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Sample records for 2-dollar triga reactor

  1. TRIGA reactor characteristics

    The general design, characteristics and parameters of TRIGA reactors and fuel are described. This is a training module with the learning objectives: to understand the basics of the physics and mechanical design of the TRIGA fuel as well as its unique operational characteristics and realize the differences between TRIGA fuels and other more traditional. 10 figs., 6 tabs. (nevyjel)

  2. TRIGA reactor main systems

    This module describes the main systems of low power (<2 MW) and higher power (≥2 MW) TRIGA reactors. The most significant difference between the two is that forced reactor cooling and an emergency core cooling system are generally required for the higher power TRIGA reactors. However, those TRIGA reactors that are designed to be operated above 3 MW also use a TRIGA fuel that is specifically designed for those higher power outputs (3 to 14 MW). Typical values are given for the respective systems although each TRIGA facility will have unique characteristics that may only be determined by the experienced facility operators. Due to the inherent wide scope of these research reactor facilities construction and missions, this training module covers those systems found at most operating TRIGA reactor facilities but may also discuss non-standard equipment that was found to be operationally useful although not necessarily required. (author)

  3. TRIGA reactor characteristics

    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)

  4. TRIGA research reactors

    TRIGA (Training, Research, Isotope production, General-Atomic) has become the most used research reactor in the world with 65 units operating in 24 countries. The original patent for TRIGA reactors was registered in 1958. The success of this reactor is due to its inherent level of safety that results from a prompt negative temperature coefficient. Most of the neutron moderation occurs in the nuclear fuel (UZrH) because of the presence of hydrogen atoms, so in case of an increase of fuel temperature, the neutron spectrum becomes harder and neutrons are less likely to fission uranium nuclei and as a consequence the power released decreases. This inherent level of safety has made this reactor fit for training tool in university laboratories. Some recent versions of TRIGA reactors have been designed for medicine and industrial isotope production, for neutron therapy of cancers and for providing a neutron source. (A.C.)

  5. Power calibrations for TRIGA reactors

    The purpose of this paper is to establish a framework for the calorimetric power calibration of TRIGA reactors so that reliable results can be obtained with a precision better than ± 5%. Careful application of the same procedures has produced power calibration results that have been reproducible to ± 1.5%. The procedures are equally applicable to the Mark I, Mark II and Mark III reactors as well as to reactors having much larger reactor tanks and to TRIGA reactors capable of forced cooling up to 3 MW in some cases and 15 MW in another case. In the case of forced cooled TRIGA reactors, the calorimetric power calibration is applicable in the natural convection mode for these reactors using exactly the same procedures as are discussed below for the smaller TRIGA reactors (< 2 MW)

  6. TRIGA reactor operating experience

    The Oregon State TRIGA Reactor (OSTR) has been in operation 3 years. Last August it was upgraded from 250 kW to 1000 kW. This was accomplished with little difficulty. During the 3 years of operation no major problems have been experienced. Most of the problems have been minor in nature and easily corrected. They came from lazy susan (dry bearing), Westronics Recorder (dead spots in the range), The Reg Rod Magnet Lead-in Circuit (a new type lead-in wire that does not require the lead-in cord to coil during rod withdrawal hss been delivered, much better than the original) and other small corrections

  7. The research reactor TRIGA Mainz

    Paper dwells upon the design and the operation of one of the German test reactors, namely, the TRIGA Mainz one (TRIGA: Training Research Isotope Production General Atomic). The TRIGA reactor is a pool test reactor the core of which contains a graphite reflector and is placed into 2 m diameter and 6.25 m height aluminum vessel. There are 75 fuel elements in the reactor core, and any of them contains about 36 g of 235U. The TRIGA reactors under the stable operation enjoy wide application to ensure tests and irradiation, namely: neutron activation analysis, radioisotope production, application of a neutron beam to ensure the physical, the chemical and the medical research efforts. Paper presents the reactor basic experimental program lines

  8. The research reactor TRIGA Mainz

    The TRIGA Mark II reactor at the Institut fuer Kernchemie became first critical on August 3rd, 1965. It 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. A survey of the research programmes performed at the TRIGA Mainz is given covering applications in basic research as well as applied science in nuclear chemistry and nuclear physics. Furthermore, the reactor is used for neutron activation analysis and for education and training of scientists, teachers, students and technical personal. Important projects for the future of the TRIGA Mainz are the UCN (ultra cold neutrons) experiment, fast chemical separation, medical applications and the use of the NAA as well as the use of the reactor facility for the training of students in the fields of nuclear chemistry, nuclear physics and radiation protection. Taking into account the past and future operation schedule and the typically low burn-up of TRIGA fuel elements (∝4 g U-235/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. (orig.)

  9. TRIGA reactor health physics considerations

    The factors influencing the complexity of a TRIGA health physics program are discussed in details in order to serve as a basis for later consideration of various specific aspects of a typical TRIGA health physics program. The health physics program must be able to provide adequate assistance, control, and safety for individuals ranging from the inexperienced student to the experienced postgraduate researcher. Some of the major aspects discussed are: effluent release and control; reactor area air monitoring; area monitoring; adjacent facilities monitoring; portable instrumentation, personnel monitoring. TRIGA reactors have not been associated with many significant occurrences in the area of health physics, although some operational occurrences have had health physics implications. One specific occurrence at OSU is described involving the detection of non-fission-product radioactive particulates by the continuous air monitor on the reactor top. The studies of this particular situation indicate that most of the particulate activity is coming from the rotating rack and exhausting to the reactor top through the rotating rack loading tube

  10. Higher power density TRIGA research reactors

    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

  11. TRIGA reactor owners' seminar. Papers and abstracts

    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

  12. Oregon State University TRIGA Reactor annual report

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

    1979-08-31

    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.

  13. Oregon State University TRIGA Reactor annual report

    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

  14. TRIGA Reactor Power Upgrading Analysis

    Reactor physics safety analysis supporting the power upgrading from 1MW to 2MW of a typical TRIGA Mark II reactor is presented for steady state and pulse operation. The analysis is performed for mixed core configuration consisting of two types of fuel elements: standard 8,5% or 12% stainless-steel clad fuel elements and LEU fuel elements (20% uranium concentration). The following reactor physics codes are applied: WIMS, TRIGAC, EXTERMINATOR, PULSTRI and TRISTAN. Results of the calculations are compared to experiments for steady state operation at 1 MW. The analysis shows that besides technical modifications of the core (installation of an additional control rod) also some strict administrative limitations have to be imposed on operational parameters (excess reactivity, pulse reactivity, core composition) to assure safe operation within design limits. (author)

  15. Operation experience with the TRIGA Reactor Vienna

    Since the last European TRIGA Users Conference in Bucharest, Romania in September 1992 the TRIGA reactor Vienna operated without any major undesired shutdown. Some problems were centred around the new microprocessor controlled instrumentation installed in summer 1992. The fuel behaviour was excellent, no fuel failures were experienced. The experimental facilities were extensively used for students education and training

