research reactor operation. Currently, the system is not satisfied with power tracking performance and can be improved. Therefore, a new design core power control is very important to improve the current performance in tracking and regulate reactor power by control the movement of control rods. In this paper, the adaptive controller and focus on Model Reference Adaptive Control (MRAC) and Self-Tuning Control (STC) were applied to the control of the core power. The model for core power control was based on mathematical models of the reactor core, adaptive controller model, and control rods selection programming. The mathematical models of the reactor core were based on point kinetics model, thermal hydraulic models, and reactivity models. The adaptive control model was presented using Lyapunov method to ensure stable close loop system and STC Generalised Minimum Variance (GMV) Controller was not necessary to know the exact plant transfer function in designing the core power control. The performance between proposed adaptive control and FCA will be compared via computer simulation and analysed the simulation results manifest the effectiveness and the good performance of the proposed control method for core power control.

  14. Computational analysis of neutronic parameters of CENM TRIGA Mark II research reactor

    International Nuclear Information System (INIS)

    El Younoussi, C.; El Bakkari, B.; Boulaich, Y.; Riyach, D.; Otmani, S.; Marrhich, I.; Badri, H.; Htet, A.; Nacir, B.; El Bardouni, T.; Boukhal, H.; Zoubair, M.; Ossama, M.; Chakir, E.

    2010-01-01

    The CENM TRIGA MARK II reactor is part of the National Center for Energy, Sciences and Nuclear Techniques (CNESTEN). It's a standard design 2MW, natural-convection-cooled reactor with a graphite reflector containing 4 beam tubes and a thermal column. The reactor has several applications in different fields as industry, agriculture, medicine, training and education. In the present work a computational study has been carried out in the framework of neutronic parameters studies of the reactor. A detailed MCNP model that include all elements of the core and surrounding structures has been developed to calculate different parameters of the core (The effective multiplication factor, reactivity experiments comprising control rods worth, excess reactivity and shutdown margin). Further calculations have been carried out to calculate the neutron flux profiles at different locations of the reactor core. The cross sections used are processed from the library provided with MCNP5 and based on the ENDF/B-VII with continuous dependence in energy and special treatment of thermal neutrons in lightweight materials. (author)

  15. Development and validation of a model TRIGA Mark III reactor with code MCNP5

    International Nuclear Information System (INIS)

    Galicia A, J.; Francois L, J. L.; Aguilar H, F.

    2015-09-01

    The main purpose of this paper is to obtain a model of the reactor core TRIGA Mark III that accurately represents the real operating conditions to 1 M Wth, using the Monte Carlo code MCNP5. To provide a more detailed analysis, different models of the reactor core were realized by simulating the control rods extracted and inserted in conditions in cold (293 K) also including an analysis for shutdown margin, so that satisfied the Operation Technical Specifications. The position they must have the control rods to reach a power equal to 1 M Wth, were obtained from practice entitled Operation in Manual Mode performed at Instituto Nacional de Investigaciones Nucleares (ININ). Later, the behavior of the K eff was analyzed considering different temperatures in the fuel elements, achieving calculate subsequently the values that best represent the actual reactor operation. Finally, the calculations in the developed model for to obtain the distribution of average flow of thermal, epithermal and fast neutrons in the six new experimental facilities are presented. (Author)

  16. Verification of MCNP simulation of neutron flux parameters at TRIGA MK II reactor of Malaysia

    International Nuclear Information System (INIS)

    Yavar, A.R.; Khalafi, H.; Kasesaz, Y.; Sarmani, S.; Yahaya, R.; Wood, A.K.; Khoo, K.S.

    2012-01-01

    A 3-D model for 1 MW TRIGA Mark II research reactor was simulated. Neutron flux parameters were calculated using MCNP-4C code and were compared with experimental results obtained by k 0 -INAA and absolute method. The average values of φ th ,φ epi , and φ fast by MCNP code were (2.19±0.03)×10 12 cm −2 s −1 , (1.26±0.02)×10 11 cm −2 s −1 and (3.33±0.02)×10 10 cm −2 s −1 , respectively. These average values were consistent with the experimental results obtained by k 0 -INAA. The findings show a good agreement between MCNP code results and experimental results. - Highlights: ► We use 3-D neutronic model to enhance the utilization of the economical use of reactor.► The MCNP code is modified to analyze the neutronic parameters. ► The neutron flux distributions are homogeneous and consistent with experimental results. ► A complete simulation of the calculated neutron flux parameters in 40 RR irradiation channels is achieved.

  17. Boron Neutron Capture Therapy activity of diffused tumors at TRIGA Mark II in Pavia

    International Nuclear Information System (INIS)

    Bortolussi, S.; Stella, S.; De Bari, A.; Altieri, S.; Bruschi, P.; Bakeine, J.G.; Clerici, A.; Ferrari, C.; Zonta, C.; Zonta, A.; Nano, R.

    2008-01-01

    The Boron neutron Capture Therapy research in Pavia has a long tradition: it begun more than 20 years ago at the TRIGA Mark II reactor of the University. A technique for the treatment of the hepatic metastases was developed, consisting in explanting the liver treated with 10 B, irradiating it in the thermal column of the reactor, and re-implanting the organ in the patient. In the last years, the possibility of applying BNCT to the lung tumours using epithermal collimated neutron beams and without explanting the organ, is being explored. The principal obtained results of the BNCT research will be presented, with particular emphasis on the following aspects: a) the project of a new thermal column configuration to make the thermal neutron flux more uniform inside the explanted liver, b) the Monte Carlo study by means of the MCNP code of the thermal neutron flux distribution inside a patient's thorax irradiated with epithermal neutrons, and c) the measurement of the boron concentration in tissues by (n,α) spectroscopy and neutron autoradiography. (authors)

  18. Neutron spectra in two beam ports of the TRIGA Mark III reactor

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H. R.; Hernandez D, V. M. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98060 Zacatecas (Mexico); Aguilar, F.; Paredes, L. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Rivera M, T., E-mail: fermineutron@yahoo.com [IPN, Centro de Investigacion en Ciencia Aplicada y Tecnologia Avanzada, Unidad Legaria, Av. Legaria 694, 11500 Mexico D. F. (Mexico)

    2013-10-15

    The neutron spectra have been measured in two beam ports, radial and tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research. Measurements were carried out with the core with mixed fuel (Leu 8.5/20 and Flip Heu 8.5/70). Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a {sup 6}Lil(Eu) scintillator and 2, 3, 5, 8, 10 and 12 inches-diameter high density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code; from each spectrum the total neutron flux, the neutron mean energy and the neutron ambient dose equivalent dose were determined. Measured spectra show fission (E≥ 0.1 MeV), epithermal (from 0.4 eV up to 0.1 MeV) and thermal neutrons (E≤ 0.4 eV). For both reactor powers the spectra in the radial beam port have similar features which are different to the neutron spectrum characteristics in the tangential beam port. (Author)

  19. Utilization of a typical 250 kW TRIGA Mark II reactor at a University

    International Nuclear Information System (INIS)

    Villa, M.; Boeck, H.; Musilek, A.

    2007-01-01

    The 250 kW TRIGA Mark-II reactor operates since March 1962 at the Atominstitut Vienna/Austria. Its main tasks are nuclear education and training in the fields of neutron- and solid state physics, nuclear technology, reactor safety, radiochemistry, radiation protection and dosimetry, and low temperature physics and fusion research. Academic research is carried out by students in the above mentioned fields coordinated and supervised by about 70 staff members with the aim of a masters- or PhD degree in one of the above mentioned areas. During the past 15 years about 580 students graduated through the Atominstitut. In addition, the Atominstitut co-operates closely with the nearby located IAEA in research projects, coordinated research programs (CRP) and supplying expert services. Regular training courses are carried out for the IAEA for Safeguard Trainees, fellowship places are offered for scientists from developing countries and staff members carry out expert missions to research centres in Africa, Asia and South America. Special Nuclear Material (SNM) is stored for calibration purposes at the Atominstitut belonging to the IAEA. (author)

  20. Study of silica (Titania) aerogels using MYSANS at MINT 1 MW TRIGA reactor

    International Nuclear Information System (INIS)

    Abdul Aziz Mohamed; Faridah Mohd Idris; Razali Kassim

    2006-01-01

    Small angle neutron scattering (SANS) technique has been widely employed in probing the microstructure of amorphous materials in the nano structure range; 1-100 nm. In this study, SANS was used to study the structure of the silica aerogels with and without titania nanoparticles. In aerogel system, the size range of 1 to 100 nm is of particular interest since the structural units, such as the pores and particles often fall in this range. Data collected was consistent with present models for the structure of silica aerogels, and an increase in mass fractal dimension from 2.3 to 2.6 for titania containing aerogels was observed. Preliminary SANS data was collected using the MYSANS instrument on the MINT PUSPATI TRIGA reactor. The neutron beam has a wavelength of 0.5 nm. The sample in powder form had dimensions 18x37mm 2 and 2 mm path length. The complete data collection consists of three measurements: sample scattering, empty sample holder scattering, and detector dark current. The scattered neutrons were detected by a 128 x 128 array area sensitive detector, proportional counter (PSD). The resulting 2D patterns were reduced to 1D profiles for further analysis. Plots of intensity, I(Q) versus momentum transfer, Q, were derived. For comparison of this work, the samples had been analysed using BATAN SANS facility. This work has demonstrated that SANS facility at MINT, mySANS, is capable to provide information of fractal dimension. (Author)

  1. Cooling Performance Analysis of ThePrimary Cooling System ReactorTRIGA-2000Bandung

    Science.gov (United States)

    Irianto, I. D.; Dibyo, S.; Bakhri, S.; Sunaryo, G. R.

    2018-02-01

    The conversion of reactor fuel type will affect the heat transfer process resulting from the reactor core to the cooling system. This conversion resulted in changes to the cooling system performance and parameters of operation and design of key components of the reactor coolant system, especially the primary cooling system. The calculation of the operating parameters of the primary cooling system of the reactor TRIGA 2000 Bandung is done using ChemCad Package 6.1.4. The calculation of the operating parameters of the cooling system is based on mass and energy balance in each coolant flow path and unit components. Output calculation is the temperature, pressure and flow rate of the coolant used in the cooling process. The results of a simulation of the performance of the primary cooling system indicate that if the primary cooling system operates with a single pump or coolant mass flow rate of 60 kg/s, it will obtain the reactor inlet and outlet temperature respectively 32.2 °C and 40.2 °C. But if it operates with two pumps with a capacity of 75% or coolant mass flow rate of 90 kg/s, the obtained reactor inlet, and outlet temperature respectively 32.9 °C and 38.2 °C. Both models are qualified as a primary coolant for the primary coolant temperature is still below the permitted limit is 49.0 °C.

  2. Upgrade of the ultracold neutron source at the pulsed reactor TRIGA Mainz

    Science.gov (United States)

    Kahlenberg, J.; Ries, D.; Ross, K. U.; Siemensen, C.; Beck, M.; Geppert, C.; Heil, W.; Hild, N.; Karch, J.; Karpuk, S.; Kories, F.; Kretschmer, M.; Lauss, B.; Reich, T.; Sobolev, Y.; Trautmann, N.

    2017-11-01

    The performance of the upgraded solid deuterium ultracold neutron source at the pulsed reactor TRIGA Mainz is described. The current configuration stage comprises the installation of a He liquefier to run UCN experiments over long-term periods, the use of stainless steel neutron guides with improved transmission as well as sputter-coated non-magnetic 58NiMo alloy at the inside walls of the thermal bridge and the converter cup. The UCN yield was measured in a "standard" UCN storage bottle (stainless steel) with a volume of 32 litres outside the biological shield at the experimental area yielding UCN densities of 8.5/cm3; an increase by a factor of 3.5 compared to the former setup. The measured UCN storage curve is in good agreement with the predictions from a Monte Carlo simulation developed to model the source. The growth and formation of the solid deuterium converter during freeze-out are affected by the ortho/para ratio of the H2 premoderator.

  3. Critical heat flux in natural convection cooled TRIGA reactors with hexagonal bundle

    International Nuclear Information System (INIS)

    Yang, J.; Avery, M.; De Angelis, M.; Anderson, M.; Corradini, M.; Feldman, E. E.; Dunn, F. E.; Matos, J. E.

    2012-01-01

    A three-rod bundle Critical Heat Flux (CHF) study at low flow, low pressure, and natural convection condition has been conducted, simulating TRIGA reactors with the hexagonally configured core. The test section is a custom-made trefoil shape tube with three identical fuel pin heater rods located symmetrically inside. The full scale fuel rod is electrically heated with a chopped-cosine axial power profile. CHF experiments were carried out with the following conditions: inlet water subcooling from 30 K to 95 K; pressure from 110 kPa to 230 kPa; mass flux up to 150 kg/m 2 s. About 50 CHF data points were collected and compared with a few existing CHF correlations whose application ranges are close to the testing conditions. Some tests were performed with the forced convection to identify the potential difference between the CHF under the natural convection and forced convection. The relevance of the CHF to test parameters is investigated. (authors)

  4. Shielding design of highly activated sample storage at reactor TRIGA PUSPATI

    International Nuclear Information System (INIS)

    Naim Syauqi Hamzah; Julia Abdul Karim; Mohamad Hairie Rabir; Muhd Husamuddin Abdul Khalil; Mohd Amin Sharifuldin Salleh

    2010-01-01

    Radiation protection has always been one of the most important things considered in Reaktor Triga PUSPATI (RTP) management. Currently, demands on sample activation were increased from variety of applicant in different research field area. Radiological hazard may occur if the samples evaluation done were misjudge or miscalculated. At present, there is no appropriate storage for highly activated samples. For that purpose, special irradiated samples storage box should be provided in order to segregate highly activated samples that produce high dose level and typical activated samples that produce lower dose level (1 - 2 mR/ hr). In this study, thickness required by common shielding material such as lead and concrete to reduce highly activated radiotracer sample (potassium bromide) with initial exposure dose of 5 R/ hr to background level (0.05 mR/ hr) were determined. Analyses were done using several methods including conventional shielding equation, half value layer calculation and Micro shield computer code. Design of new irradiated samples storage box for RTP that capable to contain high level gamma radioactivity were then proposed. (author)

  5. Shield design of concrete wall between decay tank room and primary pump room in TRIGA facility

    International Nuclear Information System (INIS)

    Khan, M. J. H.; Rahman, M.; Haque, A.; Zulquarnain, A.; Ahmed, F. U.; Bhuiyan, S. I.

    2007-01-01

    The objective of this study is to recommend the radiation protection design parameters from the shielding point of view for concrete wall between the decay tank room and the primary pump room in TRIGA Mark-II research reactor facility. The shield design for this concrete wall has been performed with the help of Point-kernel Shielding Code Micro-Shield 5.05 and this design was also validated based on the measured dose rate values with Radiation Survey Meter (G-M Counter) considering the ICRP-60 (1990) recommendations for occupational dose rate limit (10 μSv/hr). The recommended shield design parameters are: (i) thickness of 114.3 cm Ilmenite-Magnetite Concrete (IMC) or 129.54 cm Ordinary Reinforced Concrete (ORC) for concrete wall A (ii) thickness of 66.04 cm Ilmenite-Magnetite Concrete (IMC) or 78.74 cm Ordinary Reinforced Concrete (ORC) for concrete wall B and (iii) door thickness of 3.175 cm Mild Steel (MS) on the entrance of decay tank room. In shielding efficiency analysis, the use of I-M concrete in the design of this concrete wall shows that it reduced the dose rate by a factor of at least 3.52 times approximately compared to ordinary reinforced concrete

  6. Assessment of gold flux monitor at irradiation facilities of MINT TRIGA MK II reactor

    International Nuclear Information System (INIS)

    Wee Boon Siong; Abdul Khalik Wood; Mohd Suhaimi Hamzah; Shamsiah Abdul Rahman; Md Suhaimi Elias; Nazaratul Ashifa Abd Salim

    2005-01-01

    Neutron source of MINTs TRIGA MK II reactor has been used for activation analysis for many years and neutron flux plays important role in activation of samples at various positions. Currently, two irradiation facilities namely the pneumatic transfer system and rotary rack are available to cater for short and long lived irradiation. Neutron flux variation for both irradiation facilities have been determined using gold wire and gold solution as flux monitor. However, the use of gold wire as flux monitor is costlier if compared to gold solution. The results from analysis of certified reference materials showed that gold solution as flux monitors yield satisfactory results and proved to safe cost on the purchasing of gold wire. Further experiment on self-shielding effects of gold solution at various concentrations has been carried out. This study is crucial in providing vital information on the suitable concentration for gold solution as flux monitor. In the near future, gold solution flux monitor will be applied for routine analysis and hence to improve the capability of the laboratory on neutron activation analysis. (Author)

  7. Verification of the linearity of the IPR-R1 TRIGA reactor power channels

    International Nuclear Information System (INIS)

    Souza, Rose Mary Gomes do Prado; Campolina, Daniel de Almeida Magalhaes

    2013-01-01

    The aim of this paper is to verify the linearity of the three power channels of the IPR-R1 TRIGA reactor. Located at Nuclear Technology Development Center-CDTN in Belo Horizonte, the IPR-R1 reactor is a typical 100 kW Mark I light-water reactor cooled by natural convection. When the experiments were performed, the reactor core had 59 fuel elements, containing 8% by weight of uranium enriched to 20% in 235 U. The core has cylindrical configuration with an annular graphite reflector. The responses of the detectors of the Linear, Log N and Percent Power channels were compared with the responses of detectors which only depend on the overall neutron flux within the reactor. Gold and cobalt foils were activated at low and high powers, respectively, and the specific count results were compared with measurements performed, simultaneously, with a fission chamber, and with the power registered by the three channels. The results show that the Linear channel responds linearly up to 100 kW, and the Log N channel responses are linear at low powers. In the range of high power, the Log N and the Percent Power channels exhibit linearity only from 10 kW to 50 kW. (author)

  8. Temporal variation of the neutron flux in the carousel facility of a TRIGA reactor

    International Nuclear Information System (INIS)

    Jacimovic, R.; Stegnar, P.; Trkov, A.

    2003-01-01

    In this work we focused on identifying quantitatively the temporal (time-dependent) variation of neutron flux in the carousel facility (CF) of TRIGA reactor at the 'Jozef Stefan' Institute (IJS) for core No. 176, set up in April 2002. The measurements are based on neutron detectors (ionisation chambers), which surround the graphite reflector of the reactor core. In principle, the variations of the neutron flux produce a systematic error in the results obtained by absolute or 'quasi' absolute measuring techniques (such as neutron activation analysis (NAA) by the k 0 -standardization method), which assume constant conditions during irradiation. The results of our study show that for typical irradiation of 20 hours in channels of the CF aligned in the direction of the ionisation chamber (safety channel) the time-dependent variation of the neutron flux is about 6-8%. In the k 0 method, which we are using for routine work at the IJS, this variation introduced a systematic error in the results up to 4.6%, depending on the half-life of investigated radionuclide. (author)

  9. Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

    International Nuclear Information System (INIS)

    Mesquita, Amir Z.; Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L.

    2011-01-01

    The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

  10. Neutron flux measurement in the central channel (XC1) of TRIGA 14 MW LEU core

    International Nuclear Information System (INIS)

    Barbos, D.; Busuioc, P.; Paunoiu, C.; Roth, Cs.

    2008-01-01

    The full conversion of the 14 MW TRIGA Research Reactor was completed in May 2006 and each step of the conversion was achieved by removal of HEU fuel and replaced by LEU fuel. The operation was accompanied by a large set of theoretical evaluations and physical measurements intended to confirm the performances of gradual conversion. After the core full conversion, a program of measurements and comparisons with previous results of core physics and measurements is underway, allowing data acquisition for normal operation, demonstration of safety and economics of the converted core. Neutron flux spectrum measurements in the XC-1 water-filled channel were performed using multi-foil activation techniques. The neutron spectra and flux are obtained by unfolding from measured reaction rates using SAND II computer code. The integral flux measured value for LEU core and 14 MW reactor power is 4.66x10 14 n/cm 2 s. For standard core the integral neutron flux for 14 MW reactor power was 4.27x10 14 n/cm 2 s. (authors)

  11. Immobilization of Ion Exchange radioactive resins of the TRIGA Mark III Nuclear Reactor

    International Nuclear Information System (INIS)

    Garcia Martinez, H.

    1999-01-01

    In the last decades many countries in the world have taken interest in the use, availability, and final disposal of dangerous wastes in the environment, within these, those dangerous wastes that contain radioactive material. That is why studies have been made on materials used as immobilization agent of radioactive waste that may guarantee its storage for long periods of time under drastic conditions of humidity, temperature change and biodegradation. In mexico, the development of different applications of radioactive material in the industry, medicine and investigation, have generated radioactive waste, sealed and open sources, whose require a special technological development for its management and final disposal. The present work has as a finality to develop the process and define the agglutinating material, bitumen, cement and polyester resin that permits immobilization of resins of Ionic Exchange contaminated by Barium 153, Cesium 137, Europium 152, Cobalt 60 and Manganese 54 generated from the nuclear reactor TRIGA Mark III. Ionic interchange contaminated resin must be immobilized and is analysed under different established tests by the Mexican Official Standard NOM-019-NUCL-1995 L ow level radioactive wastes package requirements for its near-surface final disposal. Immobilization of ionic interchange contaminated resins must count with the International Standards applicable in this process; in these standards, the following test must be taken in prototype examples: Free-standing water, leachability, compressive strength, biodegradation, radiation stability, thermal stability and burning rate. (Author)

  12. Conversion of the core of the TRIGA Mark III reactor at the Mexican Nuclear Centre

    International Nuclear Information System (INIS)

    Moran Lopez, J.M.; Lucatero, M.A.; Reyes Andrade, B.; Rivero Gutierrez, T.; Sainz Mejia, E.

    1990-01-01

    It was decided to convert the core of the TRIGA MARK III reactor at the Mexican Nuclear Centre run by the National Nuclear Institute because of problems detected during the operation, such as a lack of excess reactivity for operation at nominal power over long periods and difficulties in the maintenance and calibration of the control panel. In order to compensate for the lack of excess reactivity the fuel elements taken to the highest burnup were replaced by fresh elements acquired for this purpose. The latter, however, had a different enrichment, and this necessitated a detailed analysis of the neutronic and thermohydraulic behaviour of the reactor with a view to determining a mixed core configuration which would meet safe operation requirements. In conducting the thermohydraulic analysis, a natural convection coolant flow model was developed to determine coolant velocity and pressure drop patterns within the core. The heat transfer equations were solved and it was found that the hottest fuel element did not attain critical heat flux conditions. In loading this core it was also necessary to analyse procedures and to consider the possible effects of reaching criticality with fuel elements having different enrichments. The loading procedure is described, as is the measurement system and the results obtained. In order to resolve the calibration and maintenance problems, a new, more advanced control panel was designed with conventional and nuclear detection systems and modern components

  13. Upgrade of the ultracold neutron source at the pulsed reactor TRIGA Mainz

    Energy Technology Data Exchange (ETDEWEB)

    Kahlenberg, J.; Ross, K.U.; Beck, M.; Heil, W.; Karch, J.; Kories, F.; Kretschmer, M. [Johannes Gutenberg University, Institute of Physics, Mainz (Germany); Ries, D. [Johannes Gutenberg University, Institute of Nuclear Chemistry, Mainz (Germany); Paul Scherrer Institute (PSI), Laboratory for Particle Physics, Villigen (Switzerland); ETH Zuerich, Institute for Particle Physics, Zuerich (Switzerland); Siemensen, C.; Geppert, C.; Karpuk, S.; Reich, T.; Sobolev, Y.; Trautmann, N. [Johannes Gutenberg University, Institute of Nuclear Chemistry, Mainz (Germany); Hild, N. [Paul Scherrer Institute (PSI), Laboratory for Particle Physics, Villigen (Switzerland); ETH Zuerich, Institute for Particle Physics, Zuerich (Switzerland); Lauss, B. [Paul Scherrer Institute (PSI), Laboratory for Particle Physics, Villigen (Switzerland)

    2017-11-15

    The performance of the upgraded solid deuterium ultracold neutron source at the pulsed reactor TRIGA Mainz is described. The current configuration stage comprises the installation of a He liquefier to run UCN experiments over long-term periods, the use of stainless steel neutron guides with improved transmission as well as sputter-coated non-magnetic {sup 58}NiMo alloy at the inside walls of the thermal bridge and the converter cup. The UCN yield was measured in a ''standard'' UCN storage bottle (stainless steel) with a volume of 32 litres outside the biological shield at the experimental area yielding UCN densities of 8.5/cm{sup 3}; an increase by a factor of 3.5 compared to the former setup. The measured UCN storage curve is in good agreement with the predictions from a Monte Carlo simulation developed to model the source. The growth and formation of the solid deuterium converter during freeze-out are affected by the ortho/para ratio of the H{sub 2} premoderator. (orig.)

  14. MCNP5 modeling of the IPR-R1 TRIGA reactor for criticality calculation and reactivity determination

    International Nuclear Information System (INIS)

    Silva, Clarysson A.M. da; Pereira, Claubia; Guerra, Bruno T.; Veloso, Maria Auxiliadora F.; Costa, Antonella L.; Dalle, Hugo M.

    2011-01-01

    Highlights: ► Two models of IPR-R1 TRIGA using the MCNP5 code were simulated. ► It obtained k eff values in some different situations of the reactor operation. ► The first model analyzes the criticality and the neutronic flux over the reactor. ► The second model includes the radial and axial neutron flux evaluation with different operation conditions. ► The results present good agreement with respect to the experimental data. - Abstract: The IPR-R1 TRIGA is a research nuclear reactor managed and located at the Nuclear Technology Development Center (CDTN) a research institute of the Brazilian Nuclear Energy Commission (CNEN). It is mainly used to radioisotopes production, scientific experiments, training of nuclear engineers for research and nuclear power plant reactor operation, experiments with materials and minerals and neutron activation analysis. In this work, criticality calculation and reactivity changes are presented and discussed using two modelings of the IPR-R1 TRIGA in the MCNP5 code. The first model (Model 1) analyzes the criticality over the reactor. On the other hand, the second model (Model 2) includes the possibility of radial and axial neutron flux evaluation with different operation conditions. The calculated results are compared with experimental data in different situations. For the two models, the standard deviation and relative error presented values of around 4.9 × 10 −4 . Both models present good agreement with respect to the experimental data. The goal is to validate the models that could be used to determine the neutron flux profiles to optimize the irradiation conditions, as well as to study reactivity insertion experiments and also to evaluate the fuel composition.

  15. Neutronic and thermal-hydraulic analysis of new irradiation channels inside the Moroccan TRIGA Mark II research reactor core.

    Science.gov (United States)

    Chham, E; El Bardouni, T; Benaalilou, K; Boukhal, H; El Bakkari, B; Boulaich, Y; El Younoussi, C; Nacir, B

    2016-10-01

    This study was conducted to improve the capacity of radioisotope production in the Moroccan TRIGA Mark II research reactor, which is considered as one of the most important applications of research reactors. The aim of this study is to enhance the utilization of TRIGA core in the field of neutron activation and ensure an economic use of the fuel. The main idea was to create an additional irradiation channel (IC) inside the core. For this purpose, three new core configurations are proposed, which differ according to the IC position in the core. Thermal neutron flux distribution and other neutronic safety parameters such as power peaking factors, excess reactivity, and control rods worth reactivity were calculated using the Monte Carlo N-Particle Transport (MCNP) code and neutron cross-section library based on ENDF/B-VII evaluation. The calculated thermal flux in the central thimble (CT) and in the added IC for the reconfigured core is compared with the thermal flux in the CT of the existing core, which is taken as a reference. The results show that all the obtained fluxes in CTs are very close to the reference value, while a remarkable difference is observed between the fluxes in the new ICs and reference. This difference depends on the position of IC in the reactor core. To demonstrate that the Moroccan TRIGA reactor could safely operate at 2MW, with new configurations based on new ICs, different safety-related thermal-hydraulic parameters were investigated. The PARET model was used in this study to verify whether the safety margins are met despite the new modifications of the core. The results show that it is possible to introduce new ICs safely in the reactor core, because the obtained values of the parameters are largely far from compromising the safety of the reactor. Copyright © 2016 Elsevier Ltd. All rights reserved.

  16. The new area monitoring system and the fuel database of the TRIGA Mark II reactor in Vienna

    International Nuclear Information System (INIS)

    Villa, M.; Boeck, H.; Hofbauer, M.; Schwarz, V.

    2004-01-01

    The 250 kW TRIGA Mark-II reactor operates since March 1962 at the Atominstitut, Vienna, Austria. Its main tasks are nuclear education and training in the fields of neutron- and solid state physics, nuclear technology, reactor safety, radiochemistry, radiation protection and dosimetry, and low temperature physics and fusion research. Academic research is carried out by students in the above mentioned fields coordinated and supervised by about 70 staff members with the aim of a masters- or PhD degree in one of the above mentioned areas. After 25 years of successful operation, it was necessary to exchange the old area monitoring system with a new digital one. The purpose of the new system is the permanent control of the reactor hall, the primary and secondary cooling system and the monitoring of the ventilation system. The paper describes the development and implementation of the new area monitoring system. The second topic in this paper describes the development of the new fuel database. Since March 7th, 1962, the TRIGA Mark II reactor Vienna operates with an average of 263 MWh per year, which corresponds to a uranium burn-up of 13.7 g per year. Presently we have 81 TRIGA fuel elements in the core, 55 of them are old aluminium clad elements from the initial criticality while the rest are stainless steel clad elements which had been added later to compensate the uranium consumption. Because 67 % of the elements are older than 40 years, it was necessary to put the history of every element in a database, to get an easy access to all the relevant data for every element in our facility. (author)

  17. Monte Carlo simulation of the thermal column and beam tube of the TRIGA Mark II research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Khan, R., E-mail: rustamzia@yahoo.com [Atominstitute (ATI), Vienna University of Technology (TU Wien), Stadion allee 2, A-1020 Vienna (Austria); Karimzadeh, S.; Stummer, T.; Boeck, H. [Atominstitute (ATI), Vienna University of Technology (TU Wien), Stadion allee 2, A-1020 Vienna (Austria)

    2011-08-15

    Highlights: > Neutronics parameters of the reactor shielding. > Biological shielding of the TRIGA reactor. > Thermal flux measurement in the thermal column and BT-A. > MCNP model validation. - Abstract: The Monet Carlo simulation of the TRIGA Mark II research reactor core has been performed employing the radiation transport computer code MCNP5. The model has been confirmed experimentally in the PhD research work at the Atominstitute (ATI) of the Vienna University of Technology. The MCNP model has been extended to complete biological shielding of the reactor including the thermal column, radiographic collimator and four beam tubes. This paper presents the MCNP simulated results in the thermal column and one of the beam tubes (beam tube A) of the reactor. To validate these theoretical results, thermal neutron flux density measurements using the gold foil activation method have been performed in the thermal column and beam tube A (BT-A). In the thermal column, the theoretical and experimental results are in fairly good agreement i.e. maximum thermal flux density in the centre decreases in radial direction. Further, it is also agreed that thermal flux densities in the lower part is greater than the upper part of the thermal column. In the BT-A experiment, the thermal flux density distribution is measured using gold foil. The experimental and theoretical diffusion lengths have been determined as 10.77 cm and 9.36 cm respectively with only 13% difference, reflecting good agreement between the experimental and simulated results. To save the computational cost and to incorporate the accurate and complete information of each individual Monte Carlo MC particle tracks, the surface source writing capability of MCNP has been utilized to the TRIGA shielding model. The variance reduction techniques have been applied to improve the statistics of the problem and to save computational efforts.

  18. Implementation of the k0 technique using multi-detectors on diverse irradiation facilities of TRIGA Reactor

    International Nuclear Information System (INIS)

    Caldera C, M. de G.

    2013-01-01

    The k 0 method with the technique of neutron activation analysis allows obtaining important characteristics parameters that describe a nuclear reactor. Among these parameters are the form factor of epithermal neutron flux, α and the ratio of thermal neutron flux with respect to the epithermal neutron flux, f. These parameters were obtained by irradiation of two different monitors, one of Au-Zr and the other of Au-Mo-Cr, where the last one was made and implemented for the first time. Both monitors were irradiated in different positions in the TRIGA Mark III Reactor at the National Institute of Nuclear Research. (Author)

  19. Capture programs, analysis, data graphication for the study of the thermometry of the TRIGA Mark III reactor core

    International Nuclear Information System (INIS)

    Paredes G, L.C.

    1991-05-01

    This document covers the explanation of the capture programs, analysis and graphs of the data obtained during the measurement of the temperatures of the instrumented fuel element of the TRIGA Mark III reactor and of the coolant one near to this fuel, using the conversion card from Analogic to Digital of 'Data Translation', and using a signal conditioner for five temperature measurers with the help of thermo par type K, developed by the Simulation and Control of the nuclear systems management department, which gives a signal from 0 to 10 Vcd for an interval of temperature of 0 to 1000 C. (Author)

  20. Epithermal neutron flux characterization of the TRIGA Mark III reactor, Salazar, Mexico, for use in Internal Neutron Activation Analysis

    International Nuclear Information System (INIS)

    Diaz Rizo, O.; Herrera Peraza, E.

    1996-01-01

    The non ideality of the epithermal neutron flux distribution at a reactor site parameter (made, using Chloramine-T method. Radiochemical purity and stability of the labelled product were determined by radiochromatography. The labelled Melagenine-II showed two radioactive fractions thermal-to-epithermal neutron ratio (f) were determined in the 3 typical irradiations positions of the TRIGA Mark III reactor of the National Nuclear Research Institute, Salazar, Mexico, using the Cd-ratio for multi monitor and bare bi-isotopic monitor methods respectively. This characterization is of use in the K o - method of neutron activation analysis, recently introduced at the Institute

  1. Handling of TRIGA spent fuel at the Medical University of Hanover and its return to the United States

    International Nuclear Information System (INIS)

    Hampel, Gabriele; Harke, Heinrich; Kelm, Wieland; Klaus, Uwe

    2008-01-01

    The Medical University of Hannover (MHH) was taking part in the US Department of Energy's (DOE) 'Research Reactor Spent Fuel Acceptance Program' to return its 76 spent TRIGA fuel elements to the United States in the middle of 1999. The fuel elements have been moved to the Idaho National Engineering and Environmental Laboratory (INEEL) in Idaho. This paper describes the technical facilities for handling the fuel elements at the MHH and the various steps in removing the fuel elements from the reactor, transferring them to the transport cask and shipping them to the INEEL. (authors)

  2. Irradiation positions for fission-track dating in the University of Pavia TRIGA Mark II nuclear reactor

    International Nuclear Information System (INIS)

    Oddone, Massimo; Meloni, Sandro; Balestrieri, Maria Laura; Bigazzi, Giulio

    2002-01-01

    An irradiation position arranged is described in the present paper for fission-track dating in the Triga Mark II reactor of the University of Pavia. Fluence values determined using the NIST glass standard SRM 962a for fission-track dating and the traditional metal foils are compared. Relatively good neutron thermalization (φ th /φ f = 0.956) and lack of significant fluence spatial gradients are good factors for fission-track dating. Finally, international age standards (or putative age standards) irradiated in this new position yielded results consistent with independent reference ages. (author)

  3. Evaluation for the status of the IAEA inspection at Hanaro and TRIGA Mark II and III reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun Sook; Lee, Byung Doo

    2007-11-15

    Safeguards implementation of nuclear material was carried out at facility level in an effect to support the peaceful nuclear activities in KAERI. Safeguards implementation is to fulfill the obligations associated with international agreements such as IAEA comprehensive safeguards agreement and additional protocol. IAEA inspection is the most important and basic factor of the safeguards implementation for the purpose of verifying whether all source or special fissionable material is diverted to nuclear weapons or other nuclear explosive devices. The status of the IAEA inspection at Hanaro and TRIGA Mark II and III reactor during 2001-2006 is evaluated in this report.

  4. Evaluation for the status of the IAEA inspection at Hanaro and TRIGA Mark II and III reactor

    International Nuclear Information System (INIS)

    Kim, Hyun Sook; Lee, Byung Doo

    2007-11-01

    Safeguards implementation of nuclear material was carried out at facility level in an effect to support the peaceful nuclear activities in KAERI. Safeguards implementation is to fulfill the obligations associated with international agreements such as IAEA comprehensive safeguards agreement and additional protocol. IAEA inspection is the most important and basic factor of the safeguards implementation for the purpose of verifying whether all source or special fissionable material is diverted to nuclear weapons or other nuclear explosive devices. The status of the IAEA inspection at Hanaro and TRIGA Mark II and III reactor during 2001-2006 is evaluated in this report

  5. Neutron flux measurements at the TRIGA reactor in Vienna for the prediction of the activation of the biological shield.

    Science.gov (United States)

    Merz, Stefan; Djuricic, Mile; Villa, Mario; Böck, Helmuth; Steinhauser, Georg

    2011-11-01

    The activation of the biological shield is an important process for waste management considerations of nuclear facilities. The final activity can be estimated by modeling using the neutron flux density rather than the radiometric approach of activity measurements. Measurement series at the TRIGA reactor Vienna reveal that the flux density next to the biological shield is in the order of 10(9)cm(-2)s(-1) at maximum power; but it is strongly influenced by reactor installations. The data allow the estimation of the final waste categorization of the concrete according to the Austrian legislation. Copyright © 2011 Elsevier Ltd. All rights reserved.

  6. Estimation of fast neutron fluence in steel specimens type Laguna Verde in TRIGA Mark III reactor; Estimacion de la fluencia de neutrones rapidos en probetas de acero tipo Laguna Verde en el reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Aguilar H, F., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    The main purpose of this work is to obtain the fluence of fast neutrons recorded within four specimens of carbon steel, similar to the material having the vessels of the BWR reactors of the nuclear power plant of Laguna Verde when subjected to neutron flux in a experimental facility of the TRIGA Mark III reactor, calculating an irradiation time to age the material so accelerated. For the calculation of the neutron flux in the specimens was used the Monte Carlo code MCNP5. In an initial stage, three sheets of natural molybdenum and molybdenum trioxide (MoO{sub 3}) were incorporated into a model developed of the TRIGA reactor operating at 1 M Wth, to calculate the resulting activity by setting a certain time of irradiation. The results obtained were compared with experimentally measured activities in these same materials to validate the calculated neutron flux in the model used. Subsequently, the fast neutron flux received by the steel specimens to incorporate them in the experimental facility E-16 of the reactor core model operating at nominal maximum power in steady-state was calculated, already from these calculations the irradiation time required was obtained for values of the neutron flux in the range of 10{sup 18} n/cm{sup 2}, which is estimated for the case of Laguna Verde after 32 years of effective operation at maximum power. (Author)

  7. Adaptive fuzzy control of neutron power of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Rojas R, E.

    2014-01-01

    The design and implementation of an identification and control scheme of the TRIGA Mark III research nuclear reactor of the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico is presented in this thesis work. The identification of the reactor dynamics is carried out using fuzzy logic based systems, in which a learning process permits the adjustment of the membership function parameters by means of techniques based on neural networks and bio-inspired algorithms. The resulting identification system is a useful tool that allows the emulation of the reactor power behavior when different types of insertions of reactivity are applied into the core. The identification of the power can also be used for the tuning of the parameters of a control system. On the other hand, the regulation of the reactor power is carried out by means of an adaptive and stable fuzzy control scheme. The control law is derived using the input-output linearization technique, which permits the introduction of a desired power profile for the plant to follow asymptotically. This characteristic is suitable for managing the ascent of power from an initial level n o up to a predetermined final level n f . During the increase of power, a constraint related to the rate of change in power is considered by the control scheme, thus minimizing the occurrence of a safety reactor shutdown due to a low reactor period value. Furthermore, the theory of stability in the sense of Lyapunov is used to obtain a supervisory control law which maintains the power error within a tolerance region, thus guaranteeing the stability of the power of the closed loop system. (Author)

  8. Elaboration of the configuration and programming of the interlocks system of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Mejia C, M. A.

    2016-01-01

    The modernization of the TRIGA Mark III reactor interlock system requires a system that provides high reliability, flexibility and ease of operation during reactor operation. With this modernization of the system, is intended to prevent, control and mitigate the causes of probable accidents reported in the reactor accident analysis. On the other hand, is foreseen the ease reactor operation in a simple, safe and efficient way. The programmable logic controller can be programmed by programming instructions using simple language and easy to develop, these can be modified from a computer using the programming software. In addition, another of the advantages offered by the controller is that can be modified from a touch screen (human-machine interface) that allows adjustment, without the need to use programming software and diagnostic functions during the process. As a result of the present work, a situation of improvement in the reactor operation was generated, facilitating the handling of the bridge and increasing the efficiency of the system in the execution of the operating conditions of the installations external to the reactor. A modern, more reliable and much less expensive system was achieved than the previous one, avoiding that the maintenance to the system generates high expenses. With respect to the development of the application programming, a control was implemented that allows to select a zone of the five that have inside the pool to carry out the displacement of automatic way and later to be located in that zone, having in this way a greater efficiency and ease in bridge control. (Author)

  9. Confirmation of a realistic reactor model for BNCT dosimetry at the TRIGA Mainz

    Energy Technology Data Exchange (ETDEWEB)

    Ziegner, Markus, E-mail: Markus.Ziegner.fl@ait.ac.at [AIT Austrian Institute of Technology GmbH, Vienna A-1220, Austria and Institute of Atomic and Subatomic Physics, Vienna University of Technology, Vienna A-1020 (Austria); Schmitz, Tobias; Hampel, Gabriele [Institut für Kernchemie, Johannes Gutenberg-Universität, Mainz DE-55128 (Germany); Khan, Rustam [Pakistan Institute of Engineering and Applied Sciences (PIEAS), Islamabad PK-44000 (Pakistan); Blaickner, Matthias [AIT Austrian Institute of Technology GmbH, Vienna A-1220 (Austria); Palmans, Hugo [Acoustics and Ionising Radiation Division, National Physical Laboratory, Teddington TW11 0LW, United Kingdom and Medical Physics Group, EBG MedAustron GmbH, Wiener Neustadt A-2700 (Austria); Sharpe, Peter [Acoustics and Ionising Radiation Division, National Physical Laboratory, Teddington TW11 0LW (United Kingdom); Böck, Helmuth [Institute of Atomic and Subatomic Physics, Vienna University of Technology, Vienna A-1020 (Austria)

    2014-11-01

    Purpose: In order to build up a reliable dose monitoring system for boron neutron capture therapy (BNCT) applications at the TRIGA reactor in Mainz, a computer model for the entire reactor was established, simulating the radiation field by means of the Monte Carlo method. The impact of different source definition techniques was compared and the model was validated by experimental fluence and dose determinations. Methods: The depletion calculation code ORIGEN2 was used to compute the burn-up and relevant material composition of each burned fuel element from the day of first reactor operation to its current core. The material composition of the current core was used in a MCNP5 model of the initial core developed earlier. To perform calculations for the region outside the reactor core, the model was expanded to include the thermal column and compared with the previously established ATTILA model. Subsequently, the computational model is simplified in order to reduce the calculation time. Both simulation models are validated by experiments with different setups using alanine dosimetry and gold activation measurements with two different types of phantoms. Results: The MCNP5 simulated neutron spectrum and source strength are found to be in good agreement with the previous ATTILA model whereas the photon production is much lower. Both MCNP5 simulation models predict all experimental dose values with an accuracy of about 5%. The simulations reveal that a Teflon environment favorably reduces the gamma dose component as compared to a polymethyl methacrylate phantom. Conclusions: A computer model for BNCT dosimetry was established, allowing the prediction of dosimetric quantities without further calibration and within a reasonable computation time for clinical applications. The good agreement between the MCNP5 simulations and experiments demonstrates that the ATTILA model overestimates the gamma dose contribution. The detailed model can be used for the planning of structural

  10. Licensing of the TRIGA Mark III reactor at the Mexican Nuclear Centre

    International Nuclear Information System (INIS)

    Ramirez, R.M.; Arrendondo, R.R.

    1990-01-01

    The TRIGA Mark III reactor at the Mexican Nuclear Centre went critical in 1968 and remained so until 1979 when the National Commission for Nuclear Safety and Safeguards (CNSNS), the Mexican regulatory authority, was set up. The reactor was therefore operating without a formal operating license, and the CNSNS accordingly requested the ININ to license the reactor under the existing conditions and to ensure that any modification of the original design complied with Standards ANSI/ANS-15 and with the code of practice set out in IAEA Safety Series No. 35. The most relevant points in granting the operating licence were: (a) the preparation of the Safety Report; (b) the formulation and application of the Quality Assurance Programme; (c) the reconditioning of the following reactor systems: the cooling systems; the ventilation and exhaust system; the monitoring system and control panel; (d) the training of the reactor operating staff at junior and senior levels; and (e) the formulation of procedures and instructions. Once the provisional operating license was obtained for the reactor it was considered necessary to modify the reactor core, which has been composed of 20% enriched standards fuel, to a mixed core based on a mixture of standard fuel and FLIP-type fuel with 70% 235 U enrichment. The CNSNS therefore requested that the mixed core be licensed and a technical report was accordingly annexed to the Safety Report, its contents including the following subjects: (a) neutron analysis of the proposed configuration; (b) reactor shutdown margins; (c) accident analysis; and (d) technical specifications. The licensing process was completed this year and we are now hoping to obtain the final operating license

  11. Three-dimensional modeling and virtual TRIGA reconfigure for specialized training

    International Nuclear Information System (INIS)

    Plata M, A. C.; Morales S, J. B.; Flores, M.

    2009-10-01

    The news products that have been realized for the training virtual room which is developing in the Engineering Faculty of National Autonomous University of Mexico are presented. These improvements are mainly in modeling of virtual reality of the reactor building, as well as internal parts of reactor. It was modified the dynamic modeling of control rods of reaction in chain and included new elements to reactor. which exist not necessarily in all the TRIGA, but that, for educational purposes are highly useful. Such is the case of addition of valves, pumps, tanks, injection lines of light or borated water, as well as a heat exchanger, with it can recycle only pool water from side to other, or to extract energy toward a secondary controller from the operator console. The models of heat decay were included, of subcooled and nucleated boiling of coolant-moderator in the core, the dynamics of xenon and samarium. These last with independent multipliers of simulation time to allow variations very fast that real time. All these additions modify the coolant-moderator characteristics and consequently the answer of simulator. The controls are separated in: an operator console (student) very similar to the real systems, another of instructor that has additional access to parameters not directly measurement in the facilities but that allow to modify the system to illustrate another not easily possible effects in the real system. The traveling crane is also modeled and is controlled in a third console from where can to replacement to reactor as well as to add or to replacement: intakes and discharges of coolant circulators, measuring instruments, reflectors and neutron sources. The dynamic models have been tested in SCILAB and SCICOS. At present is working in the integration of the dynamic simulator and the virtual reality mainly with the design requirement of allowing functions of increased reality. (Author)

  12. Experimental study of radiation dose rate at different strategic points of the BAEC TRIGA Research Reactor.

    Science.gov (United States)

    Ajijul Hoq, M; Malek Soner, M A; Salam, M A; Haque, M M; Khanom, Salma; Fahad, S M

    2017-12-01

    The 3MW TRIGA Mark-II Research Reactor of Bangladesh Atomic Energy Commission (BAEC) has been under operation for about thirty years since its commissioning at 1986. In accordance with the demand of fundamental nuclear research works, the reactor has to operate at different power levels by utilizing a number of experimental facilities. Regarding the enquiry for safety of reactor operating personnel and radiation workers, it is necessary to know the radiation level at different strategic points of the reactor where they are often worked. In the present study, neutron, beta and gamma radiation dose rate at different strategic points of the reactor facility with reactor power level of 2.4MW was measured to estimate the rising level of radiation due to its operational activities. From the obtained results high radiation dose is observed at the measurement position of the piercing beam port which is caused by neutron leakage and accordingly, dose rate at the stated position with different reactor power levels was measured. This study also deals with the gamma dose rate measurements at a fixed position of the reactor pool top surface for different reactor power levels under both Natural Convection Cooling Mode (NCCM) and Forced Convection Cooling Mode (FCCM). Results show that, radiation dose rate is higher for NCCM in compared with FCCM and increasing with the increase of reactor power. Thus, concerning the radiological safety issues for working personnel and the general public, the radiation dose level monitoring and the experimental analysis performed within this paper is so much effective and the result of this work can be utilized for base line data and code verification of the nuclear reactor. Copyright © 2017 Elsevier Ltd. All rights reserved.

  13. Modernization of the CDTN IPR-R1 TRIGA reactor instrumentation and control

    International Nuclear Information System (INIS)

    Mesquita, A.Z.; Costa, A.C.L.; Souza, R.M.G.P.