  16. Design improvements in TRIGA reactors

    There have been many design improvements to TRIGA reactor hardware in the past twelve years. One of the more important and most obvious improvements has been in the area of reactor instrumentation. The low profile, completely transistorized Mark III console was a great step forward in a low maintenance, high reliability instrumentation system. Other design improvements include the lazy susan specimen pickup assembly; the specimen container; an empty stainless steel fuel element which can be filled with samples and can be located anywhere in the core; the flexible fuel handling tool; a new fuel measuring tool design; the shock absorber on the adjustable transient rod drive; new testing and evaluation procedures on the thermocouples and other

  17. Dynamics of TRIGA-3 Salazar Reactor

    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)

  18. Upgrading Status Of Bandung Triga 2000 Reactor

    Upgrading Status Of Bandung TRIGA 2000 Reactor. Upgrading of TRIGA Mark II Reactor from 1000 k W to 2000 k W has been done. On June 24, 2000 it has been inaugurated by the Vice President, Madame Megawati Soekarnoputri. The solution of the problems faced in the upgrading should be described here since some experiences got during the process probably are very useful, especially the methods in finishing the project

  19. TRIGA research reactor activities around the world

    Recent activities at several overseas TRIGA installations are discussed in this paper, including reactor performance, research programs under way, and plans for future upgrades. The following installations are included: (1) 14,000-kW TRIGA at the Institute for Nuclear Research, Pitesti, Romania; (2) 2,000-kW TRIGA Mark II at the Institute of Nuclear Technology, Dhaka, Bangladesh; (3) 3,000-kW TRIGA conversion, Philippine Nuclear Research Institute, Quezon City, Philippines; and (4) other ongoing installations, including a 1,500-kW TRIGA Mark II at Rabat, Morocco, and a 1,000-kW conversion/upgrade at the Institute Asunto Nucleares, Bogota, Columbia

  20. TRIGA research reactor activities around the world

    Chesworth, R.H.; Razvi, J.; Whittemore, W.L. (General Atomics, San Diego, CA (United States))

    1991-11-01

    Recent activities at several overseas TRIGA installations are discussed in this paper, including reactor performance, research programs under way, and plans for future upgrades. The following installations are included: (1) 14,000-kW TRIGA at the Institute for Nuclear Research, Pitesti, Romania; (2) 2,000-kW TRIGA Mark II at the Institute of Nuclear Technology, Dhaka, Bangladesh; (3) 3,000-kW TRIGA conversion, Philippine Nuclear Research Institute, Quezon City, Philippines; and (4) other ongoing installations, including a 1,500-kW TRIGA Mark II at Rabat, Morocco, and a 1,000-kW conversion/upgrade at the Institute Asunto Nucleares, Bogota, Columbia.

  1. BNCT activities at Slovenian TRIGA research reactor

    It has been reported that satisfactory thermal/epithermal neutron beams for Boron Neutron Capture Therapy (BNCT) could be designed at TRIGA research reactors These reactors are generally perceived as being safe to install and operate in populated areas. This contribution presents the most recent BNCT research activities on the 'Jozef Stefan' Institute, where epithermal neutron beam for 'in-vitro' irradiation has been developed and experimentally verified. Furthermore, The Monte Carlo feasibility study of development of the epithermal neutron beam for BNCT clinical trials of human patients in thermalising column (TC) of TRIGA reactor has been carried out. The simulation results prove, that a BNCT irradiation facility with performances, comparable to existing beam throughout the world, could be installed in TC of the TRIGA reactor. (author)

  2. Component failure data base of TRIGA reactors

    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)

  3. PUSPATI Triga reactor fuel worth measurement

    The reactivity worth of fuel elements in the B, C, D, E and F rings in the PUSPATI TRIGA Reactor core with respect to water as well as that of dummy fuel element (graphite filled) in the G ring were measured. The reactivity worth of 8.5 w/o standard TRIGA fuel element with respect to the dummy element in the B to F rings were also determined. The measured results agreed with the typical values given by the reactor supplier, General Atomatic Company, to within eight percents. (author)

  4. Operation experience with the TRIGA reactor Wien

    Boeck, H. (Atominstitut, Vienna (Austria))

    1999-12-15

    The TRIGA Mark-II reactor Wien has been in operation more than 36 years. The average operation time is about 230 days per year with 90 % of this time at nominal power of 250 kW. The remaining 10 % operation time is used for students' training cources at low power level. Pulse operation is rather infrequent with about 5 to 10 pulses per year. The TRIGA reactor Wien is well utilized and in an excellent technical state. There are no technical or economical reasons to consider an imminent shut-down. However, the present fuel return policy might influence the destiny of this facility in the next decade. (orig.)

  5. TRIGA Mark-II, III reactor operation

    TRIGA Mark-II reactor has been primarily utilized as usual for the fundamental reactor experiments for university students. The annual operating time is 1,100 hours and the gross thermal output is 17,159 KWH, having consumed 0.88g of U-235. The reconstuction work for the control console of this reactor is now in progress and will be completed in early part of 1982. TRIGA Mark-III reactor has been operated mainly for radioisotope production, test pin irradiation and activation analysis, etc., as well as solid state physics experiments using the beamports. The annual operatino. time is amounted to 3,530 hours being the longest since the beginning of its criticality, and the gross thermal output is 4,113,013 KWH, whereas the U-235 consumption is estimated at 212.82 g. 462 samples were irradiated to produce 9 kinds of radioisotopes. In order to carry out the test pin irradiation experiment, the core configuration of TRIGA Mark-III was changed by loadinq 6 fresh fuels at G-ring as of July 1981 and a new irradiation facility consisting of 14 tubes was manufactured in place of Rotary Specimen Rack. Then 7 kinds of physics experiments were performed over a two week period to scrutinize the chanaed core characteristics. In addition, the present TRIGA Mark-III reactor fuel storage tank was enlarged and the distilled water production facility was renewed to improve its production efficiency. (Author)

  6. Optimum burnup of BAEC TRIGA research reactor

    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

  7. Assessment criteria for TRIGA reactors performances

    Full text: The international statistic data show that a number of 325 research reactors are now in operation. Their constructional and functional diversity is very large, a great share being represented by the TRIGA family reactors. Such reactors are now operating at: Tucson, Arizona - USA (1958); Austin, Texas - USA (1963); Belo Horizonte - Brazil (1960); Mainz - Germany (1975); Omaha - Veterans (1959); Heidelberg - Germany (1966); Bandung - Indonesia (1964/1971); Dalat - Vietnam (1963); Pavia - Italy (1965);, Rikkyo, Yokosuka - Japan (1961); Rome, Casaccia - Italy (1960); Seoul - Rep. of Korea (1962); Wien - Austria (1962); Dasa Bethesda, MD - USA (1962); Pitesti, Arges - Romania (1979), etc. In the proposed paper, the author sets the evaluation criteria for the TRIGA-type reactors performances. The treated phenomena can be described through functions of the type φ(N1,N2,N3,...) = constant, where for example, N1 FAα1Bβ1Cγ1Eδ1. (author)

  8. Utilization of Slovenian TRIGA Mark II reactor

    TRIGA Mark II research reactor at the Jozef Stefan Institute [JSI] is 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 briefly describes the aforementioned activities and shows that even such small reactors are still indispensable in nuclear science and technology. (author)