    2009-01-01

    The control system of the IPR-R1 was changed in 1995. Although since the year's 80 was generalized the use of microprocessor technology and video monitors for visual interface, in the IPR-R1 control room it was used analogical system by relay-based logic, and were maintained the mechanical strip chart recorders (ink-pen drive) to measure, monitor and store the operational parameters. It was maintained the measure and the control of, practically, the same variables of the original system, although the reactor power already have been upgraded to 100 kW and began the studies to increase it to 250 kW, which is the current core configuration. For 250 kW operations the fuel heat transfer becomes important and new parameters should be used as safety operational limits. A state-of-the-art instrumentation and control system using microprocessor technology is proposed to replace the present analogical systems. The new system can eliminates most manual data logging, provides automatic or manual reactor operation modes, provides complete real-time operator display, replays historical operating data on monitor or printer, eliminates spare parts replacement problems and meets all applicable international standards as NRC and IEE specifications. This paper describes the research project in process in CDTN that has as objective the modernization of the IPR-R1 TRIGA reactor instrumentation and control of the operational variables. The project also will improve the accomplishment of neutronic and thermal-hydraulic experiments, foreseen in the CDTN research program. (author)

  14. Visual inspection system and sipping design for spent fuel at TRIGA MARK III reactor of Mexico

    International Nuclear Information System (INIS)

    Delfin, A.; Mazon, R.

    2002-01-01

    In the framework of the Technical Cooperation Regional Project for Latin America RLA/4/018 for the biennium 2001-2002, one of the activities identified is the characterization of spent fuel. Of these activities an important one is not doubt the physical condition of spent fuel because an appropriate identification of the fuel status will prevent problems of fuel leaks, corrosion problems etc. As part of the activities of the project was decided that countries no having visual inspection and sipping systems should be very desirable to have them as a result of this project. The Triga reactor of Mexico does not have both of them, therefore, it was decided the need of having both system. The paper describe first the way we designed and constructed a remote Visual Inspection System and example of how is operated. Along the experience and problems we have had with the system. Also we will present the design of the Sipping system were two option were considered. First to take a sample of water after a convenient period of time passing through a circuit to a multichannel analyzer and to identify leakage by way of measuring Caesium-137. Second, exists the possibility that the Stainless Steel sleeve of the fuel has only very small failures, so it is going to be very difficult to have leakages unless the fuel is hot. Therefore we are evaluating the possibility of using heaters to increase the temperature of the fuel and succeed on detecting leakages. The results - we hope - will be ready to be presented at the meeting. (author)

  15. On the air coolability of TRIGA reactors following a loss-of-coolant accident

    International Nuclear Information System (INIS)

    El-Genk, Mohamed S.; Kim, Sung-Ho; Zaki, Galal M.; Foushee, Fabian; Philbin, Jeffrey S.; Schulze, James

    1986-01-01

    This paper describes the experiments on the air-coolability of a heated rod in a vertical open annulus at near atmospheric pressure. This data can be applied to the coolability of reactor fuel rods that are totally uncovered in a Loss-of-Coolant Accident (LOCA). As a prelude to measuring air coolability of specific core geometries (bundles), heat transfer data was collected for natural convection of atmospheric air in open vertical annuli with an isoflux inner wall and an insulated outer wall (diameter ratios, annulus ratio, of 1.155, 1.33, 1.63, and 12). Although the inner heated tube had the same overall dimensions as the fuel rod in the Annular Core Research Reactor (ACRR) at Sandia National Laboratories (3.81 cm o.d. and 55.5 cm long), the heated length was only 36.0 cm rather than the entire 50.5 cm for the ACRR's rods. The test assembly was operated at heat fluxes up to 1.38 W/cm 2 with a corresponding surface temperature of 852 K. The annulus data was extrapolated to an equilibrium surface temperature of 1200 K (as a coolability limit of TRIGA reactors) to provide a qualitative estimate of the coolability of multirod bundles by free convection of atmospheric air. The results suggest that for a typical pitch-to-diameter ratio of 1.12 in the ACRR the decay heat removal level is about 1.0 kW/m. This corresponds to an initial decay power following sustained operations at about 12.5 kW/m in the ACRR. However, because of the uncertainties in duplicating the actual thermal-hydraulic conditions in a multirod bundle using a single rod annulus, the actual coolability of open pool reactors could be different from those suggested in this paper. (author)

  16. License renewal and power upgrade of the Cornell University TRIGA reactor

    International Nuclear Information System (INIS)

    Aderhold, Howard C.

    1984-01-01

    The Cornell Mark II TRIGA reactor has been a principal facility for instruction and research in nuclear science and engineering at Cornell, and it has been extensively used by other departments at Cornell and by nearby universities and industries. Initially the fuel was low hydride, 8.5w/o 19%-enriched, aluminum clad; in 1974 it was changed to high-hydride, stainless-steel-clad. The maximum power has been 100 kW, with pulses to $2, and operation has been on a one-shift demand basis. Annual energy generation of 50 MWH has been typical. Standard features include a 4-inch tangential port and our 6-inch radial ports, a thermal column with hohlraum and vertical access, a central thimble, a 'rabbit', and a set of dry irradiation tubes, replacing the 'Lazy Susan'. The license was renewed and amended in November 1983; the new limits are 500 kW and $3 pulses. Physical changes to the facility included addition of a water-to-water heat exchanger and of a diffuser at the water outlet ∼ 60 cm above the core. The flow rate is 300 liters per minute in the primary (reactor) side of the heat exchanger. The temperature of the chilled water entering the secondary of the exchanger is ∼ 12?C; its flow rate is adjusted by a servo-controlled by-pass valve to maintain the desired range of pool water temperature. Steps taken to go to higher power included rearrangement of fuel elements to increase excess reactivity, recalibration of control rods, and power vs ion chamber current calibrations at successively higher power by comparing the rate of rise of pool temperature with a known rate using electrical heating elements. Steady-state operation has been done up to 480 kW (nominal) but pulsing at the newly allowed higher levels has not been tested as yet

  17. PENGARUH GEMPA PATAHAN LEMBANG TERHADAP FLEKSIBILITAS PIPA DAN KEGAGALAN NOZEL PERALATAN SISTEM PENDINGIN PRIMER REAKTOR TRIGA 2000 BANDUNG

    Directory of Open Access Journals (Sweden)

    H.P. Raharjo

    2012-02-01

    Full Text Available Gempa bumi di suatu instalasi nuklir dapat menyebabkan terjadinya kerusakan sistem peralatan karena adanya perubahan fleksibilitas pipa akibat beban berlebih pada sistem perpipaannya. Beban dari sistem perpipaan dapat membebani nozel peralatan seperti pompa, penukar panas, dll. Apabila beban tersebut terlalu besar dan melebihi beban yang diizinkan akan mengakibatkan kegagalan nozel peralatan. Penelitian ini menganalisis besarnya beban-beban yang terjadi di semua nozel  sistem pendingin primer Reaktor TRIGA 2000 Bandung. Analisis dilakukan dengan bantuan perangkat lunak Caesar II yang berbasis metode elemen hingga (finite element method. Acuan analisis yang digunakan adalah code ASME B31.1 yang mengatur tentang perancangan sistem perpipaan untuk sistem pembangkit daya (Power Piping. Hasil analisis menunjukkan bahwa gaya yang terjadi pada nozel masukan pompa arah aksial (FZ melebihi batas yang diijinkan. Oleh karena itu perlu dilakukan modifikasi terhadap jalur perpipaan sistem pendingin tersebut agar nozel tidak menerima gaya yang berlebih. Penyangga pipa titik(node 22 di jalur PriOut dilepas dan/atau dipindahkan pada nozel masukkan pompa. Perpipaan sistem pendingin primer reaktor TRIGA 2000 Bandung akan aman beroperasi jika terjadi gempa.yang berasal dari patahan Lembang Kata kunci : gempa bumi, patahan, perpipaan, reaktor, sistem pendingin

  18. Transient behavior during reactivity insertion in the Moroccan TRIGA Mark II reactor using the PARET/ANL code

    International Nuclear Information System (INIS)

    Boulaich, Y.; Nacir, B.; El Bardouni, T.; Boukhal, H.; Chakir, E.; El Bakkari, B.; El Younoussi, C.

    2015-01-01

    Highlights: • PARET model for the Moroccan TRIGA MARK II reactor has been developed. • Transient behavior under reactivity insertion has been studied based on PARET code. • Power factors required by PARET code have been calculated by using MCNP5 code. • The dependence on time of the main thermal-hydraulic parameters was calculated. • Results are largely far to compromise the thermal design limits. - Abstract: A three dimensional model for the Moroccan 2 MW TRIGA MARK II reactor has been developed for thermal-hydraulic and safety analysis by using the PARET/ANL and MCNP5 codes. This reactor is located at the nuclear studies center of Mâamora (CENM), Morocco. The model has been validated through temperature measurements inside two instrumented fuel elements located near the center of the core, at various power levels, and also through the power and fuel temperature evolution after the reactor shutdown (SCRAM). The axial distributions of power factors required by the PARET code have been calculated in each fuel element rod by using MCNP5 code. Based on this thermal-hydraulic model, a safety analysis under the reactivity insertion phenomenon has been carried out and the dependence on time of the main thermal-hydraulic parameters was calculated. Results were compared to the thermal design limits imposed to maintain the integrity of the clad

  19. Experimental study of the temperature distribution in the TRIGA IPR-R1 Brazilian research reactor; Investigacao experimental da distribuicao de temperaturas no reator nuclear de pesquisa TRIGA IPR-R1

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Zacarias

    2005-07-01

    The TRIGA-IPR-R1 Research Nuclear Reactor has completed 44 years in operation in November 2004. Its initial nominal thermal power was 30 kW. In 1979 its power was increased to 100 kW by adding new fuel elements to the reactor. Recently some more fuel elements were added to the core increasing the power to 250 kW. The TRIGA-IPR-R1 is a pool type reactor with a natural circulation core cooling system. Although the large number of experiments had been carried out with this reactor, mainly on neutron activation analysis, there is not many data on its thermal-hydraulics processes, whether experimental or theoretical. So a number of experiments were carried out with the measurement of the temperature inside the fuel element, in the reactor core and along the reactor pool. During these experiments the reactor was set in many different power levels. These experiments are part of the CDTN/CNEN research program, and have the main objective of commissioning the TRIGA-IPR-R1 reactor for routine operation at 250 kW. This work presents the experimental and theoretical analyses to determine the temperature distribution in the reactor. A methodology for the calibration and monitoring the reactor thermal power was also developed. This methodology allowed adding others power measuring channels to the reactor by using thermal processes. The fuel thermal conductivity and the heat transfer coefficient from the cladding to the coolant were also experimentally valued. lt was also presented a correlation for the gap conductance between the fuel and the cladding. The experimental results were compared with theoretical calculations and with data obtained from technical literature. A data acquisition and processing system and a software were developed to help the investigation. This system allows on line monitoring and registration of the main reactor operational parameters. The experiments have given better comprehension of the reactor thermal-fluid dynamics and helped to develop numerical

  20. Fuel element reactivity worth in different rings of the IPR-R1 TRIGA reactor

    International Nuclear Information System (INIS)

    Gomes do Prado Souza, Rose Mary

    2008-01-01

    The thermal power of the IPR-R1 TRIGA Reactor will be upgraded from 100 kW to 250 kW. Starting core: loaded with 59 aluminum cladded fuel elements; 1.34 $ excess reactivity; and 100 kW power. It is planned to go 2.5 times the power licensed, i.e., 250 kW. This forces to enlarge the reactivity level. Nuclear reactors must have sufficient excess reactivity to compensate the negative reactivity feedback effects caused by: the fuel temperature, fuel burnup, fission poisoning production, and to allow full power operation for predetermined period of time. To provide information for the calculation of the new core arrangement, the reactivity worth of some fuel elements in the core were measured as well as the determination of the core reactivity increase in the substitution of the original fuels, cladded with aluminium, for new ones, cladded with stainless steel. The reactivity worth of fuel element was measured from the difference in critical position of the control rods, calibrated by the positive period method, before and after the fuel element was withdrawn from the core. The magnitude of reactivity increase was determined when withdrawing the original Al-clad fuel (a little burned up) and the graphite elements, and inserting a fresh Al-clad fuel element, one by one. Experimental results indicated that to obtain enough reactivity excess to increase the rector power the addition of 4 new fuel elements in the core would be sufficient: - Substitution of 4 Al-clad fuel elements in ring C for fresh stainless steel clad fuel elements; - increase the reactivity ≅ 4 x 6.5 = 26 cents; - The removed 4 Al-clad F. E. (a little burned up) put in the core periphery, ring F, replacing graphite elements; - add exc ) of the core was estimated from control rod worth critical positions at low power, and from the calibration control rod curves. The excess reactivity of the core was increased by: re-shuffling 4 Al-clad fuel elements from C ring to F ring, replacing graphite elements

  1. Plans for an on-line computer at the Finnish TRIGA Laboratory

    International Nuclear Information System (INIS)

    Junttila, J.; Tamminen, A.; Palmgren, A.

    1970-01-01

    Full text: The future plans at the Helsinki Triga reactor include an on-line computer system. The project is in its pre-planning stage, and if the financing can be arranged, the instancing is preliminarily scheduled to start in 1972. We estimate the probability of getting a computer to be 90%. The tasks of the system would be the monitoring and control of reactor power level and other operating conditions, and, on the other hand, data acquisition, analysis and apparatus control in different reactor laboratory experiments. These include the neutron diffractometer, the cold neutron facility and neutron TOF-measurements, gamma-spectroscopy, reactor noise measurements, and neutron activation analysis. It is our intention to purchase a medium-sized computer which is presently being developed by the Finnish company Oy Stroemberg Ab. This computer has been constructed by using the time sharing principle. Consequently, background jobs like testing of new programs, off-line calculations etc., can be executed simultaneously with the actual data taking and control operations. The interfering of programs with each other is prevented by dividing the programs in protected segments. Different kinds of devices and functions can be added to the basic control processing unit, e.g. an automatic interrupt unit with up to 128 priority levels, a fast floating point arithmetic unit, and the control of a direct memory access channel. An additional fact, which makes this particular computer attractive to us, is that we wish to have the manufacturer's support in designing our special applications. The initial system cost of about 120.000 US$ is considered justified for three reasons of about equal importance: 1. The operation and utilization of our reactor will become more efficient. 2. All people involved in the project will learn how to handle a rather complicated-computerization project. 3. The Finnish Government can support the development of this computer by ordering one or two

  2. Operation experience with the 3 MW TRIGA Mark-II research reactor of Bangladesh

    International Nuclear Information System (INIS)

    Islam, M.S.; Haque, M.M.; Salam, M.A.; Rahman, M.M.; Khandokar, M.R.I.; Sardar, M.A.; Saha, P.K.; Haque, A.; Malek Sonar, M.A.; Uddin, M.M.; Hossain, S.M.S.; Zulquarnain, M.A.

    2004-01-01

    The 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since September 14, 1986. The reactor is used for radioisotope production ( 131 I, 99m Tc, 46 Sc), various R and D activities and manpower training. The reactor has been operated successfully since it's commissioning with the exception of a few reportable incidents. Of these, the decay tank leakage incident of 1997 is considered to be the most significant one. As a result of this incident, reactor operation at full power under forced-convection mode remained suspended for about 4 years. During that time, the reactor was operated at a power level of 250 kW so as to carry out experiments that require lower neutron flux. This was made possible by establishing a temporary by pass connection across the decay tank using local technology. The other incident was the contamination of the Dry Central Thimble (DCT) that took place in March 2002 when a pyrex vial containing 50 g of TeO 2 powder got melted inside the DCT. The vial was melted due to high heat generation on its surface while the reactor was operated for 8 hours at 3 MW for trial production of Iodine-131 ( 131 I). A Wet Central Thimble (WCT) was used to replace the damaged DCT in June 2002 such that the reactor operation could be resumed. The WCT was again replaced by a new DCT in June 2003 such that radioisotope production could be continued. A total of 873 irradiation requests (IRs) have been catered for different reactor uses. Out of these, 114 IRs were for radioisotope (RI) production and 759 IRs for different experiments. The total amount of RI produced stands at about 2100 GBq. The total amount of burn-up-fuel is about 6158 MWh. Efforts are on to undertake an ADP project so as to convert the analog console and I and C system of the reactor into digital one. The paper summarizes the reactor operation experiences focusing on troubleshooting, rectification, modification, RI production, various R and D

  3. On-site emergency intervention plan for nuclear accident situation at INR-Pitesti TRIGA reactor

    International Nuclear Information System (INIS)

    Oprea, I.; Margenu, S.; Preda, M.

    2001-01-01

    evaluate alternative countermeasure strategies in the various stage of an accident. In this way, we have developed at INC Pitesti a computer code namely DOZIM which is presented. The structure of the emergency intervention plan, as objectives, general features and procedures is given. Each year there is organized at INR Pitesti an emergency intervention exercise for a nuclear accident situation at the TRIGA reactor. The main objective of such an exercise is to test the response capability in a nuclear accident situation for all the institutions and organizations from all levels as well as to test the implementation of specific procedures

  4. Use of the TRIGA Reactor by the Radiochemistry Group of the Atominstitute of the Austrian Universities

    International Nuclear Information System (INIS)

    Bichler, M.; Buchtela, K.; Grass, F.; Ismail, S.

    1994-01-01

    The Radiochemistry Group of the Atominstitute of the Austrian Universities uses the TRIGA Mark II Reactor mainly for neutron activation analysis. Transport of samples to and from the irradiation positions in the reactor is performed by fast pneumatic transfer systems (transfer time 20 msec and 300 msec) and slow conventional transport facilities. Gamma-spectrometric instrumentation equipped with loss free counting systems is used to handle the high count rates up to 500 000 counts/sec. During the last years neutron activation analysis was applied to investigate environmental samples (soil, dust, incineration ash), geological samples (rocks, sediments, fossils, volcanic gases), biological materials (lichens, mushrooms and other plant materials, human diet, biological reference materials), raw materials (phosphate, coal) and archaeological materials (ancient glass). Lichen analysis was used for environmental monitoring. The content of some of the trace elements can be correlated with industrial activities, like manganese content with steel industry, the occurrence of vanadium and nickel with oil firing plants and stainless steel industry, selenium is found in lichen near coal firing plants. The amount of chlorine and sodium indicates the application of salt for road treatment during winter time, aluminum, scandium and hafnium content depends on the amount of dust in the environment. A further environmental application of neutron activation analysis is the determination of trace elements in volcanic gases. The halogens, arsenic, antimony, selenium, tellurium and mercury were determined and their daily output was calculated. The distribution of trace elements in fossils of known age gives us a geochemical key to condition and development of the paleo-environment. For this purpose we determined rare earth elements in 250 million years old microfossils (conodonts). Neutron activation analysis served also for some non scientific but nevertheless useful purposes: Organic

  5. Behavior of exposed human lymphocytes to a neutron beam of the reactor TRIGA Mark III

    International Nuclear Information System (INIS)

    Carbajal R, M. I.

    2012-01-01

    Excessive exposure to ionizing radiation occurs in people who require radiation treatment, also in those for work can come to receive doses above the permitted levels. A third possibility of exposure is the release of radioactive material in which the general population is affected. Most of the time the exhibition is partial and only rarely occurs throughout the body. For various reasons, situations arise where it is impossible to determine by conventional physical methods, the amount of radiation you were exposed to the affected person and in these cases where the option to follow is the Biological Dosimetry, where the analysis of chromosomes dicentrics is used to estimate the dose of ionizing radiation exposure. A calibration curve is generated from in vitro analysis of dicentric chromosome, which are found in human lymphocytes, treated with different types and doses of radiation. The dicentric is formed from two lesions, one on each chromosome and their union results in a structure having two centromeres, acentric fragment with her for the union of several chromosomes leads to more complex structures as tri-centric s, tetra or penta-centric s, which have the same origin. The dose-response curve is estimated by observing the frequency of dicentrics and extrapolated to a dose-effect curve previously established, for which it is necessary that each lab has its own calibration curves, taking into account that for a Let low radiation, dose-effect curve follows a linear-quadratic model Y=C + αD + βD. The production of dicentric chromosomes with a high Let, was studied using a beam of neutrons generated in the reactor TRIGA Mark III with an average energy of 1 MeV, adjusting the linear model Y=αD. The dose-response relationship is established in blood samples from the same donor, the coefficient α of the dose-response is Y = (0.3692 ± 0.011 * D), also shows that saturation is reached in system 4 Gy. (Author)

  6. Boron Neutron Capture Therapy at the TRIGA Mark II of Pavia, Italy - The BNCT of the diffuse tumours

    International Nuclear Information System (INIS)

    Altieri, S.; Bortolussi, S.; Stella, S.; Bruschi, P.; Gadan, M.A.

    2008-01-01

    The selectivity based on the B distribution rather than on the irradiation field makes Boron neutron Capture Therapy (BNCT) a valid option for the treatment of the disseminated tumours. As the range of the high LET particles is shorter than a cell diameter, the normal cells around the tumour are not damaged by the reactions occurring in the tumoral cells. PAVIA 2001: first treatment of multiple hepatic metastases from colon ca by BNCT and auto-transplantation technique: TAOrMINA project. The liver was extracted after BPA infusion, irradiated in the Thermal Column of the Pavia TRIGA Mark II reactor, and re-implanted in the patient. Two patients were treated, demonstrating the feasibility of the therapy and the efficacy in destroying the tumoral nodules sparing the healthy tissues. In the last years, the possibility of applying BNCT to the lung tumours using epithermal collimated neutron beams and without explanting the organ, is being explored. The principal obtained results of the BNCT research are presented, with particular emphasis on the following aspects: a) the project of a new thermal column configuration to make the thermal neutron flux more uniform inside the explanted liver, b) the Monte Carlo study by means of the MCNP code of the thermal neutron flux distribution inside a patient's thorax irradiated with epithermal neutrons, and c) the measurement of the boron concentration in tissues by (n,α) spectroscopy and neutron autoradiography. The dose distribution in the thorax are simulated using MCNP and the anthropomorphic model ADAM. To have a good thermal flux distribution inside the lung epithermal neutrons must be used, which thermalize crossing the first tissue layers. Thermal neutrons do not penetrate and the obtained uniformity is poor. In the future, the construction of a PGNAA facility using a horizontal channel of the TRIGA Mark II is planned. With this method the B concentration can be measured also in liquid samples (blood, urine) and in those

  7. Computer aided design (CAD) for electronics improvement of the nuclear channels of TRIGA Mark III reactor of the ININ; Diseno asistido por computadora (DAC) para mejorar la electronica de los canales nucleares del reactor TRIGA Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J.L.; Rivero G, T.; Aguilar H, F. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: jlgm@nuclear.inin.mx

    2007-07-01

    The 4 neutron measurement channels of the digital control console (CCD) of the TRIGA Mark III reactor (RTMIII) of the ININ, its were designed and built with the corresponding Quality Guarantee program, being achieved the one licensing to replace the old console. With the time they were carried out some changes to improve and to not solve some problems detected in the tests, verification and validation, requiring to modify the circuits originally designed. In this work the corrective actions carried out to eliminate the Non Conformity generated by these problems, being mentioned the advantages of using modern tools, as the software applied to the Attended Engineering by Computer, and those obtained results are presented. (Author)

  8. Experimental measurement of the refrigerant temperature of the TRIGA Mark III reactor of the ININ; Medicion experimental de la temperatura del refrigerante del reactor TRIGA Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Gallardo S, L.F.; Alonso V, G

    1991-08-15

    With the object of knowing the axial temperature profile of the refrigerant in the core of the TRIGA Mark III reactor of the ININ, the temperatures of this, at the enter, in the center and the exit of the core were measured, in the positions: west 2, north 2 and south 1. This was made by means of the thermo pars introduction mounted in aluminum guides, connected to a measurer of digital temperature, whose resolution is of {+-} 0.1 C. The measurements showed a bigger heating of the refrigerant in the superior half of the core, that which suggests that the axial profile of temperature of the reactor is not symmetrical with respect to the center or that those temperature measurements in the center are not correct. (Author)

  9. Main configurations of the reactor core TRIGA Mark III of the ININ, during their operation; Principales configuraciones del nucleo del reactor TRIGA Mark III del ININ, durante su operacion

    Energy Technology Data Exchange (ETDEWEB)

    Nava S, W.; Raya A, R., E-mail: Wenceslao.nava@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    The Reactor TRIGA Mark III is 43 years old since was put lay critical on November 8 of 1968 for the first time, along their operative life there have been 18 different configurations of the core, being three those more important: the first configuration with elements standard with an enrichment lightly minor than 20% in U-235, the second configuration that deserves out attention is when a mixed core was charged, composite of two different fuels as for their enrichment, the core consisted of 26 fuel elements Flip (of high enrichment approximately of 70%) more 3 control bars with follower of fuel Flip and 59 standard fuel elements, as those mentioned previously, finally is necessary to consider the recent reload of the reactor, with a compound core by fuel elements of low enrichment LEU 30/20. In this work the characteristics more important of the reactor are presented as well as of each one of the described cores. (Author)

  10. Development of a software for the control of the quality management system of the TRIGA-Mark III reactor; Desarrollo de un software para el control del sistema de gestion de calidad del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Herrera A, E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Hernandez, L.V.; Hernandez, J.A. [UAEM, Depto. de Ingenieria en Computacion, 50000 Toluca, estado de Mexico (Mexico)]. e-mail: eha@nuclear.inin.mx

    2006-07-01

    The quality has not only become one of the essential requirements of the product but rather at the presenme it is a strategic factor key of which depends the bigger part of the organizations, not only to maintain their position in the market but also to assure their survival. The good organizations will have processes, procedures and standards to confront these challenges. The big organizations require of the certification of their administration systems, and once the organization has obtained this certification the following step it is to maintain it. The implementation and certification of an administration system requires of an appropriate operative organization that achieves continuous improvements in their operation. This is the case of the TRIGA Mark III reactor, which contains a computer program that upgrades, it controls and it programs activities to develop in the Installation, allowing one operative organization to the whole personnel of the same one. With the purpose of avoiding activities untimely. (Author)

  11. Justine user`s manual

    Energy Technology Data Exchange (ETDEWEB)

    Lee, S.R.

    1995-10-01

    Justine is the graphical user interface to the Los Alamos Radiation Modeling Interactive Environment (LARAMIE). It provides LARAMIE customers with a powerful, robust, easy-to-use, WYSIWYG interface that facilitates geometry construction and problem specification. It is assumed that the reader is familiar with LARAMIE, and the transport codes available, i.e., MCNPTM and DANTSYSTM. No attempt is made in this manual to describe these codes in detail. Information about LARAMIE, DANTSYS, and MCNP are available elsewhere. It i also assumed that the reader is familiar with the Unix operating system and with Motif widgets and their look and feel. However, a brief description of Motif and how one interacts with it can be found in Appendix A.

  12. Phosphated in aluminium 6061-T651 used in the pool of the TRIGA Mark III nuclear reactor

    International Nuclear Information System (INIS)

    Aguilar H, F.; Espinosa L, J.; Pena B, A.; Perez F, C.; Sanchez C, M.; Vite T, M.; Vite T, J.

    2001-01-01

    We characterized a phosphated coating used a substrate in aluminium 6061-T651, which is used in the container of the TRIGA Mark III nuclear reactor. Characterization was made using MEB and X-ray diffraction techniques. Coating application has the function to prevent the corrosion. Coating was probed to test adhesion in accordance with the Standard ASTM D-4541, and the corrosion process using a salt spray (fog) camera, in accordance with the Standard ASTM B-117, so as we could phosphate the welding cord. These experiences were obtained using a Deep cell. Results obtained are going to phosphate 'in situ' using a mobile device which was patented for the National Institute of Nuclear Research (ININ) in the Mexican Institute of Intellectual Property (INPI). (Author)

  13. An epithermal irradiation terminal project for the IPR-R1 Triga Mark I reactor, CDTN/CNEN

    International Nuclear Information System (INIS)

    Oliveira, Paulo Fernando; Menezes, Maria Angela de Barros Correia; Maretti Junior, Fausto; Leal, Alexandre Soares

    2009-01-01

    The IPR-R1 Triga Mark I is a research reactor operating since 1960. It has being used mainly for training neutron activation analysis and production of some special radioisotopes. In the last years, it is coming up the necessity of using a thermal neutrons filter during neutron activation. It is in order to solve many specific situations where only activation by fast and epithermal neutrons is required. For instance, the labeling of some special molecules used for pharmaceutical investigations and the activation of biological samples in which thermal sodium activation may cause undesired analysis interferences. The usual procedure used in such cases - to irradiate the samples in the rotary specimen rack inside a cadmium box with a 1 mm wall thickness - normally offers radiological risks due to the high exposure dose. The aim of the project presented here is to optimize the procedures when the epithermal irradiation is needed. (author)

  14. Computer aided design (CAD) for electronics improvement of the nuclear channels of TRIGA Mark III reactor of the ININ

    International Nuclear Information System (INIS)

    Gonzalez M, J.L.; Rivero G, T.; Aguilar H, F.

    2007-01-01

    The 4 neutron measurement channels of the digital control console (CCD) of the TRIGA Mark III reactor (RTMIII) of the ININ, its were designed and built with the corresponding Quality Guarantee program, being achieved the one licensing to replace the old console. With the time they were carried out some changes to improve and to not solve some problems detected in the tests, verification and validation, requiring to modify the circuits originally designed. In this work the corrective actions carried out to eliminate the Non Conformity generated by these problems, being mentioned the advantages of using modern tools, as the software applied to the Attended Engineering by Computer, and those obtained results are presented. (Author)

  15. Validating the Serpent Model of FiR 1 Triga Mk-II Reactor by Means of Reactor Dosimetry

    Science.gov (United States)

    Viitanen, Tuomas; Leppänen, Jaakko

    2016-02-01

    A model of the FiR 1 Triga Mk-II reactor has been previously generated for the Serpent Monte Carlo reactor physics and burnup calculation code. In the current article, this model is validated by comparing the predicted reaction rates of nickel and manganese at 9 different positions in the reactor to measurements. In addition, track-length estimators are implemented in Serpent 2.1.18 to increase its performance in dosimetry calculations. The usage of the track-length estimators is found to decrease the reaction rate calculation times by a factor of 7-8 compared to the standard estimator type in Serpent, the collision estimators. The differences in the reaction rates between the calculation and the measurement are below 20%.

  16. Neutron flux measurement in the thermal column of the Malaysian TRIGA mark II reactor with MCNP verification

    International Nuclear Information System (INIS)

    Abdel Munem, E.; Shukri, A.; Tajuddin, A.A.

    2006-01-01

    A study of the thermal column of the Malaysian TRIGA Mark II reactor, forming part of a feasibility study for BNCT was proposed in 2001. In the current study, pure metals were used to measure the neutron flux at selected points in the thermal column and the neutron flux determined using SAND-II. Monte Carlo simulation of the thermal column was also carried out. The reactor core was homogenized and calculations of the neutron flux through the graphite stringers performed using MCNP5. The results show good agreement between the measured flux and the MCNP calculated flux. An obvious extension from this is that the MCNP neutron flux output can be utilized as an input spectrum for SAND-II for the flux iteration. (author)

  17. Determination of the fluence profile in three dimension for the thermal column of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    Herrera A, E.; Urena N, F.; Delfin L, A.; Garcia M, T.

    2006-01-01

    In this work the results of the dosimetric properties of the lithium carbonate are presented (detecting), before the thermal neutrons. The process consists on irradiating samples of lithium carbonate in the installation of the thermal column of the TRIGA Mark III reactor, with a controlled period and with time intervals of 20 hours of irradiation. It is necessary to mention that the detectors were placed in different internal positions of the thermal column. With the purpose of being used these results for future studies, like it is the fluence profile in the thermal column. To use the BNCT technique (Boron Neutron Capture Therapy). Which is a binary technique that requires the simultaneous presence of a neutron flux with appropriate energy and a neutron captor (10B), those which interacting to attack to the tumor cells without producing significant damage to the tissues when both agents are separated. (Author)

  18. Thermal neutron flux measurements in the rotary specimen rack of the IPR-R1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Rose Mary G. do Prado; Rodrigues, Rogério R.; Souza, Luiz Claudio A., E-mail: souzarm@cdtn.br, E-mail: rrr@cdtn.br, E-mail: lcas@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2017-07-01

    The thermal neutron flux in the rotary specimen rack of the IPR-R1 TRIGA reactor at the Nuclear Technology Development Center (CDTN), Belo Horizonte, Brazil, has been measured by the neutron activation method, using bare and cadmium covered gold foils. Those foils were irradiated in the rotary specimen rack with the reactor at 100 kW. The reactor core configuration has 63 fuel elements, composed of 59 original aluminum-clad elements and 4 stainless steel-clad fuel elements. The gamma activities of the foils were measured using Ge spectrometer. The perturbations of the thermal neutron flux caused by the introduction of an absorbing foil into the medium were considered in order to obtain accurate determination of the flux. The thermal neutron flux obtained was 7.4 x 10{sup 11} n.cm{sup -2}.s{sup -1}. (author)

  19. Verification of Monte Carlo calculations of the neutron flux in the carousel channels of the TRIGA Mark II reactor, Ljubljana

    International Nuclear Information System (INIS)

    Jacimovic, R.; Maucec, M.; Trkov, A.

    2002-01-01

    In this work experimental verification of Monte Carlo neutron flux calculations in the carousel facility (CF) of the 250 kW TRIGA Mark II reactor at the Jozef Stefan Institute is presented. Simulations were carried out using the Monte Carlo radiation-transport code, MCNP4B. The objective of the work was to model and verify experimentally the azimuthal variation of neutron flux in the CF for core No. 176, set up in April 2002. '1'9'8Au activities of Al-Au(0.1%) disks irradiated in 11 channels of the CF covering 180'0 around the perimeter of the core were measured. The comparison between MCNP calculation and measurement shows relatively good agreement and demonstrates the overall accuracy with which the detailed spectral characteristics can be predicted by calculations.(author)

  20. Neutron measurement at the thermal column of the Malaysian Triga Mark II reactor using gold foil activation method and TLD

    Science.gov (United States)

    Shalbi, Safwan; Salleh, Wan Norhayati Wan; Mohamad Idris, Faridah; Aliff Ashraff Rosdi, Muhammad; Syahir Sarkawi, Muhammad; Liyana Jamsari, Nur; Nasir, Nur Aishah Mohd

    2018-01-01

    In order to design facilities for boron neutron capture therapy (BNCT), the neutron measurement must be considered to obtain the optimal design of BNCT facility such as collimator and shielding. The previous feasibility study showed that the thermal column could generate higher thermal neutrons yield for BNCT application at the TRIGA MARK II reactor. Currently, the facility for BNCT are planned to be developed at thermal column. Thus, the main objective was focused on the thermal neutron and epithermal neutron flux measurement at the thermal column. In this measurement, pure gold and cadmium were used as a filter to obtain the thermal and epithermal neutron fluxes from inside and outside of the thermal column door of the 200kW reactor power using a gold foil activation method. The results were compared with neutron fluxes using TLD 600 and TLD 700. The outcome of this work will become the benchmark for the design of BNCT collimator and the shielding

  1. Development of a software for the control of the quality management system of the TRIGA-Mark III reactor

    International Nuclear Information System (INIS)

    Herrera A, E.; Hernandez, L.V.; Hernandez, J.A.

    2006-01-01

    The quality has not only become one of the essential requirements of the product but rather at the present time it is a strategic factor key of which depends the bigger part of the organizations, not only to maintain their position in the market but also to assure their survival. The good organizations will have processes, procedures and standards to confront these challenges. The big organizations require of the certification of their administration systems, and once the organization has obtained this certification the following step it is to maintain it. The implementation and certification of an administration system requires of an appropriate operative organization that achieves continuous improvements in their operation. This is the case of the TRIGA Mark III reactor, which contains a computer program that upgrades, it controls and it programs activities to develop in the Installation, allowing one operative organization to the whole personnel of the same one. With the purpose of avoiding activities untimely. (Author)

  2. A study on source term assessment and waste disposal requirement of decontamination and decommissioning for the TRIGA research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Whang, Joo Ho; Lee, Kyung JIn; Lee, Jae Min; Choi, Gyu Seup; Shin, Byoung Sun [Kyunghee Univ., Seoul (Korea, Republic of)

    1999-08-15

    The objective and necessity of the project : TRIGA is the first nuclear facility that decide to decommission and decontamination in our nation. As we estimate the expected life of nuclear power generation at 30 or 40 years, the decommissioning business should be conducted around 2010, and the development of regulatory technique supporting it should be developed previously. From a view of decommissioning and decontamination, the research reactor is just small in scale but it include all decommissioning and decontamination conditions. So, the rules by regulatory authority with decommissioning will be a guide for nuclear power plant in the future. The basis of regulatory technique required when decommissioning the research reactor are the radiological safety security and the data for it. The source term is very important condition not only for security of worker but for evaluating how we dispose the waste is appropriate for conducting the middle store and the procedure after it when the final disposal is considered. The content and the scope in this report contain the procedure of conducting the assessment of the source term which is most important in understanding the general concept of the decommissioning procedure of the decommissioning and decontamination of TRIGA research reactor. That is, the sampling and measuring method is presented as how to measure the volume of the radioactivity of the nuclear facilities. And also, the criterion of classifying the waste occurred in other countries and the site release criteria which is the final step of decommissioning and decontamination presented through MARSSIM. Finally, the program to be applicable through comparing the methods of our nation and other countries ones is presented as plan for disposal of the waste in the decommissioning.

  3. Thermal power calibrations of the IPR-R1 TRIGA reactor by the calorimetric and the heat balance methods

    International Nuclear Information System (INIS)

    Mesquita, Amir Zacarias; Rezende, Hugo Cesar; Souza, Rose Mary Gomes do Prado

    2009-01-01

    Since the first nuclear reactor was built, a number of methodological variations have been evolved for the calibration of the reactor thermal power. Power monitoring of reactors is done by means of neutronic instruments, but its calibration is always done by thermal procedures. The purpose of this paper is to present the results of the thermal power calibration carried out on March 5th, 2009 in the IPR-R1 TRIGA reactor. It was used two procedures: the calorimetric and heat balance methods. The calorimetric procedure was done with the reactor operating at a constant power, with primary cooling system switched off. The rate of temperature rise of the water was recorded. The reactor power is calculate as a function of the temperature-rise rate and the system heat capacity constant. The heat balance procedure consists in the steady-state energy balance of the primary cooling loop of the reactor. For this balance, the inlet and outlet temperatures and the water flow in the primary cooling loop were measured. The heat transferred through the primary loop was added to the heat leakage from the reactor pool. The calorimetric method calibration presented a large uncertainty. The main source of error was the determination of the heat content of the system, due to a large uncertainty in the volume of the water in the system and a lack of homogenization of the water temperature. The heat balance calibration in the primary loop is the standard procedure for calibrating the power of the IPR-R1 TRIGA nuclear reactor. (author))

  4. Evaluation of the aptitude for the service of the pool of the TRIGA Mark III reactor of the National Institute of Nuclear Research of Mexico; Evaluacion de la aptitud para el servicio de la piscina del reactor TRIGA Mark III del Instituto Nacional de Investigaciones Nucleares de Mexico

    Energy Technology Data Exchange (ETDEWEB)

    Merino C, J.; Gachuz M, M.; Diaz S, A.; Arganis J, C.; Gonzalez R, C.; Nava G, T.; Medina R, M.J. [Departamento de Sintesis y Caracterizacion de Materiales del ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)

    2001-07-01

    This work describes the evaluation of the structural integrity of the pool of the TRIGA Mark III reactor of the National Institute of Nuclear Research of Mexico, which was realized in July 2001, as an element to determine those actions for preventive and corrective maintenance which owner must do it for a safety and efficient operation of the component in the next years. (Author)

  5. Current utilization and long term strategy of the Finnish TRIGA research reactor FiR 1

    International Nuclear Information System (INIS)

    Auterinen, Iiro; Salmenhaara, Seppo

    2008-01-01

    FiR 1 (TRIGA Mark II, 250 kW) has an important international role in the development of boron neutron capture therapy (BNCT) for cancer. The safety and efficacy of BNCT is studied for several different cancers: - primary glioblastoma, a highly malignant brain tumour (since 1999); - recurrent glioblastoma or anaplastic astrocytoma (since 2001); - recurrent inoperable head and neck carcinoma (since 2003). It is one of the few facilities in the world providing this kind of treatments. The successes in the BNCT development have now created a demand for these treatments, although they are given on an experimental basis. Well over 100 patients treated now since May 1999: - at least 1 patient irradiation / week, often 2 (Tuesday and Thursday) - patients are referred to BNCT-treatments from several hospitals, also outside research protocols; - the hospitals pay for the treatment. The FiR 1 reactor has proven to be a reliable neutron source for the BNCT treatments; no patient irradiations have been cancelled because of a failure of the reactor. The BNCT facility has become a center of extensive academic research especially in medical physics. Nuclear education and training continue to play also a role at FiR 1 in the form of university courses and training of nuclear industry personnel. FiR 1 is one of the two sources in Scandinavia for short lived radioisotopes used in tracer studies in industry. The main isotope produced is Br-82 in the form of either KBr or ethylene bromide. Other typical isotopes are Na-24, Ar-41, La-140. The isotopes are used mainly in tracer studies in industry (Indmeas Inc., Finland). Typical activity of one irradiated Br-sample is 20 - 80 GBq; total activity produced in one year is over 3 TBq; the reactor operating time needed for the isotope production is one or two days per week. Accelerator based neutron sources are developed for BNCT. The prospect is that when BNCT will achieve a status of a fully accepted and efficient treatment modality for

  6. Compliance with federal and state regulations regarding the emergency response plan and physical security plan at the Oregon State TRIGA reactor

    International Nuclear Information System (INIS)

    Johnson, A.G.; Ringle, J.C.; Anderson, T.V.

    1976-01-01

    Recent legislative actions within the State of Oregon have had a significant impact upon the OSU TRIGA Emergency Response Plan, and to a lesser extent upon the Physical Security Plan. These state imposed changes will be reviewed in light of existing federal requirements. With the upcoming acquisition of FLIP fuel in August 1976, NRC required several major changes to the existing Physical Security Plan. Within the limitations of public disclosure, these changes will be contrasted to the present plan. (author)

  7. Neutronic study of nuclear reactors. Complete calculation of TRIGA MARKII reactor and calculations of fuel temperature coefficients. (Qualification of WIMS code)

    International Nuclear Information System (INIS)

    Benmansour, L.

    1992-01-01

    The present work shows a group of results, obtained by a neutronic study, concerning the TRIGA MARK II reactor and LIGHT WATER reactors. These studies aim to make cell and diffusion calculations. WIMS D-4 with extended library and DIXY programs are used and tested for those purposes. We also have proceeded to a qualification of WIMS code based on the fuel temperature coefficient calculations. 33 refs.; 23 figs.; 30 tabs. (author)

  8. Generation of seven group cross section library for TRIGA LEU fuel in CITATION format and benchmarking some experimental and operational data

    International Nuclear Information System (INIS)

    Sarker, M.M.; Bhuiyan, S.I.; Akramuzzaman, M.

    2007-01-01

    The principal objective of this study is to validate the seven group cross section library in CITATION format for TRIGA LEU Fuel. This presentation deals with the 'generation of a cross section library for the CITATION and its validation. We used WIMSD-5B version for the generation of all group constants. The overall strategy is: (1) use WIMS package to generate few group neutron macroscopic cross section (cell constants) for all of the materials in the core and its immediate neighborhood (2) use 3-D code CITATION to perform the global analysis of the core to study: multiplication factor, neutron flux distribution and power peaking factors. Various options available in WIMS program were studied in depth to finalize the models to generate the most appropriate group constants. For the global analysis the code CITATION and a post processing program FCAP were chosen. Thus a seven group cross section library for the calculations of TRIGA Research Reactor was generated. To investigate the validity of the generated library a critical experiment of the TRIGA research reactor was benchmarked. (author)

  9. Characterization of the neutron flux in the Hohlraum of the thermal column of the TRIGA Mark III reactor of the ININ; Caracterizacion del flujo neutronico en el Hohlraum de la columna termica del reactor TRIGA Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Delfin L, A.; Palacios, J.C.; Alonso, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: adl@nuclear.inin.mx

    2006-07-01

    Knowing the magnitude of the neutron flux in the reactor irradiation facilities, is so much importance for the operation of the same one, like for the investigation developing. Particularly, knowing with certain precision the spectrum and the neutron flux in the different positions of irradiation of a reactor, it is essential for the evaluation of the results obtained for a certain irradiation experiment. The TRIGA Mark III reactor account with irradiation facilities designed to carry out experimentation, where the reactor is used like an intense neutron source and gamma radiation, what allows to make irradiations of samples or equipment in radiation fields with components and diverse levels in the different facilities, one of these irradiation facilities is the Thermal Column where the Hohlraum is. In this work it was carried out a characterization of the neutron flux inside the 'Hohlraum' of the irradiation facility Thermal Column of the TRIGA Mark III reactor of the Nuclear Center of Mexico to 1 MW of power. It was determined the sub cadmic neutron flux and the epi cadmic by means of the neutron activation technique of thin sheets of gold. The maps of the distribution of the neutron flux for both energy groups in three different positions inside the 'Hohlraum' are presented, these maps were obtained by means of the irradiation of undressed thin activation sheets of gold and covered with cadmium in arrangements of 10 x 12, located parallel to 11.5 cm, 40.5 cm and 70.5 cm to the internal wall of graphite of the installation in inverse address to the position of the reactor core. Starting from the obtained values of neutron flux it was found that, for the same position of the surface of irradiation of the experimental arrangement, the relative differences among the values of neutron flux can be of 80%, and that the differences among different positions of the irradiation surfaces can vary until in a one order of magnitude. (Author)

  10. Modeling user navigation

    NARCIS (Netherlands)

    Herder, E.; Brusilovsky, Peter; Corbett, Albert; de Rosis, Fiorella

    2003-01-01

    For providing users with navigation aids that best serve their needs, user models for adaptive hypermedia should include user navigation patterns. This paper describes elements needed and how these elements can be gathered.

  11. Investigation of a superthermal ultracold neutron source based on a solid deuterium converter for the TRIGA Mainz reactor

    International Nuclear Information System (INIS)

    Lauer, Thorsten

    2010-01-01

    Research in fundamental physics with the free neutron is one of the key tools for testing the Standard Model at low energies. Most prominent goals in this field are the search for a neutron electric dipole moment (EDM) and the measurement of the neutron lifetime. Significant improvements of the experimental performance using ultracold neutrons (UCN) require reduction of both systematic and statistical errors.The development and construction of new UCN sources based on the superthermal concept is therefore an important step for the success of future fundamental physics with ultracold neutrons. Significant enhancement of today available UCN densities strongly correlates with an efficient use of an UCN converter material. The UCN converter here is to be understood as a medium which reduces the velocity of cold neutrons (CN, velocity of about 600 m/s) to the velocity of UCN (velocity of about 6 m/s).Several big research centers around the world are presently planning or constructing new superthermal UCN sources, which are mainly based on the use of either solid deuterium or superfluid helium as UCN converter.Thanks to the idea of Yu.Pokotilovsky, there exists the opportunity to build competitive UCN sources also at small research reactors of the TRIGA type. Of course these smaller facilities don't promise high UCN densities of several 1000 UCN/cm 3 , but they are able to provide densities around 100 UCN/cm 3 for experiments.In the context of this thesis, it was possible to demonstrate succesfully the feasibility of a superthermal UCN source at the tangential beamport C of the research reactor TRIGA Mainz. Based on a prototype for the future UCN source at the Forschungs-Neutronenquelle Heinz Maier-Leibnitz (FRMII) in Munich, which was planned and built in collaboration with the Technical University of Munich, further investigations and improvements were done and are presented in this thesis. In parallel, a second UCN source for the radial beamport D was designed and

  12. Development and validation of a model TRIGA Mark III reactor with code MCNP5; Desarrollo y validacion de un modelo del reactor Triga Mark III con el codigo MCNP5

    Energy Technology Data Exchange (ETDEWEB)

    Galicia A, J.; Francois L, J. L. [UNAM, Facultad de Ingenieria, Departamento de Sistemas Energeticos, Ciudad Universitaria, 04510 Ciudad de Mexico (Mexico); Aguilar H, F., E-mail: blink19871@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2015-09-15

    The main purpose of this paper is to obtain a model of the reactor core TRIGA Mark III that accurately represents the real operating conditions to 1 M Wth, using the Monte Carlo code MCNP5. To provide a more detailed analysis, different models of the reactor core were realized by simulating the control rods extracted and inserted in conditions in cold (293 K) also including an analysis for shutdown margin, so that satisfied the Operation Technical Specifications. The position they must have the control rods to reach a power equal to 1 M Wth, were obtained from practice entitled Operation in Manual Mode performed at Instituto Nacional de Investigaciones Nucleares (ININ). Later, the behavior of the K{sub eff} was analyzed considering different temperatures in the fuel elements, achieving calculate subsequently the values that best represent the actual reactor operation. Finally, the calculations in the developed model for to obtain the distribution of average flow of thermal, epithermal and fast neutrons in the six new experimental facilities are presented. (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. User Innovation Management

    DEFF Research Database (Denmark)

    Kanstrup, Anne Marie; Bertelsen, Pernille

    User Innovation Management (UIM) is a method for fo-opereation with users in innovation projects. The UIM method emphasizes the practice of a participatorty attitude.......User Innovation Management (UIM) is a method for fo-opereation with users in innovation projects. The UIM method emphasizes the practice of a participatorty attitude....