  9. Physics and kinetics of TRIGA reactor

    This training module is written as an introduction to reactor physics for reactor operators. It assumes the reader has a basic, fundamental knowledge of physics, materials and mathematics. The objective is to provide enough reactor theory knowledge to safely operate a typical research reactor. At this level, it does not necessarily provide enough information to evaluate the safety aspects of experiment or non-standard operation reviews. The material provides a survey of basic reactor physics and kinetics of TRIGA type reactors. Subjects such as the multiplication factor, reactivity, temperature coefficients, poisoning, delayed neutrons and criticality are discussed in such a manner that even someone not familiar with reactor physics and kinetics can easily follow. A minimum of equations are used and several tables and graphs illustrate the text. (author)

  10. Pneumatic transport systems for TRIGA reactors

    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 research reactors with higher power density

    The recent trend in new or upgraded research reactors is to higher power densities (hence higher neutron flux levels) but not necessarily to higher power levels. The TRIGA LEU fuel with burnable poison is available in small diameter fuel rods capable of high power per rod (∼48 kW/rod) with acceptable peak fuel temperatures. The performance of a 10-MW research reactor with a compact core of hexagonal TRIGA fuel clusters has been calculated in detail. With its light water coolant, beryllium and D2O reflector regions, this reactor can provide in-core experiments with thermal fluxes in excess of 3 x 1014 n/cm2·s and fast fluxes (> 0.1 MeV) of 2 x 1014 n/cm2·s. The core centerline thermal neutron flux in the D2O reflector is about 2 x 1014 n/cm2·s and the average core power density is about 230 kW/liter. Using other TRIGA fuel developed for 25-MW test reactors but arranged in hexagonal arrays, power densities in excess of 300 kW/liter are readily available. A core with TRIGA fuel operating at 15-MW and generating such a power density is capable of producing thermal neutron fluxes in a D2O reflector of 3 x 1014 n/cm2·s. A beryllium-filled central region of the core can further enhance the core leakage and hence the neutron flux in the reflector. (author)

  12. Monte Carlo modelling of TRIGA research reactor

    El Bakkari, B., E-mail: bakkari@gmail.co [Reactor Operating Unit (UCR), National Centre of Sciences, Energy and Nuclear Techniques (CNESTEN/CENM), POB 1382, Rabat (Morocco); ERSN-LMR, Department of Physics, Faculty of Sciences, POB 2121, Tetuan (Morocco); Nacir, B. [Reactor Operating Unit (UCR), National Centre of Sciences, Energy and Nuclear Techniques (CNESTEN/CENM), POB 1382, Rabat (Morocco); El Bardouni, T. [ERSN-LMR, Department of Physics, Faculty of Sciences, POB 2121, Tetuan (Morocco); El Younoussi, C. [Reactor Operating Unit (UCR), National Centre of Sciences, Energy and Nuclear Techniques (CNESTEN/CENM), POB 1382, Rabat (Morocco); ERSN-LMR, Department of Physics, Faculty of Sciences, POB 2121, Tetuan (Morocco); Merroun, O. [ERSN-LMR, Department of Physics, Faculty of Sciences, POB 2121, Tetuan (Morocco); Htet, A. [Reactor Technology Unit (UTR), National Centre of Sciences, Energy and Nuclear Techniques (CNESTEN/CENM), POB 1382, Rabat (Morocco); Boulaich, Y. [Reactor Operating Unit (UCR), National Centre of Sciences, Energy and Nuclear Techniques (CNESTEN/CENM), POB 1382, Rabat (Morocco); ERSN-LMR, Department of Physics, Faculty of Sciences, POB 2121, Tetuan (Morocco); Zoubair, M.; Boukhal, H. [ERSN-LMR, Department of Physics, Faculty of Sciences, POB 2121, Tetuan (Morocco); Chakir, M. [EPTN-LPMR, Faculty of Sciences, Kenitra (Morocco)

    2010-10-15

    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({alpha}, {beta}) 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

    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.

  14. Monte Carlo modelling of TRIGA research reactor

    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.

  15. Decommissioning of TRIGA Mark II type reactor

    The first research reactor in Korea, KRR 1, is a TRIGA Mark II type with open pool and fixed core. Its power was 100 kWth at its construction and it was upgraded to 250 kWth. Its construction was started in 1957. The first criticality was reached in 1962 and it had been operated for 36,000 hours. The second reactor, KRR 2, is a TRIGA Mark III type with open pool and movable core. These reactors were shut down in 1995, and the decision was made to decommission both reactors. The aim of the decommissioning activities is to decommission the KRR 2 reactor and decontaminate the residual building structures and site, and to release them as unrestricted areas. The KRR 1 reactor was decided to be preserve as a historical monument. A project was launched for the decommissioning of these reactors in 1997, and approved by the regulatory body in 2000. A total budget for the project was 20.0 million US dollars. It was anticipated that this project would be completed and the site turned over to KEPCO by 2010. However, it was discovered that the pool water of the KRR 1 reactor was leaked into the environment in 2009. As a result, preservation of the KRR 1 reactor as a monument had to be reviewed, and it was decided to fully decommission the KRR 1 reactor. Dismantling of the KRR 1 reactor takes place from 2011 to 2014 with a budget of 3.25 million US dollars. The scope of the work includes licensing of the decommissioning plan change, removal of pool internals including the reactor core, removal of the thermal and thermalizing columns, removal of beam port tubes and the aluminum liner in the reactor tank, removal of the radioactive concrete (the entire concrete structure will not be demolished), sorting the radioactive waste (concrete and soil) and conditioning the radioactive waste for final disposal, and final statuses of the survey and free release of the site and building, and turning over the site to KEPCO. In this paper, the current status of the TRIGA Mark-II type reactor

  16. Decommissioning of TRIGA Mark II type reactor

    Hwang, Dooseong; Jeong, Gyeonghwan; Moon, Jeikwon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    The first research reactor in Korea, KRR 1, is a TRIGA Mark II type with open pool and fixed core. Its power was 100 kWth at its construction and it was upgraded to 250 kWth. Its construction was started in 1957. The first criticality was reached in 1962 and it had been operated for 36,000 hours. The second reactor, KRR 2, is a TRIGA Mark III type with open pool and movable core. These reactors were shut down in 1995, and the decision was made to decommission both reactors. The aim of the decommissioning activities is to decommission the KRR 2 reactor and decontaminate the residual building structures and site, and to release them as unrestricted areas. The KRR 1 reactor was decided to be preserve as a historical monument. A project was launched for the decommissioning of these reactors in 1997, and approved by the regulatory body in 2000. A total budget for the project was 20.0 million US dollars. It was anticipated that this project would be completed and the site turned over to KEPCO by 2010. However, it was discovered that the pool water of the KRR 1 reactor was leaked into the environment in 2009. As a result, preservation of the KRR 1 reactor as a monument had to be reviewed, and it was decided to fully decommission the KRR 1 reactor. Dismantling of the KRR 1 reactor takes place from 2011 to 2014 with a budget of 3.25 million US dollars. The scope of the work includes licensing of the decommissioning plan change, removal of pool internals including the reactor core, removal of the thermal and thermalizing columns, removal of beam port tubes and the aluminum liner in the reactor tank, removal of the radioactive concrete (the entire concrete structure will not be demolished), sorting the radioactive waste (concrete and soil) and conditioning the radioactive waste for final disposal, and final statuses of the survey and free release of the site and building, and turning over the site to KEPCO. In this paper, the current status of the TRIGA Mark-II type reactor