  15. User Behavior Analytics

    Energy Technology Data Exchange (ETDEWEB)

    Turcotte, Melissa [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Moore, Juston Shane [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-02-28

    User Behaviour Analytics is the tracking, collecting and assessing of user data and activities. The goal is to detect misuse of user credentials by developing models for the normal behaviour of user credentials within a computer network and detect outliers with respect to their baseline.

  16. The User Experience

    Science.gov (United States)

    Schmidt, Aaron

    2010-01-01

    User experience (UX) is about arranging the elements of a product or service to optimize how people will interact with it. In this article, the author talks about the importance of user experience and discusses the design of user experiences in libraries. He first looks at what UX is. Then he describes three kinds of user experience design: (1)…

  17. Lazy User Behaviour

    OpenAIRE

    Collan, Mikael

    2007-01-01

    In this position paper we suggest that a user will most often choose the solution (device) that will fulfill her (information) needs with the least effort. We call this “lazy user behavior”. We suggest that the principle components responsible for solution selection are the user need and the user state. User need is the user’s detailed (information) need (urgency, type, depth, etc.) and user state is the situation, in which the user is at the moment of the need (location, time, etc.); the use...

  18. Neutron activation analysis at CDTN/CNEN using the IPR-R1 Triga Mark I reactor

    International Nuclear Information System (INIS)

    Menezes, Maria Angela de B.C.; Maretti Junior, Fausto; Kastner, Geraldo Frederico; Amaral, Angela Maria; Souza, Wagner de

    2009-01-01

    This paper describes in summary the activities developed by the Laboratory for Neutron Activation Analysis since the starting up of the IPR-R1 TRIGA Mark I research reactor in 1960. This Laboratory is located at Centro de Desenvolvimento da Tecnologia Nuclear (Nuclear Technology Development Centre) / Comissao Nacional de Energia Nuclear (Brazilian Commission for Nuclear Energy), CDTN/CNEN. The activities of the Laboratory comprise the delayed fission neutron activation analysis, instrumental (comparative and parametric methods) and radiochemical / chemical methods. These methods are responsible for significant percentage of CDTN's analytical demand, meeting the clients' analytical needs and researches developed by the Laboratory, by CDTN and by other institutions. Over the years the work has been linked to the goals of the country and the institutions. Nowadays the neutron activation analysis is responsible for 70% of the analytical demand and the k 0 - Instrumental method for 80% of this demand answering clients' request and researches. In Brazil, CDTN is the only Institute that fully masters the Instrumental Neutron Activation Analysis k 0 -method using its own nuclear reactor. (author)

  19. An investigation into the feasibility of thorium fuels utilization in seed-blanket configurations for TRIGA PUSPATI Reactor (RTP)

    Science.gov (United States)

    Damahuri, Abdul Hannan Bin; Mohamed, Hassan; Aziz Mohamed, Abdul; Idris, Faridah

    2018-01-01

    Thorium is one of the elements that needs to be explored for nuclear fuel research and development. One of the popular core configurations of thorium fuel is seed-blanket configuration or also known as Radkowsky Thorium Fuel concept. The seed will act as a supplier of neutrons, which will be placed inside of the core. The blanket, on the other hand, is the consumer of neutrons that is located at outermost of the core. In this work, a neutronic analysis of seed-blanket configuration for the TRIGA PUSPATI Reactor (RTP) is carried out using Monte Carlo method. The reactor, which has been operated since 1982 use uranium zirconium hydride (U-ZrH1.6) as the fuel and have multiple uranium weight which are 8.5, 12 and 20 wt.%. The pool type reactor is one and only research reactor that located in Malaysia. The design of core included the Uranium Zirconium Hydride located at the centre of the core that will act as the seed to supply neutron. The thorium oxide that will act as blanket situated outside of seed region will receive neutron to transmute 232Th to 233U. The neutron multiplication factor or criticality of each configuration is estimated. Results show that the highest initial criticality achieved is 1.30153.

  20. Determination of the subcadmium flux in the Triga Mark III reactor of the Nuclear Center in Mexico

    International Nuclear Information System (INIS)

    Rodiguez V, F.J.

    1975-01-01

    The determination of the subcadmium flux profile (neutrons with energy between 0 and 0.4 eV approximatley) in the core of the reactor Triga Mark III in the Nuclear Center of Mexico was made. The technique used for that purpose consists in placing alternativelly indium sheets covered and uncovered with cadmium and separated by fragments of aluminium 2.54cm wide in an aluminium tube whose external diameter reaches 0.635cm. This tube is introduced in each of the 21 axial irradiation places of the sheets the activity of gamma rays of 1.293 MeV of In-116m produced during the irradiation, is measured using a monochannel analyzer. We obtain graphs of the specific counting ratio vs position for the sheets uncovered and covered with cadmium; using quadratic interpolation we compute in each position the missing specific measuring ratio. The difference between the specific counting ratios with our cadmium and with cadmium multiplied by a constant gives the subcadmium flux in the z point. For irradiation at 10 watts, 1 Kilowatt and 1 megawatt the results show profiles which in general terms are in accordance with those which were obtained theoretically for a cylyndrical reactor using group theory. The greatest flow given by the manufacturer of the reactor is in the values range obtained experimentally.For the calculations, the method uses a FORTRAN IV program so that the determinations can be made routinely. (author)

  1. Design of a digital system for operational parameters simulation of IPR-R1 TRIGA nuclear research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lage, Aldo M.F.; Mesquita, Amir Z.; Felippe, Adriano de A.M., E-mail: aldo@cdtn.br, E-mail: amir@cdtn.br, E-mail: adrianoamfelippe@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN /CNEN-MG), Belo Horizonte, MG (Brazil); Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

    2017-11-01

    The instrumentation of nuclear reactors is designed based on the reliability, redundancy and diversification of control systems. The monitoring of the parameters is of crucial importance with regard to the operational efficiency and safety of the installation. Since the first criticality of a nuclear reactor, achieved by Fermi et al. in 1942, there has been concern about the reliable monitoring of the parameters involved in the chain reaction. This paper presents the current stage of the system of simulation, which is under development at the CDTN, which intends to simulate the operation of the TRIGA IPR-R1 nuclear reactor, involving the evolution of neutron flux and reactor power related events. The system will be developed using LabVIEW® software, using the modern concept of virtual instruments (VIs) that are visualized in a video monitor. For the implementation of this model, computational tools and systems analysis are necessary, which help and facilitate the implementation of the simulator. In this article we will show some of these techniques and the initial design of the model to be implemented. The design of a computational system is of great importance, since it guides in the implementation stages and generates the documentation for later maintenance and updating of the computational system. It is noteworthy that the innovations developed in research reactors are normally used in power reactors. The relatively low costs enable research reactors to be an excellent laboratory for developing techniques for future reactors. (author)

  2. Analysis of occupational doses from the CDTN reactor TRIGA IPR-R1 for radiation protection optimization

    Energy Technology Data Exchange (ETDEWEB)

    Maletta, Paulo G.M.; Leal, Marcos A.; Wakabayashi, Tetsuaki, E-mail: pgmm@cdtn.br, E-mail: amgr@cdtn.br, E-mail: leal@cdtn.br, E-mail: tw@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Guimaraes, Adriana M.; Silva, Teogenes A. da, E-mail: silvata@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Post Graduation Programme in Sciences and Technology of Radiations

    2011-07-01

    Worker and area monitoring have routinely been done around the CDTN TRIGA IPR-R1 research reactor aiming to optimize and assure radiation protection safety. As part of the implementation of the ALARA program, individual doses from planned practices were analyzed. Personnel dose equivalents, Hp(10), from up to 39 occupationally exposed workers were daily reported during their stay in the restricted area. Measurements were done with a RAD-60 Rados electronic personal dosimeter with a 1 {mu}Sv low detection limit. Results of about 5000 measurements in a year obtained during 2009 and 2010 showed that monthly doses did not exceed 60 {mu}Sv, except in very specific cases of non-routine practices. Results also suggested that values of 40, 100 and 800 {mu}Sv could be adopted as weekly, monthly and annual dose constraints. Considering that measured doses were very small when compared to the 20 mSv/year dose limit, it was concluded that the adoption of the dose constraints was enough to assume the compliance with the ALARA principle and that changes in the routine procedure or in the reactor facility design are not needed. (author)

  3. Design of a digital system for operational parameters simulation of IPR-R1 TRIGA nuclear research reactor

    International Nuclear Information System (INIS)

    Lage, Aldo M.F.; Mesquita, Amir Z.; Felippe, Adriano de A.M.

    2017-01-01

    The instrumentation of nuclear reactors is designed based on the reliability, redundancy and diversification of control systems. The monitoring of the parameters is of crucial importance with regard to the operational efficiency and safety of the installation. Since the first criticality of a nuclear reactor, achieved by Fermi et al. in 1942, there has been concern about the reliable monitoring of the parameters involved in the chain reaction. This paper presents the current stage of the system of simulation, which is under development at the CDTN, which intends to simulate the operation of the TRIGA IPR-R1 nuclear reactor, involving the evolution of neutron flux and reactor power related events. The system will be developed using LabVIEW® software, using the modern concept of virtual instruments (VIs) that are visualized in a video monitor. For the implementation of this model, computational tools and systems analysis are necessary, which help and facilitate the implementation of the simulator. In this article we will show some of these techniques and the initial design of the model to be implemented. The design of a computational system is of great importance, since it guides in the implementation stages and generates the documentation for later maintenance and updating of the computational system. It is noteworthy that the innovations developed in research reactors are normally used in power reactors. The relatively low costs enable research reactors to be an excellent laboratory for developing techniques for future reactors. (author)

  4. Calculations to Support On-line Neutron Spectrum Adjustment by Measurements with Miniature Fission Chambers in the JSI TRIGA Reactor

    Science.gov (United States)

    Kaiba, Tanja; Radulović, Vladimir; Žerovnik, Gašper; Snoj, Luka; Fourmentel, Damien; Barbot, LoÏc; Destouches, Christophe AE(; )

    2018-01-01

    Preliminary calculations were performed with the aim to establish optimal experimental conditions for the measurement campaign within the collaboration between the Jožef Stefan Institute (JSI) and Commissariat à l'Énergie Atomique et aux Énergies Alternatives (CEA Cadarache). The goal of the project is to additionally characterize the neutron spectruminside the JSI TRIGA reactor core with focus on the measurement epi-thermal and fast part of the spectrum. Measurements will be performed with fission chambers containing different fissile materials (235U, 237Np and 242Pu) covered with thermal neutron filters (Cd and Gd). The changes in the detected signal and neutron flux spectrum with and without transmission filter were studied. Additional effort was put into evaluation of the effect of the filter geometry (e.g. opening on the top end of the filter) on the detector signal. After the analysis of the scoping calculations it was concluded to position the experiment in the outside core ring inside one of the empty fuel element positions.

  5. Comparison of EPR response of alanine and Gd₂O₃-alanine dosimeters exposed to TRIGA Mainz reactor.

    Science.gov (United States)

    Marrale, M; Schmitz, T; Gallo, S; Hampel, G; Longo, A; Panzeca, S; Tranchina, L

    2015-12-01

    In this work we report some preliminary results regarding the analysis of electron paramagnetic resonance (EPR) response of alanine pellets and alanine pellets added with gadolinium used for dosimetry at the TRIGA research reactor in Mainz, Germany. Two set-ups were evaluated: irradiation inside PMMA phantom and irradiation inside boric acid phantom. We observed that the presence of Gd2O3 inside alanine pellets increases the EPR signal by a factor of 3.45 and 1.24 in case of PMMA and boric acid phantoms, respectively. We can conclude that in the case of neutron beam with a predominant thermal neutron component the addition of gadolinium oxide can significantly improve neutron sensitivity of alanine pellets. Monte Carlo (MC) simulations of both response of alanine and Gd-added alanine pellets with FLUKA code were performed and a good agreement was achieved for pure alanine dosimeters. For Gd2O3-alanine deviations between MC simulations and experimental data were observed and discussed. Copyright © 2015 Elsevier Ltd. All rights reserved.

  6. Calculations to Support On-line Neutron Spectrum Adjustment by Measurements with Miniature Fission Chambers in the JSI TRIGA Reactor

    Directory of Open Access Journals (Sweden)

    Kaiba Tanja

    2018-01-01

    Full Text Available Preliminary calculations were performed with the aim to establish optimal experimental conditions for the measurement campaign within the collaboration between the Jožef Stefan Institute (JSI and Commissariat à l’Énergie Atomique et aux Énergies Alternatives (CEA Cadarache. The goal of the project is to additionally characterize the neutron spectruminside the JSI TRIGA reactor core with focus on the measurement epi-thermal and fast part of the spectrum. Measurements will be performed with fission chambers containing different fissile materials (235U, 237Np and 242Pu covered with thermal neutron filters (Cd and Gd. The changes in the detected signal and neutron flux spectrum with and without transmission filter were studied. Additional effort was put into evaluation of the effect of the filter geometry (e.g. opening on the top end of the filter on the detector signal. After the analysis of the scoping calculations it was concluded to position the experiment in the outside core ring inside one of the empty fuel element positions.

  7. PIE of a CANDU fuel element irradiated for a load following test in the INR TRIGA reactor

    International Nuclear Information System (INIS)

    Parvan, Marcel; Matei, Ana; Mincu, Marin; Uta, Octavian; Palleck, Steve; Montin, John; Abbas, Shaun

    2007-01-01

    As part of the collaboration under the Romania - Canada Memorandum for co-operation in research and development of nuclear energy and technology, a load following test has been devised to demonstrate the load following capability of CANDU-6 fuel within the established design envelope for operating powers. A 37-element CANDU-6 fuel bundle element fabricated by AECL was irradiated in the TRIGA 14 MW(th) material testing reactor at the Institute for Nuclear Research (INR) in Pitesti, Romania. The load following cycle consisted of 200 daily cycles from 100% power to 50% power within the reference overpower envelope for fuel in a CANDU-6 reactor. Full power operation was 57 kW/m Element Linear Power. The paper provides the results obtained by post-irradiation examination of the fuel element in the INR hot cells. The following techniques were used: - Visual inspection and photography by periscope; - Profilometry; - Axial gamma scanning; - Fuel element puncturing and fission gas analysis; - Metallographic and ceramographic examinations by optical microscopy; - Burn-up measurement by mass spectrometry using the 235 U depletion method. (authors)

  8. Determination of the neutrons energy spectrum in the central thimble of the reactor core TRIGA Mark III

    International Nuclear Information System (INIS)

    Parra M, M. A.; Luis L, M. A.; Raya A, R.; Cruz G, H. S.

    2013-10-01

    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 -10 to 20 MeV) of the neutrons spectrum in the reactor core, for this purpose thin sheets were used of 197 Au, 58 Ni, 115 In, 24 Mg, 27 Al, 58 Fe, 59 Co and 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)

  9. Training courses for the staff of the nuclear power station KRSKO conducted at the TRIGA reactor center in Ljubljana

    International Nuclear Information System (INIS)

    Pregl, G.; Najzer, M.

    1976-01-01

    The training program for the Nuclear Power Station Krsko was divided into two modules: fundamentals of nuclear engineering and specialized training according to duties that candidates are supposed to take at the power station. Basic training was organized at the TRIGA Reactor Center in Ljubljana in two different versions. The first version intended for plant operators and all engineers lasted for six months and included about 500 hours of classroom lessons and seminars and 31 laboratory experiments. The educational program was conventional. The following topics were covered: nuclear and atomic physics, reactor theory, reactor dynamics, reactor instrumentation and control, heat transfer in nuclear power plants, nuclear power plant systems, reactor materials, reactor safety, and radiation protection. Until now, two groups, consisting of 37 candidates altogether, have attended this basic course. Plans have been made to conduct two additional courses of about 20 students each for technicians other than operators. The program of this second version will be reduced, with the emphasis on reactor core physics and radiation protection. Classroom lessons will be strongly supported by laboratory experiments. (author)

  10. Neutron flux and gamma dose measurement in the BNCT irradiation facility at the TRIGA reactor of the University of Pavia

    Science.gov (United States)

    Bortolussi, S.; Protti, N.; Ferrari, M.; Postuma, I.; Fatemi, S.; Prata, M.; Ballarini, F.; Carante, M. P.; Farias, R.; González, S. J.; Marrale, M.; Gallo, S.; Bartolotta, A.; Iacoviello, G.; Nigg, D.; Altieri, S.

    2018-01-01

    University of Pavia is equipped with a TRIGA Mark II research nuclear reactor, operating at a maximum steady state power of 250 kW. It has been used for many years to support Boron Neutron Capture Therapy (BNCT) research. An irradiation facility was constructed inside the thermal column of the reactor to produce a sufficient thermal neutron flux with low epithermal and fast neutron components, and low gamma dose. In this irradiation position, the liver of two patients affected by hepatic metastases from colon carcinoma were irradiated after borated drug administration. The facility is currently used for cell cultures and small animal irradiation. Measurements campaigns have been carried out, aimed at characterizing the neutron spectrum and the gamma dose component. The neutron spectrum has been measured by means of multifoil neutron activation spectrometry and a least squares unfolding algorithm; gamma dose was measured using alanine dosimeters. Results show that in a reference position the thermal neutron flux is (1.20 ± 0.03) ×1010 cm-2 s-1 when the reactor is working at the maximum power of 250 kW, with the epithermal and fast components, respectively, 2 and 3 orders of magnitude lower than the thermal component. The ratio of the gamma dose with respect to the thermal neutron fluence is 1.2 ×10-13 Gy/(n/cm2).

  11. Inspection with non destructive assay techniques of the aluminium coating of the TRIGA Mark III reactor vat

    International Nuclear Information System (INIS)

    Reyes A, A.I.; Gonzalez M, A.; Castaneda J, G.; Rivera M, H.; Sandoval G, I.

    2001-01-01

    In June 2000, the Reactor Department assigned to the Scientific Research Direction of the National Institute of Nuclear Research requested to the Non-destructive Assays Laboratory (LEND), assigned to the Materials Science Management, the inspection and measurement of thickness of the aluminium coating (liner) of the TRIGA Mark III reactor vat with non-destructive assay techniques, due to that the aluminium coating is exposed mainly to undergo slimming on its back side due to corrosion phenomena. Activity that was able to be carried out from april until august 2001. It is worth pointing out that this type of inspection with these techniques was realized by first time. The non-destructive assays (NDA) are techniques which use indirect physical methods for inspecting the sanitation of components in process or in service, for detect lack of continuity or defects which affect their quality or usefulness. The application of those do not alter the physical, chemical, mechanical or dimensional properties of the part subject of inspection. The results of the application of the ultrasound inspection techniques, industrial radiography and penetrating liquids are presented. (Author)

  12. Computational Analysis of Nuclear Safety Parameters of 3 MW TRIGA Mark-II Research Reactor Based on Evaluated Nuclear Data Libraries JENDL-3.3 and ENDF/B-VII.0

    International Nuclear Information System (INIS)

    Khan, Jahirul Haque

    2013-01-01

    The objective of this study is to explain the main nuclear safety parameters of 3 MW TRIGA Mark-II Research Reactor at AERE, Savar, Dhaka, Bangladesh from the viewpoint of reactor safety and also reactor operator. The most important nuclear reactor physics safety parameters are power distribution, power peaking factors, shutdown margin, control rod worth, excess reactivity and fuel temperature reactivity coefficient. These parameters are calculated using the chain of the computer codes the SRAC-PIJ for cell calculation based on neutron transport theory and the SRAC-CITATION for core calculation based on neutron diffusion equation. To achieve this objective the TRIGA model is developed by the 3-D diffusion code SRAC-CITATION based on the group constants that come from the collision probability transport code SRAC-PIJ. In this study the evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 are used. The calculated most important reactor physics parameters are compared to the safety analysis report (SAR) values as well as earlier published MCNP results (numerically benchmark). It was found that the calculated results show a good agreement between the said libraries. Besides, in most cases the calculated results reveal a reasonable agreement with the SAR values (by General Atomic) as well as the MCNP results. In addition, this analysis can be used as the inputs for thermal-hydraulic calculations of the TRIGA fresh core in the steady state and pulse mode operation. Because of power peaking factors, power distributions and temperature reactivity coefficients are the most important reactor safety parameters for normal operation and transient safety analysis in research as well as in power reactors. They form the basis for technical specifications and limitations for reactor operation such as loading pattern limitations for pulse operation (in TRIGA). Therefore, this analysis will be very important to develop the nuclear safety parameters data of 3 MW TRIGA Mark

  13. A user-friendly, digital console for the control room parameters supervision in old-generation nuclear plants

    International Nuclear Information System (INIS)

    Memmi, Fabrizio; Falconi, Luca; Cappelli, Mauro; Palomba, Mario; Sepielli, Massimo

    2016-01-01

    Highlights: • We propose a user-friendly monitoring system for reactor supervision. • Statistics from data analysis can be used to optimize reactor management. • The tool has been designed in order to include a simulation tool for prediction. • The proposed system could help operators in training and continuous learning. - Abstract: In this work a user-friendly, digital monitoring system for supervision of process variables coming from a fission nuclear reactor of TRIGA type (1-MW TRIGA reactor RC-1) is presented. The system, developed on the basis of COTS tools, can easily interface the control room instrumentation and display the typical monitoring parameters (e.g. nuclear power, temperatures, flow rates, radiological variables) in an intuitive, user-adjustable way for plant operators. A front panel of a virtual instrument allows for a direct measure while the acquisition system, for signals coming from the reactor, can process the data and generate a detailed representation of the results, whose statistics can be interpreted to optimize the reactor management parameters. This system has been also designed so as to include a simulation tool able to predict specific performances and investigate critical phenomena, and to optimize overall plant performances. In particular, it allows to have a feedback control and to perform several predictive statistical surveys of all main process parameters. The proposed system could help operators in training and continuous learning activities, and serve as a basis for an advanced decision support system and for a remote training tool for students and trainees not authorized to work in a radiation environment.

  14. The construction of TRIGA-TRAP and direct high-precision Penning trap mass measurements on rare-earth elements and americium

    International Nuclear Information System (INIS)

    Ketelaer, Jens

    2010-01-01

    The construction of TRIGA-TRAP and direct high-precision Penning trap mass measurements on rare-earth elements and americium: Nuclear masses are an important quantity to study nuclear structure since they reflect the sum of all nucleonic interactions. Many experimental possibilities exist to precisely measure masses, out of which the Penning trap is the tool to reach the highest precision. Moreover, absolute mass measurements can be performed using carbon, the atomic-mass standard, as a reference. The new double-Penning trap mass spectrometer TRIGA-TRAP has been installed and commissioned within this thesis work, which is the very first experimental setup of this kind located at a nuclear reactor. New technical developments have been carried out such as a reliable non-resonant laser ablation ion source for the production of carbon cluster ions and are still continued, like a non-destructive ion detection technique for single-ion measurements. Neutron-rich fission products will be available by the reactor that are important for nuclear astrophysics, especially the r-process. Prior to the on-line coupling to the reactor, TRIGA-TRAP already performed off-line mass measurements on stable and long-lived isotopes and will continue this program. The main focus within this thesis was on certain rare-earth nuclides in the well-established region of deformation around N ∝ 90. Another field of interest are mass measurements on actinoids to test mass models and to provide direct links to the mass standard. Within this thesis, the mass of 241 Am could be measured directly for the first time. (orig.)

  15. The construction of TRIGA-TRAP and direct high-precision Penning trap mass measurements on rare-earth elements and americium

    Energy Technology Data Exchange (ETDEWEB)

    Ketelaer, Jens

    2010-06-14

    The construction of TRIGA-TRAP and direct high-precision Penning trap mass measurements on rare-earth elements and americium: Nuclear masses are an important quantity to study nuclear structure since they reflect the sum of all nucleonic interactions. Many experimental possibilities exist to precisely measure masses, out of which the Penning trap is the tool to reach the highest precision. Moreover, absolute mass measurements can be performed using carbon, the atomic-mass standard, as a reference. The new double-Penning trap mass spectrometer TRIGA-TRAP has been installed and commissioned within this thesis work, which is the very first experimental setup of this kind located at a nuclear reactor. New technical developments have been carried out such as a reliable non-resonant laser ablation ion source for the production of carbon cluster ions and are still continued, like a non-destructive ion detection technique for single-ion measurements. Neutron-rich fission products will be available by the reactor that are important for nuclear astrophysics, especially the r-process. Prior to the on-line coupling to the reactor, TRIGA-TRAP already performed off-line mass measurements on stable and long-lived isotopes and will continue this program. The main focus within this thesis was on certain rare-earth nuclides in the well-established region of deformation around N {proportional_to} 90. Another field of interest are mass measurements on actinoids to test mass models and to provide direct links to the mass standard. Within this thesis, the mass of {sup 241}Am could be measured directly for the first time. (orig.)