  17. Operation experience with the TRIGA reactor Wien

    The TRIGA Mark-II reactor Wien is now in operation for more than 38 years. The average operation time is about 230 days per year with 90% of this time at nominal power of 250 kW. The remaining 10% operation time is used for students' training courses at low power level. Pulse operation is rather infrequent with about 5 to 10 pulses per year. The utilization of this facility is excellent. All experimental facilities are intensively used, therefore, neither from a technical nor from an economical and utilization viewpoint a need for decommissioning is necessary and it is intended to operate the reactor as long as possible into the next decade. The on-going US fuel return program has been discussed with the Regulatory Body and the authority's viewpoint is to return the nine HEU fuel elements at present installed in the core and to continue reactor operation beyond 2006 only with LEU standard TRIGA fuel. All components and systems are reinspected following an elaborate reinspection program. This consumes about 4 man-days per month. Once a year all the reactor systems are inspected in presence of an expert nominated by the regulatory body and his expertise is the basis for the annual renewal of the operation license valid again for the coming year. This annual inspection requires approximately 1 man-month (four persons for two weeks). Some of the inspection methods have been successfully applied in other TRIGA reactors. The paper has the following structure: - 1. Introduction; - 2. Status of Main Reactor Systems; - 2.1 Instrumentation; - 2.2 Fuel Elements; - 2.3 Cooling Circuits; - 2.4 Ventilation System; - 2.5 Area Monitoring System; - 2.6 Reinspection and Maintenance Program; - 3. Summary and Outlook

  18. Safety Management at PUSPATI TRIGA Reactor (RTP)

    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)

  19. Update of recent TRIGA reactor projects

    Recent activities related to several TRIGA projects are reported, including reactor upgrades, fuel processing, and instrumentation and control system upgrades at existing research reactors. The installations reported include: Instituto de Asuntos Nucleares, Bogota, Colombia; Centre National de l'Energie des Sciences et des Techniques Nucleaires, Rabat, Morocco; Institute for Nuclear Research, Pitesti, Romania. New computer-based digital instrumentation and control systems are in various stages of completion at: - Imperial Chemical Industries, Billingham, England; - University of Illinois, Urbana, Illinois; - National Tsing Hua University, Hsinchu, Taiwan; - Research Centre for Nuclear Techniques, Bandung, Indonesia

  20. Decontamination of TRIGA Mark II reactor, Indonesia

    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. Stack Monitoring System At PUSPATI TRIGA Reactor

    This paper describes the current Stack Monitoring System at PUSPATI TRIGA Reactor (RTP) building. A stack monitoring system is a continuous air monitor placed at the reactor top for monitoring the presence of radioactive gaseous in the effluent air from the RTP building. The system consists of four detectors that provide the reading for background, particulate, Iodine and Noble gas. There is a plan to replace the current system due to frequent fault of the system, thus thorough understanding of the current system is required. Overview of the whole system will be explained in this paper. Some current results would be displayed and moving forward brief plan would be mentioned. (author)

  2. United States Domestic Research Reactor Infrastructure TRIGA Reactor Fuel Support

    The purpose of this technical paper is to provide status of the United State domestic Research Reactor Infrastructure (RRI) Program at the Idaho National Laboratory. This paper states the purpose of the program, lists the universities operating TRIGA reactors that are supported by the program, identifies anticipated fresh fuel needs for the reactor facilities, discusses spent fuel activities associated with the program, and addresses successes and planned activities for the program. (author)

  3. Industrial and commercial applications for a Triga reactor

    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)

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

    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

  5. Utilisation of the Research Reactor TRIGA Mainz

    The TRIGA Mark II reactor of the University of Mainz can be operated in the steady state mode with thermal powers up to a maximum of 100 kW and in the pulse mode with a maximum peak power of 250 MW. So far, more than 17 000 pulses have been performed. For irradiations the TRIGA Mainz has a central experimental tube, three pneumatic transfer systems and a rotary specimen rack. In addition, the TRIGA Mainz includes four horizontal beam ports and a graphite thermal column which provides a source of well-thermalised neutrons. A broad spectrum of commercial applications, scientific research and training can be executed. For education and training various courses in nuclear and radiochemistry, radiation protection, reactor operation and physics are held for scientists, advanced students, teachers, engineers and technicians. Isotope production and Neutron Activation Analysis (NAA) are applied in in-core positions for different applications. NAA in Mainz is focused to determine trace elements in different materials such as in archaeometry, forensics, biology and technical materials including semiconductors for photovoltaics. The beam ports and the thermal column are used for commercial as well as for special basic and applied research in medicine, biology, chemistry and physics. Experiments are in preparation to determine the fundamental neutron properties with very high precision using ultra cold neutrons (UCN) produced at the tangential beam port. A second source is under development at the radial piercing beam port. Another experiment under development is the determination of ground-state properties of radioactive nuclei with very high precision using a penning trap and collinear laser spectroscopy. For many years fast chemical separation procedures combining a gas-jet transport system installed in one beam tube with either continuous or discontinuous chemical separation are carried out. In addition the thermal column of the reactor is also used for medical and

  6. Triga Mark III Reactor in paleotemperatures determination

    The Triga Mark III reactor produces neutron fluxes which are used to irradiate geologic specimens with age estimation purposes. Irradiation produces radioactive nucleus in the sample as well as: 39 Ar used in the age estimation 40 Ar/39 Ar ratio, and fission fragments for the age estimation by fission tracks detection. This document presents the basis for both methods, as well as the attained results, and has the purpose to perform joint experimentation in order to extend the usefulness of the method to paleotemperature determination. A brief comment about the associated problematic of the sample irradiation is made

  7. Nondestructive examination of TRIGA reactor fuel elements

    Neutron radiography has proved to be a very useful method for nondestructive examination of used and nonused reactor elements. The method can be used for determination of homogenity and burn-up of fuel and burnable poisons, for detection of fuel and full clad damage and taking into account the capability to perform accurate geometrical measurements it is also possible to assess mechanical deformations of fuel elements. Active fuel elements of TRIGA reactor have been examined for deformations and fuel clad damage. In the course of these investigations the following methods were tested and compared: - transfer neutronradiographic techniques using In and Dy converter screens, - direct neutrongraphic method using solid state track detectors, - X-ray radiography employing lead shielding masks and highly selective photographic material. Considerable information on the burn-up of reactor fuel elements can be obtained from measuring the distribution of radioactive isotopes in the fuel element by gamma ray spectroscopy. For a used TRIGA fuel element the axial distribution of the isotope Cs-137 has been measured and the burn-up determined. We compare the experimental results with a crude estimate of burn-up

  8. Modeling the PUSPATI TRIGA Reactor using MCNP code

    The 1 MW TRIGA MARK II research reactor at Malaysian Nuclear Agency achieved initial criticality on June 28, 1982. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes. This paper describes the reactor parameters calculation for the PUSPATI TRIGA REACTOR (RTP); focusing on the application of the developed reactor 3D model for criticality calculation, analysis of power and neutron flux distribution and depletion study of TRIGA fuel. The 3D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA reactor. The model represents in detailed all important components of the core and shielding with literally no physical approximation. (author)

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

    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

  10. Fuel element situation and performance data TRIGA Mark II reactor

    Electronic data acquisition of the position and movement of Triga fuel elements (FE) in the TRIGA II Vienna reactor was the objective of this project. Using one month power data and the Fuel element position in core it is possible to calculate their burnup. Fuel element performance data during 1962 to 2003 are provided. (nevyjel)

  11. Neutron flux measurements in PUSPATI Triga Reactor

    Neutron flux measurement in the PUSPATI TRIGA Reactor (PTR) was initiated after its commissioning on 28 June 1982. Initial measured thermal neutron flux at the bottom of the rotary specimen rack (rotating) and in-core pneumatic terminus were 3.81E+11 n/cm2 sec and 1.10E+12n/cm2 sec respectively at 100KW. Work to complete the neutron flux data are still going on. The cadmium ratio, thermal and epithermal neutron flux are measured in the reactor core, rotary specimen rack, in-core pneumatic terminus and thermal column. Bare and Cadmium covered gold foils and wires are used for the above measurement. The activities of the irradiated gold foils and wires are determined using Ge(Li) and hyperpure germinium detectors. (author)

  12. The TRIGA reactor as chemistry apparatus

    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)

  13. TRIGA reactor dynamics: Frequency response tests

    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)

  14. The construction, installation and commissioning of the PUSPATI TRIGA reactor

    A TRIGA Mark II research reactor has been installed at the Tun Ismail Atomic Research Centre (PUSPATI), Selangor, Malaysia. The reactor was commissioned in July 1982. With the commissioning of the reactor, a new era in the development of nuclear science and technology in Malaysia has just begun. This report describes the construction, installation and commissioning of the reactor. (author)

  15. Oregon State TRIGA reactor power calibration study

    As a result of a recent review of the Oregon State TRIGA Reactor (OSTR) power calibration procedure, an investigation was performed on the origin and correctness of the OSTR tank factor and the calibration method. It was determined that there was no clear basis for the tank factor which was being used (0.0525 deg. C/kwh) and therefore a new value was calculated (0.0493 deg. C/kwh). The calculational method and likely errors are presented in the paper. In addition, a series of experimental tests were conducted to decide if the power calibration was best performed with or without a mixer, at 100 KW or at 1 MW. The results of these tests along with the final recommendation are presented. (author)

  16. TRIGA Mark-III reactor dismantling program

    The activation assessment of the main parts of the TRIGA Mark-III (KRR-2) was estimated to effectively dismantle the activated and contaminated areas. All of the method and the order for decommissioning the KRR-2 have been chosen as a result of the examination of the physical structure and radiological conditions of the reactor component. These decommissioning methods and orders were reviewed as part of the Hazard and Operability (HAZOP) studies for the project. Radiological assessment is also done to protect the workers and the environment from the dismantling work. License documents were submitted to the Ministry of Science and Technology (MOST) at the end of 1998. Practical work of the D and D will start at the end of 1999 once the government issues the license. Radiation protection plan was also set up to control the workers and environment. This paper summarized the main lines of those studies. (author)

  17. Reconditioning of the TRIGA Mark III reactor

    The paper describes the activities carried out to recondition the TRIGA Mark III reactor at the Mexican Nuclear Centre, namely repair of its containment system, maintenance of its operational systems, and the obtaining of a licence for the facility and its operating staff. The process of initially obtaining the operating licence from the regulatory authority was affected by the existence of water leaks in the pool which were detected in March 1985 and were caused by corrosion in the reactor containment system. Reconditioning began with a series of activities aimed at locating, delimiting and repairing the areas damaged by corrosion and involved establishing criteria for selecting the most appropriate inspection, testing and repair methods. In order to obtain the operating licence, it was necessary to comply with various requirements laid down by the regulatory body. The most important requirements included: (a) repair of the reactor pool; (b) maintenance of its operational systems; (c) preparation and implementation of the Quality Control Programme; (d) updating of the Safety Report; (e) updating and preparation of operating, repair, radiation safety, emergency and administrative procedures; and (f) training of operating staff. In addition, the paper describes the work carried out at this reactor to widen its field of research and range of utilization. This work includes the reconditioning of a neutron diffractometer, the design and construction of a neutron diffractometer to determine the textures of materials, and the analysis of a new mixed core configuration based on fuels with 20% and 70% 235U enrichment. (author). 7 refs

  18. Ageing management for reactor TRIGA PUSPATI

    The probability of a component, system or structure failure resulting from ageing degradation normally increases with the time of exposure to service condition unless countermeasures are taken. The objective of the management of ageing is to determine and apply these countermeasures. The Reactor TRIGA PUSPATI ageing management includes activities such as protection, repair, refurbishment or replacement, which are similar to other activities carried out at a reactor facility during routine maintenance or when a modification project takes place. However, it is important to distinguish between these different activities, because the management of ageing requires the use of methodology which will detect and evaluate deficiencies produced by the service conditions and will lead to the application of countermeasures for prevention and mitigation of the deficiencies. One approach to this methodology is a determination that the reactor systems and components can perform their safety functions during their service life and under the service conditions. This can be achieved through appropriately selecting systems and components which should be included in long term surveillance program, through data collection and through evaluation of the potential ageing effects. The above activities will be followed by countermeasures for prevention and mitigation of the ageing effects to ensure an adequate level of safety for the reactor facility. (author)

  19. Current activities at the Finnish TRIGA reactor

    The FiR 1 reactor, a 250 kW TRIGA reactor, with its subsystems has experienced a large renovation work. The main purpose of the upgrading has been to install the new Boron Neutron Capture Therapy (BNCT) irradiation facility. The epithermal neutrons are produced from the fast fission neutrons by a moderator block consisting of Al+AlF3 (FLUENTAL), which showed to be the optimum material for this purpose. The BNCT work dominates the current utilisation of the reactor: four days per week for BNCT purposes and only one day per week for neutron activation analysis and isotope production. The first ten patients have already been irradiated during a period of about twelve months. The Council of State (government) granted at the end of last year a new operating license for the reactor for twelve years. There is a special condition in the new license. One has now about four years' time to achieve a binding agreement between VTT and the Nuclear Power Plants about the possibility to use the final disposal facility of the Nuclear Power Plants for the spent fuel. If this will not happen, one intends to use the USDOE alternative with the well-known time limits. Recently it was started a project to study the possibilities and limitations to increase the power of the reactor to 500 kW or more. In the BNC Therapy in some cases there is the need to increase the penetration depth of the neutrons. This can be arranged by filtering low energy neutrons away from the epithermal beam. The only way to compensate the loss of neutron intensity caused by the filter is to increase the power of the reactor. (authors)

  20. Evaluation of TRIGA Mark II reactor in Turkey

    There are two research reactors in Turkey and one of them is the university Triga Mark II reactor which was in service since 1979 both for education and industrial application purposes. The main aim of this paper is to evaluate the spectrum of the services carried by Turkish Triga Mark II reactor. In this work, statistical distribution of the graduate works and applications, by using Triga Mark II reactor is examined and evaluated. In addition to this, technical and scientific uses of this above mentioned reactor are also investigated. It was already showed that the uses and benefits of this reactor can not be limited. If the sufficient work and service is given, NDT and industrial applications can also be carried economically. (orig.)