  16. Accelerator facilities users' guide

    International Nuclear Information System (INIS)

    Walter, H.C.; Adrion, L.; Frosch, R.; Salzmann, M.

    1994-07-01

    In 1981 the ''Green Book'' of SIN was distributed, a User Handbook serving the needs of people already working at SIN as well as informing new users about our installations. An update of the Green Book is necessary because many beams have disappeared, been modified or added, and the installation has been upgraded in intensity and versatility quite considerably. The spectrum of users has shifted away from nuclear and particle physics; applications in medicine, solid state physics and materials science have gained in importance. This Users' Guide is intended to inform our users about the changes, and to interest potential new users in coming to PSI. (author) figs., tabs

  17. Calculation of the Activity Inventory for the TRIGA Reactor at the Medical University of Hannover (MHH) in Preparation for Dismantling the Facility

    Energy Technology Data Exchange (ETDEWEB)

    Hampel, G.; Scheller, F.; Bernnat, W.; Pfister, G.; Klaux, U.; Gerhards, E.

    2002-02-25

    It is planned to dismantle the TRIGA reactor facility at the Medical University of Hannover (MHH). Radioactive waste resulting from this dismantling will be disposed of externally, any remaining materials as well as the building structures will then be measured to ensure there is no residual activity. In preparation for this and to plan the techniques which will be used to dismantle the reactor, calculations were made in order to determine the amount of activity and the dose rates for the reactor tank and its inside components as well as for the biological shield and its radial beam tube.

  18. Safety-evaluation report related to the renewal of the operating license for the Cornell University TRIGA Research Reactor. Docket No. 50-157

    International Nuclear Information System (INIS)

    1983-08-01

    This Safety Evaluation Report for the application filed by the Cornell University for a renewal of Operating License R-80 to continue to operate a research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by Cornell University and is located on the Cornell campus in Ithaca, New York. The staff concludes that the TRIGA reactor facility can continue to be operated by Cornell without endangering the health and safety of the public

  19. Safety evaluation report related to the renewal of the operating license for the Washington State University TRIGA reactor. Docket No. 50-27

    International Nuclear Information System (INIS)

    1982-05-01

    This Safety Evaluation Report for the application filed by the Washington State University (WSU) for a renewal of operating license number R-76 to continue to operate a research reactor has been prepared by the Office of Nuclear Reactor Regulation of the US Nuclear Regulatory Commission. The facility is owned and operated by the Washington State University and is located on the WSU campus in Pullman, Whitman County, Washington. The staff concludes that the TRIGA reactor facility can continue to be operated by WSU without endangering the health and safety of the public

  20. Elaboration of the configuration and programming of the interlocks system of the TRIGA Mark III reactor; Elaboracion de la configuracion y programacion del sistema de interbloqueos del reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Mejia C, M. A.

    2016-07-01

    The modernization of the TRIGA Mark III reactor interlock system requires a system that provides high reliability, flexibility and ease of operation during reactor operation. With this modernization of the system, is intended to prevent, control and mitigate the causes of probable accidents reported in the reactor accident analysis. On the other hand, is foreseen the ease reactor operation in a simple, safe and efficient way. The programmable logic controller can be programmed by programming instructions using simple language and easy to develop, these can be modified from a computer using the programming software. In addition, another of the advantages offered by the controller is that can be modified from a touch screen (human-machine interface) that allows adjustment, without the need to use programming software and diagnostic functions during the process. As a result of the present work, a situation of improvement in the reactor operation was generated, facilitating the handling of the bridge and increasing the efficiency of the system in the execution of the operating conditions of the installations external to the reactor. A modern, more reliable and much less expensive system was achieved than the previous one, avoiding that the maintenance to the system generates high expenses. With respect to the development of the application programming, a control was implemented that allows to select a zone of the five that have inside the pool to carry out the displacement of automatic way and later to be located in that zone, having in this way a greater efficiency and ease in bridge control. (Author)

  1. Refurbishment, Modernization and Ageing Management Program of The 3MW TRIGA Mark-II Research Reactor of Bangladesh

    International Nuclear Information System (INIS)

    Salam, M. A.

    2013-01-01

    The 3 MW TRIGA MK-II research reactor of Bangladesh Atomic Energy Commission (BAEC) achieved its first criticality on 14 September 1986. The reactor has been used for manpower training, radioisotope production and various R and D activities in the field of neutron activation analysis, neutron radiography and neutron scattering. Reactor Operation and Maintenance Unit (ROMU) is responsible for operation and maintenance of the research reactor. During the past twenty seven years ROMU carried out several refurbishments, replacement, modification and modernization activities in the reactor facility. The major tasks carried out under refurbishment program were replacement of the corrosion damaged N-16 decay tank by a new one, replacement of the fouled shell and tube type heat exchanger by a plate type one, modification of the shielding arrangements around the N-16 decay tank and ECCS system and solving the radial beam port-1 leakage problem. All of these refurbishment activities were performed under an annual development project (ADP) funded by Bangladesh government. BAEC research reactor (RR) was operated by analogue console system from its commissioning to July, 2011. Old analog based console has been replaced by digital console on June, 2012. Modernization program for the reactor control console due to obsolescence and unavailability of spare parts of I and C system was vital to restore the safe operation of the reactor. Considering these facts, installation of a digital control console and I and C system based on the state-of-the-art digital technology became necessary. Reactor digital console system installation tasks were performed under another ADP funded project by Bangladesh government. Now the reactor is operating with the digital control system. Besides this, the Neutron Radiography (NR) facility has been modernized by the addition of a digital neutron radiography set-up at the tangential beam port. The Neutron Scattering (NS) facility also has been upgraded

  2. Getting grid users together

    CERN Multimedia

    Appleton, Owen

    2007-01-01

    "While Grid conferences are becoming ever more popular, many of them remain primarily IT events, with few if any users attending. Not so the second EGEE User Forum, an event specifically designed to bring together the diverse user community that makes use of the EGEE grid infrastructure." (1 page)

  3. NPAS Users Guide

    International Nuclear Information System (INIS)

    1984-01-01

    This NPAS Users Guide is primarily intended as a source of information about policies, procedures, and facilities appropriate for users in the program of Nuclear Physics at SLAC (NPAS). General policies and practices are described, the preparation of proposals is discussed, and the services for users is outlined. SLAC experimental facilities are described, and contacts are listed

  4. Determination of the flows profile in the role of power in the central thimble of TRIGA Mark III Reactor

    International Nuclear Information System (INIS)

    Garcia F, A.

    2010-01-01

    The overall objective of the thesis project is to determine the flow profiles sub cadmic and epi cadmic in the central thimble to different powers and operation times of TRIGA Mark III Reactor, using activation foils as detectors. In the reactor operation, it is necessary to know the neutron flow profile for to realize other tasks as: the radioisotopes production, research in reactors physics and fuel burning. The distribution of the neutron flow, accurately reflects what is happening in the reactor core, plus the flows value in this distribution is directly related to the power generated. For this reason it is performed the sub cadmic flow measurement with energies between 0 and 0.4 eV (energy of the cadmium cut E cd ∼ 0.4 eV) and epi cadmic flow with energies greater than 0.4 eV, in the central thimble powers to the powers of 10, 100 W, 1, 10 100 Kw and 1 MW. The method used is known as flakes activation, which is to be arranged by placing flakes ( 3 mm of diameter and 0.0508 mm of thickness) of a given material (either Au, In, Cu, Mn, etc.) into an aluminum tube outside diameter equal to 6.35 mm, alternating flakes with lids covered and discovered of cadmium (3.4 mm of diameter and 0.508 mm of thickness) and separated by lucite pieces of 3 mm of diameter and 25.4 mm in length. After irradiating the flakes for some time, is measured the gamma activity of each of them, using a hyper pure germanium detector of high resolution. Already known gamma activity, proceed to calculate the epi cadmic and sub cadmic flows using a computer program in Fortran language, called Caflu. (Author)

  5. Operation and maintenance of the 250 kW TRIGA Mark II reactor at the J. Stefan Institute

    International Nuclear Information System (INIS)

    Dimic, V.

    1980-01-01

    Since the last TRIGA Conference, the reactor has completed approximately 4800 operating hours without major problems. The problem with the lack of fresh fuel elements is going to be solved after the signing of a new agreement for the supply of fuel between the IAEA, the Yugoslav and US Governments. In order to increase the reactivity the fuel elements from the outer zone we shuffled to the inner zone, and old fuel elements from the fuel container were added to the F ring. Due to the large demand for irradiation, a new pneumatic facility for loading and unloading the samples in the rotary specimen rack or central thimble has been constructed and installed. A configuration based on two microcomputers in a master slave hierarchical organisation for automatic data logging and direct has been finished and the system was installed after extensive testing. The reactor operation is now more reliable and simpler for the operators. Some of the original instrumentation of the reactor has been gradually substituted because of ageing: a start-up channel with digital display, a power integrator, a digital electronic rod position indicator, a digital power range switch without resistors and a new 2-pen recorder have been installed. The following instrumentation was ordered by the IAEA from the Hartmann and Brawn company: a start-up channel, a log channel, a safety channel, an automatic power control and water temperature, conductivity, level and activity measuring units. During the last year, with the help of our nuclear chemistry department, the production of high concentration and high purity technetium-99 m for medical use was developed by a solvent extraction method

  6. Feasibility study of the underwater neutron radiography facility using the University of Utah 100 kW TRIGA (UUTR) reactor

    International Nuclear Information System (INIS)

    Choe, D.; Xiao, S.; Jevremovic, T.; Yang, X.

    2010-01-01

    The University of Utah 100 kW TRIGA (UUTR) reactor provides usable neutron yields for neutron radiography. Currently, UUTR reactor has three irradiators (Central, Pneumatic, and Thermal irradiators) and one Fast neutron Irradiation Facility (FNIF). These irradiators are very small so they are not suitable for neutron radiography. UUTR has three beam ports but they are not available due to the structure of the core. All sides of the core are occupied by FNIF, Thermal Irradiator, and three ion chambers. The only available position for underwater vertical beam port is on the top of the FNIF. There are two factors necessary to fulfill to be able to realize vertical underwater beam port: noninterruption to other facilities and radiation shielding. Designing the vertical beam port as movable ensures good access to the core and pool, while still providing a good neutron radiography environment. Keeping the top of the beam port below the surface of the pool the water represents biological shield. Neutron radiographs, with a simple setup of efficient neutron converters and digital camera systems, can produce acceptable resolution with an exposure time as short as a few minutes. It is important to validate the design with calculations before constructing the beam port. The design of the beam port is modeled using the MCNP5 transport code. A minimum of 10 5 neutrons/cm 2 -sec thermal neutron flux is required for high resolution neutron radiography. Currently, the UUTRIGA is in the process of upgrading its power from 100 kW to 250 kW. Upon the completion of the upgrading, the maximum neutron flux in the core will be ∼7x10 12 neutrons/cm 2 -sec. This paper discusses a modeling and evaluation of capability for a neutron radiography facility. (author)

  7. Neutron flux of 100kW in the irradiation terminals of the IPR-R1 Triga Reactor

    International Nuclear Information System (INIS)

    Zangirolami, Dante Marco

    2009-01-01

    In this work, it was carried out a study of the neutron flux in the IPR-R1 TRIGA reactor irradiation facilities: rotary specimen rack (RSR), pneumatic transfer tube two (PTT2) and the central thimble (CT). The objective was to obtain the neutron flux profile on the RSR, which has forty irradiation positions, and also values for the thermal and epithermal neutron fluxes of some RSR positions and also of the PTT2 and of the CT facility. It was applied the neutron activation analysis of a reference material, Al-Au (0.1%) alloy. Irradiations were performed on 16 different dates. It was concluded that for the RSR, the average value of thermal and epithermal neutron fluxes depends on the vertical position of the reactor control rods. Neutron flux variations along the RSR form a characteristic profile, whose values depend on the location of the irradiation position in the reactor core and on the control rods vertical position. In the RSR, the obtained values of thermal and epithermal neutron flux were (8.1 +- 0.3) x 10 11 n.cm -2 .s -1 , and (3.4 +- 0.2)x10 10 n.cm -2 .s -1 , respectively. For the PTT2 and the CT, the values for the epithermal neutron flux were respectively (3.3 +- 0.2) x 10 9 n.cm -2 .s -1 and (2.6 +- 0.1) x 10 11 n.cm -2 .s -1 . For these facilities, the thermal neutron flux was estimated, and the obtained values were (2.4 +- 0.2) x 10 11 n.cm -2 .s -1 and (2.8 +- 0.1)x10 12 n.cm -2 .s -1 for the PTT2 and the CT, respectively. (author)

  8. A high performance neutron powder diffractometer at 3 MW Triga Mark-II research reactor in Bangladesh

    Energy Technology Data Exchange (ETDEWEB)

    Kamal, I., E-mail: imtiaz-kamal26@yahoo.com; Yunus, S. M., E-mail: yunussm11@yahoo.com; Datta, T. K., E-mail: tk-datta4@yahoo.com; Zakaria, A. K. M.; Das, A. K.; Aktar, S.; Hossain, S. [Institute of Nuclear Science and Technology, Atomic Energy Research Establishment, Ganakbari, Savar, Dhaka (Bangladesh); Berliner, R., E-mail: RB@instrumentationAssociates.com [Instrumentation Associates, Durham, North Carolina (United States); Yelon, W. B., E-mail: yelonwb@hotmail.com [Oak Ridge National Laboratory, Oak Ridge, Tennessee (United States)

    2016-07-12

    A high performance neutron diffractometer called Savar Neutron Diffractometer (SAND) was built and installed at radial beam port-2 of TRIGA Mark II research reactor at AERE, Savar, Dhaka, Bangladesh. Structural studies of materials are being done by this technique to characterize materials crystallograpohically and magnetically. The micro-structural information obtainable by neutron scattering method is very essential for determining its technological applications. This technique is unique for understanding the magnetic behavior in magnetic materials. Ceramic, steel, electronic and electric industries can be benefited from this facility for improving their products and fabrication process. This instrument consists of a Popovicimonochromator with a large linear position sensitive detector array. The monochromator consists of nine blades of perfect single crystal of silicon with 6 mm thickness each. The monochromator design was optimized to provide maximum flux on 3 mm diameter cylindrical sample with a relatively flat angular dependence of resolution. Five different wave lengths can be selected by orienting the crystal at various angles. A sapphire filter was used before the primary collimator to minimize the first neutron. The detector assembly is composed of 15 linear position sensitive proportional counters placed at either 1.1 m or 1.6 m from the sample position and enclosed in a air pad supported high density polythene shield. Position sensing is obtained by charge division using 1-wide NIM position encoding modules (PEM). The PEMs communicate with the host computer via USB. The detector when placed at 1.1 m, subtends 30° (2θ) at each step and covers 120° in 4 steps. When the detector is placed at 1.6 m it subtends 20° at each step and covers 120° in 6 steps. The instrument supports both low and high temperature sample environment. The instrument supports both low and high temperature sample environment. The diffractometer is a state-of-the art technology

  9. A high performance neutron powder diffractometer at 3 MW Triga Mark-II research reactor in Bangladesh

    Science.gov (United States)

    Kamal, I.; Yunus, S. M.; Datta, T. K.; Zakaria, A. K. M.; Das, A. K.; Aktar, S.; Hossain, S.; Berliner, R.; Yelon, W. B.

    2016-07-01

    A high performance neutron diffractometer called Savar Neutron Diffractometer (SAND) was built and installed at radial beam port-2 of TRIGA Mark II research reactor at AERE, Savar, Dhaka, Bangladesh. Structural studies of materials are being done by this technique to characterize materials crystallograpohically and magnetically. The micro-structural information obtainable by neutron scattering method is very essential for determining its technological applications. This technique is unique for understanding the magnetic behavior in magnetic materials. Ceramic, steel, electronic and electric industries can be benefited from this facility for improving their products and fabrication process. This instrument consists of a Popovicimonochromator with a large linear position sensitive detector array. The monochromator consists of nine blades of perfect single crystal of silicon with 6mm thickness each. The monochromator design was optimized to provide maximum flux on 3mm diameter cylindrical sample with a relatively flat angular dependence of resolution. Five different wave lengths can be selected by orienting the crystal at various angles. A sapphire filter was used before the primary collimator to minimize the first neutron. The detector assembly is composed of 15 linear position sensitive proportional counters placed at either 1.1 m or 1.6 m from the sample position and enclosed in a air pad supported high density polythene shield. Position sensing is obtained by charge division using 1-wide NIM position encoding modules (PEM). The PEMs communicate with the host computer via USB. The detector when placed at 1.1 m, subtends 30˚ (2θ) at each step and covers 120˚ in 4 steps. When the detector is placed at 1.6 m it subtends 20˚ at each step and covers 120˚ in 6 steps. The instrument supports both low and high temperature sample environment. The instrument supports both low and high temperature sample environment. The diffractometer is a state-of-the art technology

  10. A high performance neutron powder diffractometer at 3 MW Triga Mark-II research reactor in Bangladesh

    International Nuclear Information System (INIS)

    Kamal, I.; Yunus, S. M.; Datta, T. K.; Zakaria, A. K. M.; Das, A. K.; Aktar, S.; Hossain, S.; Berliner, R.; Yelon, W. B.

    2016-01-01

    A high performance neutron diffractometer called Savar Neutron Diffractometer (SAND) was built and installed at radial beam port-2 of TRIGA Mark II research reactor at AERE, Savar, Dhaka, Bangladesh. Structural studies of materials are being done by this technique to characterize materials crystallograpohically and magnetically. The micro-structural information obtainable by neutron scattering method is very essential for determining its technological applications. This technique is unique for understanding the magnetic behavior in magnetic materials. Ceramic, steel, electronic and electric industries can be benefited from this facility for improving their products and fabrication process. This instrument consists of a Popovicimonochromator with a large linear position sensitive detector array. The monochromator consists of nine blades of perfect single crystal of silicon with 6 mm thickness each. The monochromator design was optimized to provide maximum flux on 3 mm diameter cylindrical sample with a relatively flat angular dependence of resolution. Five different wave lengths can be selected by orienting the crystal at various angles. A sapphire filter was used before the primary collimator to minimize the first neutron. The detector assembly is composed of 15 linear position sensitive proportional counters placed at either 1.1 m or 1.6 m from the sample position and enclosed in a air pad supported high density polythene shield. Position sensing is obtained by charge division using 1-wide NIM position encoding modules (PEM). The PEMs communicate with the host computer via USB. The detector when placed at 1.1 m, subtends 30° (2θ) at each step and covers 120° in 4 steps. When the detector is placed at 1.6 m it subtends 20° at each step and covers 120° in 6 steps. The instrument supports both low and high temperature sample environment. The instrument supports both low and high temperature sample environment. The diffractometer is a state-of-the art technology

  11. User interface design considerations

    DEFF Research Database (Denmark)

    Andersen, Simon Engedal; Jakobsen, Arne; Rasmussen, Bjarne D.

    1999-01-01

    When designing a user interface for a simulation model there are several important issues to consider: Who is the target user group, and which a priori information can be expected. What questions do the users want answers to and what questions are answered using a specific model?When developing...... and output variables. This feature requires special attention when designing the user interface and a special approach for controlling the user selection of input and output variables are developed. To obtain a consistent system description the different input variables are grouped corresponding...... the user interface of EESCoolTools these issues led to a series of simulation tools each with a specific purpose and a carefully selected set of input and output variables. To allow a more wide range of questions to be answered by the same model, the user can change between different sets of input...

  12. Qualification of heavy water based irradiation device in the JSI TRIGA reactor for irradiations of FT-TIMS samples for nuclear safeguards

    Science.gov (United States)

    Radulović, Vladimir; Kolšek, Aljaž; Fauré, Anne-Laure; Pottin, Anne-Claire; Pointurier, Fabien; Snoj, Luka

    2018-03-01

    The Fission Track Thermal Ionization Mass Spectrometry (FT-TIMS) method is considered as the reference method for particle analysis in the field of nuclear Safeguards for measurements of isotopic compositions (fissile material enrichment levels) in micrometer-sized uranium particles collected in nuclear facilities. An integral phase in the method is the irradiation of samples in a very well thermalized neutron spectrum. A bilateral collaboration project was carried out between the Jožef Stefan Institute (JSI, Slovenia) and the Commissariat à l'Énergie Atomique et aux Énergies Alternatives (CEA, France) to determine whether the JSI TRIGA reactor could be used for irradiations of samples for the FT-TIMS method. This paper describes Monte Carlo simulations, experimental activation measurements and test irradiations performed in the JSI TRIGA reactor, firstly to determine the feasibility, and secondly to design and qualify a purpose-built heavy water based irradiation device for FT-TIMS samples. The final device design has been shown experimentally to meet all the required performance specifications.

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

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, B.C.

    1997-05-01

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

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

    International Nuclear Information System (INIS)

    Ryan, B.C.

    1997-05-01

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

  15. Criticality and shielding calculations for containers in dry of spent fuel of TRIGA Mark III reactor of ININ; Calculos de criticidad y blindaje para contenedores en seco de combustible gastado del reactor Triga Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Barranco R, F.

    2015-07-01

    In this thesis criticality and shielding calculations to evaluate the design of a container of dry storage of spent nuclear fuel generated in research reactors were made. The design of such container was originally proposed by Argentina and Brazil, and the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico. Additionally, it is proposed to modify the design of this container to store spent fuel 120 that are currently in the pool of TRIGA Mark III reactor, the Nuclear Center of Mexico and calculations and analyzes are made to verify that the settlement of these fuel elements is subcritical limits and dose rates to workers and the general public are not exceeded. These calculations are part of the design criteria for security protection systems in dry storage system (Dss for its acronym in English) proposed by the Nuclear Regulatory Commission (NRC) of the United States. To carry out these calculations simulation codes of Monte Carlo particle transport as MCNPX and MCNP5 were used. The initial design (design 1) 78 intended to store spent fuel with a maximum of 115. The ININ has 120 fuel elements and spent 3 control rods (currently stored in the reactor pool). This leads to the construction of two containers of the original design, but for economic reasons was decided to modify (design 2) to store in a single container. Criticality calculations are performed to 78, 115 and fresh fuel elements 124 within the container, to the two arrangements described in Chapter 4, modeling the three-dimensional geometry assuming normal operating conditions and accident. These calculations are focused to demonstrate that the container will remain subcritical, that is, that the effective multiplication factor is less than 1, in particular not greater than 0.95 (as per specified by the NRC). Spent fuel 78 and 124 within the container, both gamma radiation to neutron shielding calculations for only two cases were simulated. First actinides and fission products generated

  16. International user studies

    DEFF Research Database (Denmark)

    Nielsen, Lene; Madsen, Sabine; Jensen, Iben

    in Sydhavnen, and it is funded by InfinIT. Based on a qualitative interview study with 15 user researchers from 11 different companies, we have investigated how companies collect and present data about users on international markets. Key findings are: Companies do not collect data about end users in all...... the countries/regions they operate in. Instead, they focus on a few strategic markets. International user studies tend to be large-scale studies that involve the effort of many both internal and external/local human resources. The studies typically cover 2-4 countries/regions and many end users in each country...... across nationalities and (2) that it often is more important to focus on and take differences in market conditions into account than national culture per se. Companies are in the process of finding out how best to present the insights about international end users to their employees. However, so far...

  17. User Requirements for Wireless

    DEFF Research Database (Denmark)

    In most IT system development processes, the identification or elicitation of user requirements is recognized as a key building block. In practice, the identification of user needs and wants is a challenge and inadequate or faulty identifications in this step of an IT system development can cause...... huge problems with the final product. The elicitation of user requirements as such changes according to age groups;, to gender,; to cultural settings,; and into time; and experience in the use of the system/software. User requirements, therefore, cannot be used between projects, IT systems......, and different software. That makes the elicitation of user requirements an inherent part of any software development project and a resourceful activity as well. This book provides insights to the process of identifying user requirements and to different types by describing varying case studies in which...