  1. Analysis of fuel options in TRIGA reactor

    In this paper, nuclear characteristics of TRIGA Mark-III has been analyzed in detail for six different fuel options. Presently, 70w/o enriched FLIP fuels are adopted for TRIGA core to improve fuel lifetime. However, such highly enriched fuels are not easily obtained due to nonproliferation treaty. This research examines the possible substitution for FLIP fuels with high density fuels without reducing the nuclear performance. This work will provide long-time plan for TRIGA operation (author)

  2. Development of PUSPATI TRIGA Reactor Simulator For Education Purpose

    The PUSPATI TRIGA Reactor Simulator was developed in 2012 purposely for 3V program. It is an interactive tool for students understand how to operate the PUSPATI TRIGA Reactor. Students can feeling and understand how to operate the reactor from start-up to shut-down in manual mode of operation. Movement of control rod and reactor parameters are displaying in interactively. Behavior and characteristic for reactor console and reactor itself can be evaluated and understand. Implementation of human system interface is using computer screens, keyboard and mouse. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate and estimated reactor console parameters. The capabilities in user interface, reactor physics and thermal-hydraulics can be expanded and explored to simulation as well as modeling for New Reactor Console, Research Reactor and Nuclear Power Plant. (author)

  3. Fuel experience at a 37 year old TRIGA type reactor

    Boeck, H. [Atominstitut der Oesterreichischen Universitaeten, Wien (Austria)

    1999-07-01

    A survey is given on 37 years of TRIGA fuel experience at the 250 kW TRIGA Mark II reactor Vienna. Approximately 3000 fuel-years of experience have accumulated at this facility with only minor problems. Totally only 8 fuel elements had to be removed permanently from the core. Various inspection methods which have been developed throughout the years are described in this paper. (author)

  4. A 5 MW TRIGA reactor design for radioisotope production

    The production and preparation of commercial-scale quantities of radioisotopes has become an important activity as their medical and industrial applications continue to expand. There are currently various large multipurpose research reactors capable of producing ample quantities of radioisotopes. These facilities, however, have many competing demands placed upon them by a wide variety of researchers and scientific programs which severely limit their radioisotope production capability. A demonstrated need has developed for a simpler reactor facility dedicated to the production of radioisotopes on a commercial basis. This smaller, dedicated reactor could provide continuous fission and activation product radioisotopes to meet commercial requirements for the foreseeable future. The design of a 5 MW TRIGA reactor facility, upgradeable to 10 MW, dedicated to the production of industrial and medical radioisotopes is discussed. A TRIGA reactor designed specifically for this purpose with its demonstrated long core life and simplicity of operation would translate into increased radioisotope production. As an example, a single TRIGA could supply the entire US needs for Mo-99. The facility is based on the experience gained by General Atomics in the design, installation, and construction of over 60 other TRIGAs over the past 35 years. The unique uranium-zirconium hydride fuel makes TRIGA reactors inexpensive to build and operate, reliable in their simplicity, highly flexible due to unique passive safety, and environmentally friendly because of minimal power requirements and long-lived fuel. (author)

  5. Research work at the TRIGA Mainz reactor

    In the last two years the research activities at the TRIGA Mark II reactor in Mainz have mainly been concentrated on the investigation of short- lived nuclides of medium mass number produced by thermal-neutron induced fission of 235U and other fissile materials. For the identification of these nuclides and for detailed studies of their properties rapid chemical separation procedures in combination with high-resolution gamma-ray and neutron spectroscopy as well as mass-separated samples have been used. Fast, discontinuous separation techniques are illustrated by a procedure for technetium. Continuous separation methods from aqueous solutions and in the gas phase, accomplished by combining a gas jet recoil transport system with an on-line operating solvent extraction technique and a thermo- chromatographic method, are presented. The application of such procedures to decay scheme and delayed neutron studies is demonstrated by a few examples. The experimental set-up and the method for nuclear spin - and magnetic moment measurements on alkali isotopes far from the region of beta-stability applying the nuclear radiation detected optical pumping technique to mass- separated samples of neutron-rich alkali nuclides are briefly described. (author)

  6. Small Angle Neutron Scattering instrument at Malaysian TRIGA reactor

    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)

  7. Irradiation routine in the IPR-R1 Triga reactor

    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.)

  8. Irradiation behaviour of TRIGA-LEU fuel in the TRIGA 14 MW reactor facility

    In order to convert TRIGA reactors to low enriched uranium fuel cycle, General Atomic (USA) produces a new type of fuel elements (U 235 enrichment -19.7%). A part of the TRIGA 14 MW (th) core in Pitesti consists of this type of fuel, and six elements have been examined in our post-irradiation facility. These were the first measurements using non-destructive control on this type of fuel. In this paper, some interesting information about the fuel behaviour during irradiation, is presented. (Author)

  9. Education and Training Programme at the Research Reactor TRIGA Mainz

    Hampel, Gabriele; Eberhardt, Klaus [University of Mainz, Institute for Nuclear Chemistry, D-55099 Mainz (Germany)

    2011-07-01

    Education and training are important elements for the future of nuclear science, technology and safety. Fields of interest include high- technology applications in nuclear techniques and neutron sources, advances in the areas of power reactor safety, establishing the scientific basis of new reactors, training of personnel needed to operate, maintain, regulate and improve reactors or other facilities associated with nuclear power. Also, creating a knowledgeable public through education usually means less opposition and more support. Education and training for safeguards, operators, researchers and quality programmes (calibration services, etc.) are one of the main utilisations of TRIGA research reactors. Use of a reactor as a training tool for university students studying nuclear engineering and/or physics, where there is a growing demand at European Universities, is of vital importance. In particular, the TRIGA Mark II reactor, located at the University of Mainz, one of the largest universities in Germany, offers a broad range of nuclear-related courses for training and education. (author)

  10. Education and Training Programme at the Research Reactor TRIGA Mainz

    Education and training are important elements for the future of nuclear science, technology and safety. Fields of interest include high- technology applications in nuclear techniques and neutron sources, advances in the areas of power reactor safety, establishing the scientific basis of new reactors, training of personnel needed to operate, maintain, regulate and improve reactors or other facilities associated with nuclear power. Also, creating a knowledgeable public through education usually means less opposition and more support. Education and training for safeguards, operators, researchers and quality programmes (calibration services, etc.) are one of the main utilisations of TRIGA research reactors. Use of a reactor as a training tool for university students studying nuclear engineering and/or physics, where there is a growing demand at European Universities, is of vital importance. In particular, the TRIGA Mark II reactor, located at the University of Mainz, one of the largest universities in Germany, offers a broad range of nuclear-related courses for training and education. (author)