  18. Measuring user engagement

    CERN Document Server

    Lalmas, Mounia; Yom-Tov, Elad

    2014-01-01

    User engagement refers to the quality of the user experience that emphasizes the positive aspects of interacting with an online application and, in particular, the desire to use that application longer and repeatedly. User engagement is a key concept in the design of online applications (whether for desktop, tablet or mobile), motivated by the observation that successful applications are not just used, but are engaged with. Users invest time, attention, and emotion in their use of technology, and seek to satisfy pragmatic and hedonic needs. Measurement is critical for evaluating whether online

  19. Demonstrator 1: User Interface and User Functions

    DEFF Research Database (Denmark)

    Gram, Christian

    1999-01-01

    Describes the user interface and its functionality in a prototype system used for a virtual seminar session. The functionality is restricted to what is needed for a distributed seminar discussion among not too many people. The system is designed to work with the participants distributed at several...

  20. Criticality and shielding calculations for containers in dry of spent fuel of TRIGA Mark III reactor of ININ

    International Nuclear Information System (INIS)

    Barranco R, F.

    2015-01-01

    In this thesis criticality and shielding calculations to evaluate the design of a container of dry storage of spent nuclear fuel generated in research reactors were made. The design of such container was originally proposed by Argentina and Brazil, and the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico. Additionally, it is proposed to modify the design of this container to store spent fuel 120 that are currently in the pool of TRIGA Mark III reactor, the Nuclear Center of Mexico and calculations and analyzes are made to verify that the settlement of these fuel elements is subcritical limits and dose rates to workers and the general public are not exceeded. These calculations are part of the design criteria for security protection systems in dry storage system (Dss for its acronym in English) proposed by the Nuclear Regulatory Commission (NRC) of the United States. To carry out these calculations simulation codes of Monte Carlo particle transport as MCNPX and MCNP5 were used. The initial design (design 1) 78 intended to store spent fuel with a maximum of 115. The ININ has 120 fuel elements and spent 3 control rods (currently stored in the reactor pool). This leads to the construction of two containers of the original design, but for economic reasons was decided to modify (design 2) to store in a single container. Criticality calculations are performed to 78, 115 and fresh fuel elements 124 within the container, to the two arrangements described in Chapter 4, modeling the three-dimensional geometry assuming normal operating conditions and accident. These calculations are focused to demonstrate that the container will remain subcritical, that is, that the effective multiplication factor is less than 1, in particular not greater than 0.95 (as per specified by the NRC). Spent fuel 78 and 124 within the container, both gamma radiation to neutron shielding calculations for only two cases were simulated. First actinides and fission products generated

  1. Cognitive requirements in the redesign of a TRIGA RC-1 control room: The role of the operators' evaluations

    International Nuclear Information System (INIS)

    Visciola, M.; Bagnara, S.; Ruggeri, R.

    1986-01-01

    When a control room undergoes to a redesign process it is of crucial importance to analyze how operators critically review it and which improvements they suggest. This is even more critical when presumably the same people will operate in the new 'redesigned' control room: Consistency in the mental models possessed by the operators of the plant functions and of their control should be emphasized. Consistency in the mental models can be assumed when redesign follows well-established guidelines drawn from experiences and studies carried out in very similar situations. However, this condition is not fulfilled when a nuclear research control room has to be redesigned, since available guidelines (e.g.; NUREG-0700) are based on studies conducted on nuclear power plant control rooms. These two types of facilities are of much difference as for activities performed in the control room, goals to be aimed at, costs and risks. As a consequence, the available guidelines cannot be safely applied to such a situation as the redesign of a TRIGA RC-1 control room. So, data have to be collected in order to allow the operators to efficiently and easily adapt to the new control room by consistently 'updating' their mental models. In the present study, these data have been collected through structured interviews, which consisted of a modified version of EPRI. The results can be summarized as follows: 1) The operators critically reviewed the present control room and underlined the lack of 'transparency' of the control system as for the plant's conditions and for the feedbacks about their own activities. 2) The operators' work analysis showed that they spend much of their time out of the control room. This means that, if the operators have to stay in the control room, they should be allowed to perform more and higher-level activities than those presently required, to prevent understimulation. So, the redesign should or allow and support the central control and maintenance, and other

  2. Analysis of area events as part of probabilistic safety assessment for Romanian TRIGA SSR 14 MW reactor

    International Nuclear Information System (INIS)

    Mladin, D.; Stefan, I.

    2005-01-01

    The international experience has shown that the external events could be an important contributor to plant/ reactor risk. For this reason such events have to be included in the PSA studies. In the context of PSA for nuclear facilities, external events are defined as events originating from outside the plant, but with the potential to create an initiating event at the plant. To support plant safety assessment, PSA can be used to find methods for identification of vulnerable features of the plant and to suggest modifications in order to mitigate the impact of external events or the producing of initiating events. For that purpose, probabilistic assessment of area events concerning fire and flooding risk and impact is necessary. Due to the relatively large power level amongst research reactors, the approach to safety analysis of Romanian 14 MW TRIGA benefits from an ongoing PSA project. In this context, treatment of external events should be considered. The specific tasks proposed for the complete evaluation of area event analysis are: identify the rooms important for facility safety, determine a relative area event risk index for these rooms and a relative area event impact index if the event occurs, evaluate the rooms specific area event frequency, determine the rooms contribution to reactor hazard state frequencies, analyze power supply and room dependencies of safety components (as pumps, motor operated valves). The fire risk analysis methodology is based on Berry's method [1]. This approach provides a systematic procedure to carry out a relative index of different rooms. The factors, which affect the fire probability, are: personal presence in the room, number and type of ignition sources, type and area of combustibles, fuel available in the room, fuel location, and ventilation. The flooding risk analysis is based on the amount of piping in the room. For accuracy of the information regarding piping a facility walk-about is necessary. In case of flooding risk

  3. Neutron tomography McClellan Nuclear Radiation Center (MNRC) TRIGA reactor

    International Nuclear Information System (INIS)

    Crump, J.C. III; Shields, K.; Newell, D.; Kiger, K.; Friday, C.

    1994-01-01

    At the MNRC Reactor Facility we are currently in the process of converting one of our real-time radiography bays into a neutron tomography facility. The facility's initial use will be for the determination of the amount of hydrogen in titanium compressor blades. An increase in the amount of hydrogen within the blade can result in embriddlement and later structural failure of the fan blade. The blades are 'baked' (heated to high temperatures to.force the hydrogen from the blade) after a number of flying hours. The goal of the system is to enable the end user to know when blades have reached potentially damaging levels of hydrogen (greater than 125 ppm). The system also needs to be able to process a large number of blades in a short time (i.e. a production mode). To achieve the goal of detecting the hydrogen at the appropriate levels; maxmize through put, and give the system flexibility to be used in other applications several unique features were incorporated. These design features included; a high resolution CCD camera for imaging, multiple high capacity turn tables to minimize system fixturing time, and automatic image analysis for hydrogen content. The high resolution CCD camera was selected based on previous work at the facility done by Dr. Eric McFarlan of University of California at Santa Barbara. He originally used the ST-4 and ST-6 which had small rectangular arrays of 375x242. For the system we chose the Princeton Instruments TE-1024K which has a segmentable (can be partitioned in smaller areas) array of 1024x1024. The camera also has the ability to digitize image information at 16 bit resolution or as low as 8 bit for high speed. Additionally, the CCD camera sensitivity (quantum efficiency) was matched closely to the wavelength of light emitted by the scintillation screen. The turn tables used have a weight capacity of 2000 pounds. This is well above the weight of the fan blade (less than one pound). To increase the through put we placed 3 blades on each

  4. Main activities carried out for the conversion of the reactor core TRIGA, from HEU 8.5/70 / LEU 8.5/20 to LEU 30/20; Principales actividades llevadas a cabo para la conversion del nucleo del reactor TRIGA, de HEU 8.5/70 / LEU 8.5/20 a LEU 30/20

    Energy Technology Data Exchange (ETDEWEB)

    Flores C, J., E-mail: jorge.floresc@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    In agreement with the policies of the global initiative of threats reduction (GTRI), Mexico committed that inside the reduction program of the fuel enrichment in research and test reactors (RERTR), the conversion of the core reactor TRIGA (in the nuclear centre) would be made, to use solely fuel with low enrichment ({<=} 20% U{sup 235}). To support to the execution of this commitment, a series of accords and agreements were established. The Project Agreement and Supply among the IAEA, the United States of America and Mexico was the more relevant. In this work the main activities carried out in the Instituto Nacional de Investigaciones Nucleares (ININ) with this purpose are presented. (Author)

  5. Validation study of SRAC2006 code system based on evaluated nuclear data libraries for TRIGA calculations by benchmarking integral parameters of TRX and BAPL lattices of thermal reactors

    International Nuclear Information System (INIS)

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

    2013-01-01

    Highlights: ► To validate the SRAC2006 code system for TRIGA neutronics calculations. ► TRX and BAPL lattices are treated as standard benchmarks for this purpose. ► To compare the calculated results with experiment as well as MCNP values in this study. ► The study demonstrates a good agreement with the experiment and the MCNP results. ► Thus, this analysis reflects the validation study of the SRAC2006 code system. - Abstract: The goal of this study is to present the validation study of the SRAC2006 code system based on evaluated nuclear data libraries ENDF/B-VII.0 and JENDL-3.3 for neutronics analysis of TRIGA Mark-II Research Reactor at AERE, Bangladesh. This study is achieved through the analysis of integral parameters of TRX and BAPL benchmark lattices of thermal reactors. In integral measurements, the thermal reactor lattices TRX-1, TRX-2, BAPL-UO 2 -1, BAPL-UO 2 -2 and BAPL-UO 2 -3 are treated as standard benchmarks for validating/testing the SRAC2006 code system as well as nuclear data libraries. The integral parameters of the said lattices are calculated using the collision probability transport code PIJ of the SRAC2006 code system at room temperature 20 °C based on the above libraries. The calculated integral parameters are compared to the measured values as well as the MCNP values based on the Chinese evaluated nuclear data library CENDL-3.0. It was found that in most cases, the values of integral parameters demonstrate a good agreement with the experiment and the MCNP results. In addition, the group constants in SRAC format for TRX and BAPL lattices in fast and thermal energy range respectively are compared between the above libraries and it was found that the group constants are identical with very insignificant difference. Therefore, this analysis reflects the validation study of the SRAC2006 code system based on evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 and can also be essential to implement further neutronics calculations

  6. Who are your users?

    DEFF Research Database (Denmark)

    Nielsen, Lene; Salminen, joni; Jung, Soon-Gyo

    2017-01-01

    One of the reasons for using personas is to align user understandings across project teams and sites. As part of a larger persona study, at Al Jazeera English (AJE), we conducted 16 qualitative interviews with media producers, the end users of persona descriptions. We asked the participants about...

  7. Additional user needs

    International Nuclear Information System (INIS)

    Rorschach, H.E.; Hayter, J.B.

    1986-01-01

    This paper summarizes the conclusions of a discussion group on users' needs held at the Workshop on an Advanced Steady-State Neutron Facility. The discussion was devoted to reactor characteristics, special facilities and siting considerations suggested by user needs. (orig.)

  8. Vulnerable road users.

    NARCIS (Netherlands)

    2010-01-01

    A group of road users can be defined as ‘vulnerable’ in a number of ways, such as by the amount of protection in traffic (e.g. pedestrians and cyclists) or by the amount of task capability (e.g. the young and the elderly). Vulnerable road users do not usually have a protective 'shell', and also the

  9. User Requirements for Wireless

    DEFF Research Database (Denmark)

    In most IT system development processes, the identification or elicitation of user requirements is recognized as a key building block. In practice, the identification of user needs and wants is a challenge and inadequate or faulty identifications in this step of an IT system development can cause...... involvement and requirements elicitation Usable security requirements for design of privacy...

  10. User Interface History

    DEFF Research Database (Denmark)

    Jørgensen, Anker Helms; Myers, Brad A

    2008-01-01

    User Interfaces have been around as long as computers have existed, even well before the field of Human-Computer Interaction was established. Over the years, some papers on the history of Human-Computer Interaction and User Interfaces have appeared, primarily focusing on the graphical interface era...

  11. User Frustrations as Opportunities

    Directory of Open Access Journals (Sweden)

    Michael Weiss

    2012-04-01

    Full Text Available User frustrations are an excellent source of new product ideas. Starting with this observation, this article describes an approach that entrepreneurs can use to discover business opportunities. Opportunity discovery starts with a problem that the user has, but may not be able to articulate. User-centered design techniques can help elicit those latent needs. The entrepreneur should then try to understand how users are solving their problem today, before proposing a solution that draws on the unique skills and technical capabilities available to the entrepreneur. Finally, an in-depth understanding of the user allows the entrepreneur to hone in on the points of difference and resonance that are the foundation of a strong customer value proposition.

  12. User participation in implementation

    DEFF Research Database (Denmark)

    Fleron, Benedicte; Rasmussen, Rasmus; Simonsen, Jesper

    2012-01-01

    experienced more uncertainty and frustration than management and non-participating staff, especially concerning how to run an implementation process and how to understand and utilize the configuration possibilities of the system. This suggests that user participation in implementation introduces a need......Systems development has been claimed to benefit from user participation, yet user participation in implementation activities may be more common and is a growing focus of participatory-design work. We investigate the effect of the extensive user participation in the implementation of a clinical...... system by empirically analyzing how management, participating staff, and non-participating staff view the implementation process with respect to areas that have previously been linked to user participation such as system quality, emergent interactions, and psychological buy-in. The participating staff...

  13. The PANTHER User Experience

    Energy Technology Data Exchange (ETDEWEB)

    Coram, Jamie L. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Morrow, James D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Perkins, David Nikolaus [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    This document describes the PANTHER R&D Application, a proof-of-concept user interface application developed under the PANTHER Grand Challenge LDRD. The purpose of the application is to explore interaction models for graph analytics, drive algorithmic improvements from an end-user point of view, and support demonstration of PANTHER technologies to potential customers. The R&D Application implements a graph-centric interaction model that exposes analysts to the algorithms contained within the GeoGraphy graph analytics library. Users define geospatial-temporal semantic graph queries by constructing search templates based on nodes, edges, and the constraints among them. Users then analyze the results of the queries using both geo-spatial and temporal visualizations. Development of this application has made user experience an explicit driver for project and algorithmic level decisions that will affect how analysts one day make use of PANTHER technologies.

  14. Lead User Innovation

    DEFF Research Database (Denmark)

    Brem, Alexander; Larsen, Henry

    2015-01-01

    User innovation and especially the integration of lead users is a key topic in the innovation management literature of recent years. This paper contributes by providing a rare perspective into what easily could be seen as innovation failure, shown from two perspectives. We show how a lack of shared...... imagination hampers participation and kills innovation between interdependent stakeholders at the threshold between invention and innovation in practice. We present a first case in the fun-sport industry where an external lead user and diverse firm representatives in different functions fail to create......, deliver and capture the value of an innovatively new device together. From the perspective of the lead user, we show antecedents and effects of social interaction between organizational actors and the lead user on the development of social capital, especially trust and shared imagination. The second case...

  15. User`s guide to MIDAS

    Energy Technology Data Exchange (ETDEWEB)

    Tisue, S.A.; Williams, N.B.; Huber, C.C. [Argonne National Lab., IL (United States). Decision and Information Sciences Div.; Chun, K.C. [Argonne National Lab., IL (United States). Environmental Assessment Div.

    1995-12-01

    Welcome to the MIDAS User`s Guide. This document describes the goals of the Munitions Items Disposition Action System (MIDAS) program and documents the MIDAS software. The main text first describes the equipment and software you need to run MIDAS and tells how to install and start it. It lists the contents of the database and explains how it is organized. Finally, it tells how to perform various functions, such as locating, entering, viewing, deleting, changing, transferring, and printing both textual and graphical data. Images of the actual computer screens accompany these explanations and guidelines. Appendix A contains a glossary of names for the various abbreviations, codes, and chemicals; Appendix B is a list of modem names; Appendix C provides a database dictionary and rules for entering data; and Appendix D describes procedures for troubleshooting problems associated with connecting to the MIDAS server and using MIDAS.

  16. Dose calculation in biological samples in a mixed neutron-gamma field at the TRIGA reactor of the University of Mainz

    DEFF Research Database (Denmark)

    Schmitz, T.; Blaickner, M.; Schütz, C.

    2010-01-01

    biological effectiveness (RBE) of liver and cancer cells in our mixed neutron and gamma field. We work with alanine detectors in combination with Monte Carlo simulations, where we can measure and characterize the dose. To verify our calculations we perform neutron flux measurements using gold foil activation...... and pin-diodes. Material and methods. When L-α-alanine is irradiated with ionizing radiation, it forms a stable radical which can be detected by electron spin resonance (ESR) spectroscopy. The value of the ESR signal correlates to the amount of absorbed dose. The dose for each pellet is calculated using...... of the mixed neutron and gamma field of the TRIGA Mainz is possible in order to characterize the neutron behavior in the thermal column. Currently we also speculate on sensitizing alanine to thermal neutrons by adding boron compounds....

  17. The User Reconfigured

    DEFF Research Database (Denmark)

    Bardzell, Jeffrey; Bardzell, Shaowen

    2015-01-01

    Foundational to HCI is the notion of “the user.” Whether a cognitive processor, social actor, consumer, or even a non- user, the user in HCI has always been as much a technical construct as actual people using systems. We explore an emerging formulation of the user—the subjectivity of in- formati......, and activism. We argue that subjectivi- ties of information clarifies the relationships between de- sign choices and embodied experiences, ways that designers design users and not just products, and ways to cultivate and transform, rather than merely support, human agency.......Foundational to HCI is the notion of “the user.” Whether a cognitive processor, social actor, consumer, or even a non- user, the user in HCI has always been as much a technical construct as actual people using systems. We explore an emerging formulation of the user—the subjectivity of in- formation......—by laying out what it means and why research- ers are being drawn to it. We then use it to guide a case study of a relatively marginal use of computing—digitally mediated sexuality—to holistically explore design in rela- tion to embodiment, tactual experience, sociability, power, ideology, selfhood...

  18. Characterization of the neutron flux in the Hohlraum of the thermal column of the TRIGA Mark III reactor of the ININ

    International Nuclear Information System (INIS)

    Delfin L, A.; Palacios, J.C.; Alonso, G.

    2006-01-01

    Knowing the magnitude of the neutron flux in the reactor irradiation facilities, is so much importance for the operation of the same one, like for the investigation developing. Particularly, knowing with certain precision the spectrum and the neutron flux in the different positions of irradiation of a reactor, it is essential for the evaluation of the results obtained for a certain irradiation experiment. The TRIGA Mark III reactor account with irradiation facilities designed to carry out experimentation, where the reactor is used like an intense neutron source and gamma radiation, what allows to make irradiations of samples or equipment in radiation fields with components and diverse levels in the different facilities, one of these irradiation facilities is the Thermal Column where the Hohlraum is. In this work it was carried out a characterization of the neutron flux inside the 'Hohlraum' of the irradiation facility Thermal Column of the TRIGA Mark III reactor of the Nuclear Center of Mexico to 1 MW of power. It was determined the sub cadmic neutron flux and the epi cadmic by means of the neutron activation technique of thin sheets of gold. The maps of the distribution of the neutron flux for both energy groups in three different positions inside the 'Hohlraum' are presented, these maps were obtained by means of the irradiation of undressed thin activation sheets of gold and covered with cadmium in arrangements of 10 x 12, located parallel to 11.5 cm, 40.5 cm and 70.5 cm to the internal wall of graphite of the installation in inverse address to the position of the reactor core. Starting from the obtained values of neutron flux it was found that, for the same position of the surface of irradiation of the experimental arrangement, the relative differences among the values of neutron flux can be of 80%, and that the differences among different positions of the irradiation surfaces can vary until in a one order of magnitude. (Author)

  19. Steady-state thermal-hydraulic analysis of the Moroccan TRIGA MARK II reactor by using PARET/ANL and COOLOD-N2 codes

    International Nuclear Information System (INIS)

    Boulaich, Y.; Nacir, B.; El Bardouni, T.; Zoubair, M.; El Bakkari, B.; Merroun, O.; El Younoussi, C.; Htet, A.; Boukhal, H.; Chakir, E.

    2011-01-01

    Research highlights: → The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. → The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). → The most important conclusion is that all obtained values of DNBR, fuel center and surface temperature, cladding surface temperature and coolant temperature across the hottest channel are largely far to compromise safety of the reactor. - Abstract: The COOLOD/N2 and PARET/ANL codes were used for a steady-state thermal-hydraulic and safety analysis of the 2 MW TRIGA MARK II reactor located at the Nuclear Studies Center of Maamora (CENM), Morocco. In order to validate our PARET/ANL and COOLOD-N2 models, the fuel center temperature as function of core power was calculated and compared with the corresponding experimental values. The comparison indicates that the calculated values are in satisfactory agreement with the measurement. The main objective of this study is to ensure the safety margins of different safety related parameters by steady-state calculations at full power level (2 MW). Therefore, we have calculated the departure from nucleate boiling ratio (DNBR), fuel center and surface temperature, cladding surface temperature and coolant temperature profiles across the hottest channel. The most important conclusion is that all obtained values are largely far to compromise safety of the reactor.

  20. Fabrication and testing of a 4-node micro-pocket fission detector array for the Kansas State University TRIGA Mk. II research nuclear reactor

    Science.gov (United States)

    Reichenberger, Michael A.; Nichols, Daniel M.; Stevenson, Sarah R.; Swope, Tanner M.; Hilger, Caden W.; Unruh, Troy C.; McGregor, Douglas S.; Roberts, Jeremy A.

    2017-08-01

    Advancements in nuclear reactor core modeling and computational capability have encouraged further development of in-core neutron sensors. Micro-Pocket Fission Detectors (MPFDs) have been fabricated and tested previously, but successful testing of these prior detectors was limited to single-node operation with specialized designs. Described in this work is a modular, four-node MPFD array fabricated and tested at Kansas State University (KSU). The four sensor nodes were equally spaced to span the length of the fuel-region of the KSU TRIGA Mk. II research nuclear reactor core. The encapsulated array was filled with argon gas, serving as an ionization medium in the small cavities of the MPFDs. The unified design improved device ruggedness and simplified construction over previous designs. A 0.315-in. (8-mm) penetration in the upper grid plate of the KSU TRIGA Mk. II research nuclear reactor was used to deploy the array between fuel elements in the core. The MPFD array was coupled to an electronic support system which has been developed to support pulse-mode operation. Neutron-induced pulses were observed on all four sensor channels. Stable device operation was confirmed by testing under steady-state reactor conditions. Each of the four sensors in the array responded to changes in reactor power between 10 kWth and full power (750 kWth). Reactor power transients were observed in real-time including positive transients with periods of 5, 15, and 30 s. Finally, manual reactor power oscillations were observed in real-time.

  1. Designing for user engagement

    CERN Document Server

    Geisler, Cheryl

    2013-01-01

    Designing for User Engagement on the Web: 10 Basic Principles is concerned with making user experience engaging. The cascade of social web applications we are now familiar with - blogs, consumer reviews, wikis, and social networking - are all engaging experiences. But engagement is an increasingly common goal in business and productivity environments as well. This book provides a foundation for all those seeking to design engaging user experiences rich in communication and interaction. Combining a handbook on basic principles with case studies, it provides readers with a ric

  2. Game user experience evaluation

    CERN Document Server

    Bernhaupt, Regina

    2015-01-01

    Evaluating interactive systems for their user experience (UX) is a standard approach in industry and research today. This book explores the areas of game design and development and Human Computer Interaction (HCI) as ways to understand the various contributing aspects of the overall gaming experience. Fully updated, extended and revised this book is based upon the original publication Evaluating User Experience in Games, and provides updated methods and approaches ranging from user- orientated methods to game specific approaches. New and emerging methods and areas explored include physiologi

  3. SEVERO code - user's manual

    International Nuclear Information System (INIS)

    Sacramento, A.M. do.

    1989-01-01

    This user's manual contains all the necessary information concerning the use of SEVERO code. This computer code is related to the statistics of extremes = extreme winds, extreme precipitation and flooding hazard risk analysis. (A.C.A.S.)

  4. VIERS- User Preference Service

    Data.gov (United States)

    Department of Veterans Affairs — The Preferences service provides a means to store, retrieve, and manage user preferences. The service supports definition of enterprise wide preferences, as well as...