  11. TRIGA mark-II,III reactor safety re-evaluation

    For two years of 1990 and 1991, the safety of TRIGA Mk-II and III reactor has been re-evaluated. For this, domestic rules on research reactors has been reviewed, and as it was judged that standards on research reactors in USA is applicable to our ones it was evaluated whether TRIGA Mk-II and III reactors satisfy these standards. The site parameters and the environmental impacts during normal operation and hypothetical accident conditions have been analysed, and those parts for reactor facility and structure have been rewritten to fit SAR standard format based on the review of old SAR and maintenance manuals reflecting changes after the construction. Based on this re-evaluation, SAR, Technical Specifications, Radiation Emergency Plan, Environment Report, various procedures,etc. will be amended by the reactor management project. (Author)

  12. Operation experience operation experience with the TRIGA Reactor Wien

    The TRIGA Reactor Wien is the Closest Nuclear Facility to the IAEA. It is involved in: development of safeguards instrumentation, prevention of illicit trafficking, calibration of nuclear instrumentation, irradiation and test of safeguards instrumentation, storage of special nuclear material, training courses for junior inspectors (more than 120 trained and since 1992 more than 100 IAEA fellows from developing countries). The TRIGA Mark II Reactor Vienna is the only operating research reactor and the only nuclear facility in Austria, uniquely used for training and education of students and junior professionals in the fields of: nuclear technology, neutron and solid state physics, radiochemistry, radiation protection and dosimetry, low temperature physics, nuclear- and nuclear astrophysics, electron- and x-ray physics. The main technical data of the TRIGA Mark-II Reactor are reviewed as well as the operation experience during the 2004-2008 period: Visual inspection of beam tubes A (piercing) and D (radial), MCNP core calculations, investigation of shielding concrete for trace elements, estimation of radiation exposure during dismantling, replacement of both monitors at the console, problems with the NM-1000 wide range channel, Noise pick-up by nm-1000 due to grounding problems, Change of core configuration: 6 FLIP fuel elements transferred from C-ring into B-ring. The concrete studies at the TRIGA Vienna include: the determination of long-lived radionuclides in heavy concrete (mainly Ba-133, Eu 133, Eu-152, Eu-154, Co 154, Co-60), the measurement of the composition of heavy concrete, the estimation of the neutron attenuation in heavy concrete: aim is to establish a model and to predict the mass and activity of activated concrete in the Vienna TRIGA shield in view of future dismantling, the validation of model using the data from the dismantled 10 MW ASTRA reactor at Seibersdorf. In conclusions: after 50 years of successful TRIGA reactor operation (May 3, 1958) there

  13. Operation experience with the TRIGA reactor Wien 2004

    The TRIGA Mark-II reactor in Vienna is now in operation for more than 42 years. The average operation time is about 230 days per year with 90 % of this time at nominal power of 250 kW. The remaining 10 % operation time is used for students' training courses at low power level. Pulse operation is rather infrequent with about 5 to 10 pulses per year. The utilization of this facility is excellent, the number of students participating in practical exercises has strongly increased, and also training courses for outside groups such as the IAEA or for the 2004 Eugene Wigner Course are using the reactor, because it is the only TRIGA reactor remaining in Austria. Therefore, there is no need for decommissioning and it is intended to operate it as long as possible into the next decade. Nevertheless, in early 2004 it was decided to prepare a report on a decommissioning procedure for a typical TRIGA Mark II reactor which lists the volumes, the activity and the weight of individual materials such as concrete, aluminium, stainless steel, graphite and others which will accumulate during this process (a summary of possible activated and contaminated materials and the activity of a single TRIGA fuel element as a function of fuel type and decay time in Bq is presented). The status of the reactor (instrumentation, fuel elements, cooling circuit, ventilation system, re-inspection and maintenance program, cost/benefit) is outlined. (nevyjel)

  14. Verification of the MCNP model for the University of Texas TRIGA reactor

    An MCNP model of The University of Texas TRIGA reactor has been used for design calculations for the neutron collimator system in the through beam port. The TRIGA MCNP model was verified by comparing its results with experimentally determined values

  15. Component and operation experience of reactor TRIGA MARK II

    Reactor TRIGA MARK II is Jozef Stefan Institute's research reactor. It has been operating since 1966. A probabilistic approach of reactor safety estimation was used first in 1989 when a Probabilistic Safety Analysis (PSA) of the reactor was performed. A lack of reactor component data was found as the major problem in probabilistic assessment. It was decided to continue the work with specific data base development. The project has been divided in two phases. In the first phase specific data from year 1985 to 1990 were collected. In the second phase the collected data were treated. The comparison of generic and specific data showed significant difference between the generic and specific data and leads to a conclusion that a generic data based PSA has a limited credibility indicating that there is a need to build a specific data base for research reactors. The TRIGA MARK II research reactor has three major purposes: operator training, research involving neutrons and isotope production. The paper represents specific data base formation for TRIGA MARK II research reactor in Podgorica. Specific data on reactor scrams, components operation and human errors were collected. The data of fifteen components were estimated by classical and Bayesian method. The results of both methods are very different. Because of good specific data the results of classical methods were preferred. The comparison of specific and generic data showed that there is a great need to build a specific data base for research reactors. It is expected to use the specific data for existing PSA of TRIGA MARK II reactor reevaluation and optimisation of its operation. (authors)

  16. 78 FR 26811 - Dow Chemical Company, Dow TRIGA Research Reactor; License Renewal for the Dow Chemical TRIGA...

    2013-05-08

    ...) published a notice in the Federal Register on July 20, 2012 (77 FR 42771), ``License Renewal for the Dow... COMMISSION Dow Chemical Company, Dow TRIGA Research Reactor; License Renewal for the Dow Chemical TRIGA... Facility License No. R-108 for Dow Chemical Company which would authorize continued operation of the...

  17. Reactor calculations for improving utilization of TRIGA reactor

    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)

  18. Triga mark-II,III reactor safety re-evaluation

    In order to revise safety analysis report of old TRIGA reactors, safety re-evaluation of these reactor was started for necessary parts. This report contains the first year results of the project scheduled for two years. The guide lines of safety re-evaluation was made by translating that of nuclear power plant from the view point of TRIGA reactor confirming the basic safety philosophy as much as possible. First of all, sections of reactor history and comparison with similar reactors are made, since the actual operation records, changes, any modification of similar reactors constructed after then, etc., are realistic and valuable data from the safety aspect of old reactor. For the effectiveness of nuclear analysis, a PC based analysis system using WIMS-D/4 and VENTURE was established, and a program for the natural convection cooling analysis of TRIGA reactor was developed. As a result of thermal-hydraulic analysis it was confirmed that the operation limit of fuel temperature set at 650 deg C without any logical reason is very close to the DNB limit. (Author)