  5. Bevalac user's handbook

    International Nuclear Information System (INIS)

    1990-04-01

    This report is a users manual on the Bevalac accelerator facility. This paper discuses: general information; the Bevalac and its operation; major facilities and experimental areas; and experimental equipment

  6. User Interface History

    DEFF Research Database (Denmark)

    Jørgensen, Anker Helms; Myers, Brad A

    2008-01-01

    User Interfaces have been around as long as computers have existed, even well before the field of Human-Computer Interaction was established. Over the years, some papers on the history of Human-Computer Interaction and User Interfaces have appeared, primarily focusing on the graphical interface era...... and early visionaries such as Bush, Engelbart and Kay. With the User Interface being a decisive factor in the proliferation of computers in society and since it has become a cultural phenomenon, it is time to paint a more comprehensive picture of its history. This SIG will investigate the possibilities...... of  launching a concerted effort towards creating a History of User Interfaces. ...

  7. SILMUSCEN and CLIGEN User`s Guide

    Energy Technology Data Exchange (ETDEWEB)

    Carter, T.; Tuomenvirta, H. [Finnish Meteorological Inst., Helsinki (Finland); Posch, M. [Water and Environment Research Inst., Helsinki (Finland)

    1995-12-31

    This User`s Guide has been prepared to provide recommendations for the selection and application of climatic scenarios in the Finnish Research Programme on Climate Change (SILMU). These scenarios are required for conducting impact studies in SILMU. They should reflect the current range of estimates of future climate in the Finnish region. In addition, they should be consistent with other projections of importance in impact studies, such as future atmospheric composition and sea level. Section 2 provides some background information about the types of scenarios required in SILMU and Section 3 offers a general description of the scenarios. In Section 4 there is some advice on applying sensitivity studies to complement the use of scenarios. Section 5 explains the installation of the SILMUSCEN program and Section 6 guides the user through some examples to illustrate how SILMUSCEN can be used. Section 7 offers some recommendations on which scenarios to adopt for different impact assessments. In order to ensure some compatibility between impact studies in SILMU, it is very important that the recommendations in this section are followed as far as possible. Section 8 addresses important omissions from the computer program and suggests procedures to adopt in their absence. Section 9 explores alternative methods of specifying the baseline climate, and shows how scenario adjustments to the baseline can be made. in Section 10, the stochastic weather generator, CLIGEN, is described and its use illustrated by means of examples. Finally, possible refinements of the programs are outlined in Section 11, along with contact names and addresses for obtaining further information. (36 refs.)

  8. VOLTTRON: User Guide

    Energy Technology Data Exchange (ETDEWEB)

    Lutes, Robert G.; Katipamula, Srinivas; Akyol, Bora A.; Tenney, Nathan D.; Haack, Jereme N.; Monson, Kyle E.; Carpenter, Brandon J.

    2014-04-24

    This document is a user guide for the deployment of the Transactional Network platform and agent/application development within the VOLTTRON. The intent of this user guide is to provide a description of the functionality of the Transactional Network Platform. This document describes how to deploy the platform, including installation, use, guidance, and limitations. It also describes how additional features can be added to enhance its current functionality.

  9. Program Trains NASTRAN Users

    Science.gov (United States)

    Grooms, H. R.; Hinz, P. J.; Collier, M. A.; Cox, Kim D.; Merriman, Warren J.; Commerford, Gerry

    1994-01-01

    Rockwell Environment and NASTRAN Trainer (RENT) computer program developed to assist new and current users of NASTRAN finite-element computer code. Provides organized, systematic collection of IBM(R) features consisting of panels, clists, skeletons, and messages, along with FORTRAN and Pascal programs and example NASTRAN data files. Enables each user to learn at his or her own pace. Written in VS/FORTRAN, VS/ Pascal, and IBM(R) job-control language for an IBM(R) computer system.

  10. Metadata: A user`s view

    Energy Technology Data Exchange (ETDEWEB)

    Bretherton, F.P. [Univ. of Wisconsin, Madison, WI (United States); Singley, P.T. [Oak Ridge National Lab., TN (United States)

    1994-12-31

    An analysis is presented of the uses of metadata from four aspects of database operations: (1) search, query, retrieval, (2) ingest, quality control, processing, (3) application to application transfer; (4) storage, archive. Typical degrees of database functionality ranging from simple file retrieval to interdisciplinary global query with metadatabase-user dialog and involving many distributed autonomous databases, are ranked in approximate order of increasing sophistication of the required knowledge representation. An architecture is outlined for implementing such functionality in many different disciplinary domains utilizing a variety of off the shelf database management subsystems and processor software, each specialized to a different abstract data model.

  11. GRSAC Users Manual

    International Nuclear Information System (INIS)

    Ball, S.J.; Nypaver, D.J.

    1999-01-01

    An interactive workstation-based simulation code (GRSAC) for studying postulated severe accidents in gas-cooled reactors has been developed to accommodate user-generated input with ''smart front-end'' checking. Code features includes on- and off-line plotting, on-line help and documentation, and an automated sensitivity study option. The code and its predecessors have been validated using comparisons with a variety of experimental data and similar codes. GRSAC model features include a three-dimensional representation of the core thermal hydraulics, and optional ATWS (anticipated transients without scram) capabilities. The user manual includes a detailed description of the code features, and includes four case studies which guide the user through four different examples of the major uses of GRSAC: an accident case; an initial conditions setup and run; a sensitivity study; and the setup of a new reactor model

  12. GRSAC Users Manual

    Energy Technology Data Exchange (ETDEWEB)

    Ball, S.J.; Nypaver, D.J.

    1999-02-01

    An interactive workstation-based simulation code (GRSAC) for studying postulated severe accidents in gas-cooled reactors has been developed to accommodate user-generated input with ''smart front-end'' checking. Code features includes on- and off-line plotting, on-line help and documentation, and an automated sensitivity study option. The code and its predecessors have been validated using comparisons with a variety of experimental data and similar codes. GRSAC model features include a three-dimensional representation of the core thermal hydraulics, and optional ATWS (anticipated transients without scram) capabilities. The user manual includes a detailed description of the code features, and includes four case studies which guide the user through four different examples of the major uses of GRSAC: an accident case; an initial conditions setup and run; a sensitivity study; and the setup of a new reactor model.

  13. TRANS-USERS

    DEFF Research Database (Denmark)

    in Denmark, Sweden and France. The five case studies are: The industrialised home building concept BoKlok, a web based product configurator for kitchens by HTH, the innovative potential of the dual role of employees as both user and employee in Rockwool, the application of quality management systems......This report addresses user-orientated strategies for industrialising the construction industry. In the first part, the objectives of the study are described along with the theoretical framework and the research design of the study. The second part of this report contains five case studies conducted...... to redesign production and business processes to accommodate for users' requirements (Maisons MACCHI), and the client as driver of innovation on the construction and renovation of the low budget hotel brand Formule 1 of ACCOR Hotels. In the third part, the discussion and conclusion addresses three interlinked...

  14. Evaluating User Participation and User Influence in an Enterprise System

    Science.gov (United States)

    Gibbs, Martin D.

    2010-01-01

    Does user influence have an impact on the data quality of an information systems development project? What decision making should users have? How can users effectively be engaged in the process? What is success? User participation is considered to be a critical success factor for Enterprise Resource Planning (ERP) projects, yet there is little…

  15. HANARO user support

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jeong Soo; Kim, Y. J.; Seong B.S. [and others

    2003-06-01

    The purpose of this project is to support external user for the promotion of HANARO common utilization effectively. To do this, external manpower was recruited and trained. Also, in order to find out and cultivate HANARO user, practice-oriented education was done. The total number of project selected as the promotion of HANARO common utilization was 31 in this year. These composed of four fields such as neutron beam utilization, materials/nuclear materials irradiation test, neutron activation analysis and radioisotope production. In each field, the numbers of project were 17, 7, 4 and 3 respectively. At first, from a selected project of view, supporting ratio by external manpower was reached to the 58%, that is, 18 out of 31 project was supported. In each field, it was 82% for neutron beam utilization and 100% for neutron activation analysis. Also, from the utilization time point of view, supporting ratio of external manpower was reached to 30% for neutron beam utilization and 59% for neutron activation analysis. Otherwise, supporting ratio by manpower in KAERI was reached to 97%, that is, 30 out of 31 project was supported. Also, from the utilization time point of view, total supporting ratio was reached to 15%. In each field, it was 20% for neutron beam utilization, 18% for materials/nuclear materials irradiation test, 20% for neutron activation analysis and 6% for radioisotope production. In order to contribute finding and cultivating of HANARO potential user and increase utilization ratio of HANARO experimental facility, practice-oriented HANARO user education has been done. At first, 32 participants from industries, universities, institutes were educated and practiced on HRPD/SANS instrument in the field of neutron beam utilization. Otherwise, in order to support external user effectively, external manpower were trained. Also, more effective support for external user could be possible through the grasping difficulty and problem on the performance of project

  16. CRACUK user's guide

    International Nuclear Information System (INIS)

    Egan, M.J.; Nixon, W.; Brearley, I.R.

    1986-08-01

    The CRACUK computer code is a revised version of the US consequence modelling code CRAC2, adapted to suit UK applications. Modifications to various models within the code have led to certain changes in the input data requirements for CRACUK in comparison with CRAC2. This guide, written in the form of an Appendix to the CRAC2 User Guide, includes descriptions of the input data layout as it has been altered for use in CRACUK. Used in conjunction with the CRAC2 User Guide, this publication should allow easy use of the CRACUK code. (author)

  17. EPRINT ARCHIVE USER SURVEY

    CERN Multimedia

    2001-01-01

    University of Southampton invites the CERN community to participate in a survey Professor Stevan Harnad is conducting on current users and non-users of Eprint Archives. http://www.eprints.org/survey/ The findings will be used to suggest potential enhancements of the services as well as to get a deeper understanding of the very rapid developments in the on-line dissemination and use of scientific and scholarly research. (The survey is anonymous. Revealing your identity is optional and it will be kept confidential.)

  18. User friendly packaging

    DEFF Research Database (Denmark)

    Geert Jensen, Birgitte

    2010-01-01

    User-friendly Packaging” aims to create a platform for developing more user-friendly packaging. One intended outcome of the project is a guideline that industry can use in development efforts. The project also points the way for more extended collaboration between companies and design researchers. How...... can design research help industry in packaging innovation?......Most consumers have experienced occasional problems with opening packaging. Tomato sauce from the tinned mackerel splattered all over the kitchen counter, the unrelenting pickle jar lid, and the package of sliced ham that cannot be opened without a knife or a pair of scissors. The research project...

  19. The OSIRIS user guide

    CERN Document Server

    Telling, M T F

    2003-01-01

    This user guide contains all the information necessary to perform a successful neutron scattering experiment on the OSIRIS spectrometer at ISIS, RAL, UK. Since OSIRIS is a continually evolving and improving instrument some information contained within this manual may become redundant. However, the basic instrument operating procedures should remain essentially unchanged. While updated manuals will be produced when appropriate, the most comprehensive source of information concerning OSIRIS is the Instrument Scientist/Local Contact. It would be appreciated, however, if this user guide were the first point of call should problems arise

  20. User friendly packaging

    DEFF Research Database (Denmark)

    Geert Jensen, Birgitte

    2010-01-01

    Most consumers have experienced occasional problems with opening packaging. Tomato sauce from the tinned mackerel splattered all over the kitchen counter, the unrelenting pickle jar lid, and the package of sliced ham that cannot be opened without a knife or a pair of scissors. The research project...... “User-friendly Packaging” aims to create a platform for developing more user-friendly packaging. One intended outcome of the project is a guideline that industry can use in development efforts. The project also points the way for more extended collaboration between companies and design researchers. How...... can design research help industry in packaging innovation?...

  1. Perspectives on User Satisfaction Surveys.

    Science.gov (United States)

    Cullen, Rowena

    2001-01-01

    Discusses academic libraries, digital environments, increasing competition, the relationship between service quality and user satisfaction, and user surveys. Describes the SERVQUAL model that measures service quality and user satisfaction in academic libraries; considers gaps between user expectations and managers' perceptions of user…

  2. Users Office - Removal

    CERN Multimedia

    CERN Bulletin

    2010-01-01

    As of 8 December 2010 and until the end of February 2011, the Users Office will move from Bldg. 60. New Location : Bldg. 510-R-033 Opening Hours: Monday, Tuesday, Thursday, Friday : 08.30 – 12.30 Monday to Friday: 14.00 – 16.00 Closed Wednesday mornings.

  3. AELIB user's manual

    International Nuclear Information System (INIS)

    Evans, L.E.; Klawitter, G.L.

    1981-05-01

    This report is an updatable manual for users of AELIB, the AECL Library of FORTRAN-callable routines for the CRNL CDC 6600/CYBER 170 system. It provides general advice on the use of this library and detailed information on the selection and usage of particular library routines

  4. SALOME. User's guide

    International Nuclear Information System (INIS)

    Zastrow, K.D.

    1994-01-01

    A user's guide for a least squares fit analysis program is presented, which has been developed for VUV and visible plasma spectroscopy on the Extrap-T1 experiment. The program can be used for line and multiplet identification, multiplet intensity and line width measurements. Atomic data are used extensively to aid the interpretation of the spectra. 7 figs

  5. Neem: A User's Manual

    Indian Academy of Sciences (India)

    Home; Journals; Resonance – Journal of Science Education; Volume 2; Issue 5. Neem: A User's Manual The 'Free Tree' – Its Healing Power and Other Uses. M D Subash Chandran. Book Review Volume 2 Issue 5 May 1997 pp 84-86. Fulltext. Click here to view fulltext PDF. Permanent link:

  6. CDPOP Users Manual

    Science.gov (United States)

    E. L. Landguth; B. K. Hand; J. M. Glassy; S. A. Cushman; M. Jacobi; T. J. Julian

    2011-01-01

    The goal of this user manual is to explain the technical aspects of the current release of the CDPOP program. CDPOP v1.0 is a major extension of the CDPOP program (Landguth and Cushman 2010). CDPOP is an individual-based program that simulates the influences of landscape structure on emergence of spatial patterns in population genetic data as functions of individual-...

  7. 16. ESRF users meeting

    International Nuclear Information System (INIS)

    Coraux, J.; Renevier, H.; Favre-Nicolin, V.; Daudin, B.; Proietti, M.G.; Renaud, G.; Fowler, B.; Mercer, D.L.; Omar, A.H.; Thompson, P.; Markovic, N.M.; Stamenkovic, V.; Lucas, C.A.; Andrejczuk, A.; Kwiatkowska, J.; Dobrzynski, L.; Zukowski, E.; Bellin, Ch.; Loupias, G.; Shukla, A.; Buslaps, Th.; Stankov, S.; Sladecek, M.; Slezak, T.; Korecki, J.; Spiridis, N.; Sepiol, B.; Vogl, G.; Chumakov, A.; Ruffer, R.; Hermann, R.P.; Grandjean, F.; Schweika, W.; Long, G.J.; Leupold, O.; Belrhall, H.; Caserotto, H.; Dauvergne, F.; Geoffroy, L.; Guljarro, M.; Launer, L.; Levault, B.; Walsh, M.; Beckers, M.; Schell, N.; Martins, R.M.S.; Mucklich, A.; Moller, W.; Silva, R.J.C.; Mahesh, K.K.; Braz Fernandes, F.M.; Tejas, Parikh; Neil, Fellows; Durodola, J.; Slawinski, W.; Przenioslo, R.; Sosnowska, I.; Suard, E.

    2006-01-01

    This document gathers the posters that were presented during the poster session of this workshop. These posters highlight the results obtained by ESRF'users in different fields such as surface structure, Compton scattering studies, localized vibrational modes in thermoelectric materials, Ni-Ti thin films, residual stresses in superconducting wires, and changes in crystal and magnetic structure of NdFeO 3

  8. CONSUME: users guide.

    Science.gov (United States)

    R.D. Ottmar; M.F. Burns; J.N. Hall; A.D. Hanson

    1993-01-01

    CONSUME is a user-friendly computer program designed for resource managers with some working knowledge of IBM-PC applications. The software predicts the amount of fuel consumption on logged units based on weather data, the amount and fuel moisture of fuels, and a number of other factors. Using these predictions, the resource manager can accurately determine when and...

  9. Educating the Music User

    Science.gov (United States)

    Adams, Mark C.

    2016-01-01

    To better serve students' evolving needs in music, music educators must connect classroom learning with how students use and interact with music in their daily lives. One way to accomplish this is by approaching classrooms with the music user in mind, which can open new possibilities for meaningful music making and remove students from the…

  10. User's guide for ICE

    International Nuclear Information System (INIS)

    Fraley, S.K.

    1976-07-01

    ICE is a cross-section mixing code which will accept cross sections from an AMPX working library and produce mixed cross sections in the AMPX working library format, ANISN format, and the group-independent ANISN format. User input is in the free-form or fixed-form FIDO structure. The code is operable as a module in the AMPX system

  11. Governing the water user

    NARCIS (Netherlands)

    Rap, Edwin; Wester, Flip

    2017-01-01

    This article traces a policy shift that makes the ‘water user’ the main subject of water governance. From a Foucauldian perspective on governmentality these new subjectivities accompany neo-liberal governmental technologies to devolve autonomy from state institutions to an active user base, whilst

  12. TO STORES USERS

    CERN Multimedia

    SPL Division

    2001-01-01

    Stores users are informed that the Stores (Central, Emergency window, Raw materials, Chemical products and Prévessin Self service stores) will be closed on Friday, 7 December owing to migration of the Stores computers to Windows 2000. Thank you for your understanding.

  13. Personal lifelong user model clouds

    DEFF Research Database (Denmark)

    Dolog, Peter; Kay, Judy; Kummerfeld, Bob

    This paper explores an architecture for very long term user modelling, based upon personal user model clouds. These ensure that the individual's applications can access their model whenever it is needed. At the same time, the user can control the use of their user model. So, they can ensure...... which combines both. Finally we discuss implications of our approach for consistency and freshness of the user model information....... it is accessed only when and where they wish, by applications that they wish. We consider the challenges of representing user models so that they can be reused by multiple applications. We indicate potential synergies between distributed and centralised user modelling architectures, proposing an architecture...

  14. SHARP User Manual

    Energy Technology Data Exchange (ETDEWEB)

    Yu, Y. Q. [Argonne National Lab. (ANL), Argonne, IL (United States); Shemon, E. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Thomas, J. W. [Argonne National Lab. (ANL), Argonne, IL (United States); Mahadevan, Vijay S. [Argonne National Lab. (ANL), Argonne, IL (United States); Rahaman, Ronald O. [Argonne National Lab. (ANL), Argonne, IL (United States); Solberg, Jerome [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2016-03-31

    SHARP is an advanced modeling and simulation toolkit for the analysis of nuclear reactors. It is comprised of several components including physical modeling tools, tools to integrate the physics codes for multi-physics analyses, and a set of tools to couple the codes within the MOAB framework. Physics modules currently include the neutronics code PROTEUS, the thermal-hydraulics code Nek5000, and the structural mechanics code Diablo. This manual focuses on performing multi-physics calculations with the SHARP ToolKit. Manuals for the three individual physics modules are available with the SHARP distribution to help the user to either carry out the primary multi-physics calculation with basic knowledge or perform further advanced development with in-depth knowledge of these codes. This manual provides step-by-step instructions on employing SHARP, including how to download and install the code, how to build the drivers for a test case, how to perform a calculation and how to visualize the results. Since SHARP has some specific library and environment dependencies, it is highly recommended that the user read this manual prior to installing SHARP. Verification tests cases are included to check proper installation of each module. It is suggested that the new user should first follow the step-by-step instructions provided for a test problem in this manual to understand the basic procedure of using SHARP before using SHARP for his/her own analysis. Both reference output and scripts are provided along with the test cases in order to verify correct installation and execution of the SHARP package. At the end of this manual, detailed instructions are provided on how to create a new test case so that user can perform novel multi-physics calculations with SHARP. Frequently asked questions are listed at the end of this manual to help the user to troubleshoot issues.

  15. Demarcating User eXperience

    Science.gov (United States)

    Roto, Virpi

    This panel discusses the scoping of user experience as a research field. User experience is a crossing point of several disciplines, each of which tends to define user experience from their own perspective. The distinguished panelists from academia and industry represent the different perspectives to user experience: Traditional human-computer interaction, Psychology, Cognitive psychology, and Design. The goal of the panel is to get one step closer to a shared understanding of the concept of user experience.

  16. Effects of Cooling Fluid Flow Rate on the Critical Heat Flux and Flow Stability in the Plate Fuel Type 2 MW TRIGA Reactor

    Directory of Open Access Journals (Sweden)

    H. P. Rahardjo

    2017-12-01

    Full Text Available The conversion program of the 2 MW TRIGA reactor in Bandung consisted of the replacement of cylindrical fuel (produced by General Atomic with plate fuel (produced by BATAN. The replacement led into the change of core cooling process from upward natural convection type to downward forced convection type, and resulted in different thermohydraulic safety criteria, such as critical heat flux (CHF limit, boiling limit, and cooling fluid flow stability. In this paper, a thermohydraulic safety analysis of the converted TRIGA reactor is presented by considering the Dynamic Nucleate Boiling Ratio (DNBR criterion, Onset Nucleate Boiling Ratio (ONBR limit, and cooling fluid flow stability at various cooling fluid flow rate.The numerical analyses were performed using the HEATHYD program on the hottest channels of reactor core.The combination of heat transfer and fluid flow analysis were conducted for reactor operation at 2 MW with 20 fuel element bundles and four control rod bundles. Incoming fluid flow to the cooling channel was fixed at 44.5 °C temperature and 1.9970 bar pressure, and its flow rate was varied from 1.25 to 3.5 m3/h. By inputting these values, as well as the total power of fuel elements per bundle, the wall temperature distribution of the plate fuel element, cooling fluid temperature distribution, and pressure losses in the channels were obtained for the analysis of CHF limit, boiling limit, and flow stability. It was shown that no boiling occurred for the cooling fluid flow rate range of 2.4 to 3.5 m3/h, and even at the cooling fluid flow rate of 1.25 m3/h where some bubbles occurred, the DNBR was higher than the critical limit (more than 23 while the flow stability criterion in some channels were slightly less than 1 (unstable. At the cooling fluid flow rate of 1.4 m3/h, however, the flow became stable in all channel. Observing the user experience a practitioner's guide to user research

    CERN Document Server

    Kuniavsky, Mike; Goodman, Elizabeth

    2012-01-01

    The gap between who designers and developers imagine their users are, and who those users really are can be the biggest problem with product development. Observing the User Experience will help you bridge that gap to understand what your users want and need from your product, and whether they'll be able to use what you've created. Filled with real-world experience and a wealth of practical information, this book presents a complete toolbox of techniques to help designers and developers see through the eyes of their users. It provides in-depth coverage of 13 user experience research techniques

  17. IT User Community Survey

    CERN Multimedia

    Peter Jones (IT-CDA-WF)

    2016-01-01

    IT-CDA is gathering information to more accurately form a snapshot of the CERN IT user community and we would appreciate you taking time to complete the following survey.   We want to use this survey to better understand how the user community uses their devices and our services, and how the delivery of those services could be improved. You will need to authenticate to complete the survey. However please note that your responses are confidential and will be compiled together and analysed as a group. You can also volunteer to offer additional information if you so wish. This survey should take no longer than 5 minutes. Thanks in advance for your collaboration.

  18. Portraying User Interface History

    DEFF Research Database (Denmark)

    Jørgensen, Anker Helms

    2008-01-01

    The user interface is coming of age. Papers adressing UI history have appeared in fair amounts in the last 25 years. Most of them address particular aspects such as an in­novative interface paradigm or the contribution of a visionary or a research lab. Contrasting this, papers addres­sing UI...... history at large have been sparse. However, a small spate of publications appeared recently, so a reasonable number of papers are available. Hence this work-in-progress paints a portrait of the current history of user interfaces at large. The paper first describes a theoretical framework recruited from...... history. Next the paper analyses a selected sample of papers on UI history at large. The analysis shows that the current state-of-art is featured by three aspects: Firstly internalism, in that the papers adress the tech­nologies in their own right with little con­text­ualization, secondly whiggism...

  19. Growing Data User Communities

    Science.gov (United States)

    Wiggin, B.

    2017-12-01

    Preserving data is not only a technical challenge. Perhaps the best way to protect data is to use it. Grassroots efforts to make research-quality copies of federal data continue to energize communities of data users who often did not previously recognize themselves as open earth data users. Beyond "data rescue" events, the Data Refuge project researches how federal climate and environmental data are used downstream in a variety of local communities and municipal governments to address everyday issues: public health, municipal safety, and even the preservation of cultural heritage assets. Documenting the diverse uses made of open earth data beyond the earth sciences research community grows the community who, in making use of data, also helps to preserve it.