  19. Reactor TRIGA PUSPATI (RTP) spent fuel pool conceptual design

    Reactor TRIGA PUSPATI (RTP) is the one and only research reactor in Malaysia that has been safely operated and maintained since 1982. In order to enhance technical capabilities and competencies especially in nuclear reactor engineering a feasibility study on RTP power upgrading was proposed to serve future needs for advance nuclear science and technology in the country with the capability of designing and develop reactor system. The need of a Spent Fuel Pool begins with the discharge of spent fuel elements from RTP for temporary storage that includes all activities related to the storage of fuel until it is either sent for reprocessed or sent for final disposal. To support RTP power upgrading there will be major RTP systems replacement such as reactor components and a new temporary storage pool for fuel elements. The spent fuel pool is needed for temporarily store the irradiated fuel elements to accommodate a new reactor core structure. Spent fuel management has always been one of the most important stages in the nuclear fuel cycle and considered among the most common problems to all countries with nuclear reactors. The output of this paper will provide sufficient information to show the Spent Fuel Pool can be design and build with the adequate and reasonable safety assurance to support newly upgraded TRIGA PUSPATI TRIGA Research Reactor. (author)

  20. Isothermal temperature reactivity coefficient measurement in TRIGA reactor

    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 oC and 25 oC. 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)

  1. On-line coupling of the TRIGA-SPEC facility at the research reactor TRIGA Mainz

    To determine ground-state properties of exotic nuclides, the TRIGA-SPEC experiment at the TRIGA Mainz research reactor was recently installed. It includes the Penning-trap mass spectrometer TRIGA-TRAP and the collinear laser spectroscopy setup TRIGA-LASER. Nuclides of interest are produced via neutron-induced fission of suitable actinide isotopes, thermalized in a gas-filled volume and transported with a gas-jet system to an on-line ion source. Ionization of the fission products occurs inside a hot cavity of the ion source, which is heated by electron bombardment to temperatures of about 2000 C. The ion beam is extracted by a high potential difference and mass separated by a 90 dipole magnet. Afterwards, the ion beam is injected into an RF-cooler/buncher and finally decelerated by a pulsed drift tube so that the ions can be captured in a Penning trap. The efficiencies of the different parts of the beamline were tested recently, and the latest results about the performance are presented.

  2. Outlook on radioisotope production at TRIGA SSR 14 MW reactor

    INR Pitesti, endowed with a research nuclear reactor of TRIGA SSR 14 MW type, has developed activities of radioisotope production, being at present licensed for production and selling Ir-192 sources for industrial gamma radiography and Co-60 sources (2,000 Ci) for medical uses (cobalto therapy). A collaboration was initiated with the CPR Department of IFIN-HH Bucharest, particularly after the WWR-S reactor shutdown on December 21, 1997. In the frame of this program the INR Pitesti offers services of raw material irradiations followed by the radioisotope production performed subsequently at the Radioisotope Production Department (CPR) of IFIN-HH Bucharest which also deals with selling the product on internal market . The experimental facilities with the two TRIGA reactors (TRIGA SSR 14 MW and TRIGA ACPR) of INR Pitesti are described. The maximum neutron flux is 2.9 · 1014 n/cm2s. The irradiation channels are of two neutron spectra types. Also the neutron flux is characterized by radial and axial distribution which are taken into account when a given raw material is to be irradiated, to avoid perturbing non-homogeneities in the raw material activation. Five irradiation devices are presented. Preparations are currently under way for production of fission radioisotopes Mo-99, I-131 and Xe-133 and activation radioisotope I-125 for medical application

  3. Computational analysis of irradiation facilities at the JSI TRIGA reactor.

    Snoj, Luka; Zerovnik, Gašper; Trkov, Andrej

    2012-03-01

    Characterization and optimization of irradiation facilities in a research reactor is important for optimal performance. Nowadays this is commonly done with advanced Monte Carlo neutron transport computer codes such as MCNP. However, the computational model in such calculations should be verified and validated with experiments. In the paper we describe the irradiation facilities at the JSI TRIGA reactor and demonstrate their computational characterization to support experimental campaigns by providing information on the characteristics of the irradiation facilities. PMID:22154389

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

    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

  5. The current status of Bandung Triga Mark II reactor, Indonesia

    Full text: The Bandung TRIGA Mark II Reactor - Indonesia was started-up on October 10, 1964 and it has been operated at power level of 250 kw. The facility has been, operated for research, production of radioisotopes and training. In 1971, the reactor has been upgraded from 250 kw to 1000 kw. Since that time the facility has been safely operating at various power levels of a maximum 1000 kw until February 1996, even though the reactor tank is kept unchanged. For a highly reliable reactor that can back-up the Ga Siwabessy Multipurpose Reactor - Jakarta, Indonesia, in producing sufficient radioisotopes, a higher power reactor is needed. This can be accomplished by increasing the thermal power of current TRIGA Mark II Bandung Reactor to 2000 kw as well as by enhancing the inherent and engineered safety features of the current reactor. The upgrading of reactor power shall ensure the increasing of neutron flux in the beam ports; hence the experiments such as neutron radiography, time of flight spectrometry and other nuclear physic experiments can be conducted better. For that the reactor tank, the number and configuration of fuel element, instrumentation and control rod, primary cooling system, secondary cooling system, water treatment system, shielding, etc. have been changed, and an Emergency Core Cooling System (ECCS) was added. One additional control rod, core configuration modification and enhancement of reactor shielding, shall increase the safety margin so that the reactor could be operated at a maximum power of 2000 kw. At the middle of May 2000 cold test (non-nuclear commissioning) was done, and continued to hot test (nuclear commissioning). Since June 24, 2000 the TRIGA Mark II Bandung has been operated at 2000 kw

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

    Ketelaer, J.; Krämer, J.; Beck, D; Blaum, K.; Block, M.; Eberhardt, K.; Eitel, G.; Ferrer, R.; Geppert, C.; George, S; Herfurth, F.; Ketter, J.; Nagy, Sz.; Neidherr, D.; Neugart, R

    2008-01-01

    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 beam line 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 h...

  7. Thermal - hydraulic analysis of the ITU TRIGA Mark - II reactor

    Experimental and analytical studies have been performed to find out the temperature distribution, as a function of reactor power, in the TRIGA Mark-II reactor at Istanbul Technical University. A two-dimensional computer code was written in FORTRAN-77 language numerically solves heat conduction equation using finite difference method at the steady state. The calculated results for fuel temperature and coolant temperature distribution in the reactor core for different reactor power were compared with the experimental data. Agreements between experiment and results from the computer program are fairly good

  8. Research projects at the TRIGA-reactor Vienna

    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)

  9. Preliminary neutronic design of TRIGA Mark II Reactor

    It is very important to analyse the behaviour of the research reactors, since, they play a key role in developing the power reactor technology and radiation applications such as isotope generation for medical treatments. In this study, the neutronic behaviour of the TRIGA MARK II reactor, owned and operated by Istanbul Technical University is analysed by using the SCALE code system. In the analysis, in order to overcome the disadvantages of special TRIGA codes, such as TRIGAP, the SCALE code system is chosen to perform the calculations. TRIGAP and similar codes have limited geometrical (one-dimensional geometry) and cross sectional options (two-group calculations), however, SCALE has the capability of wider range of geometrical modelling capability (three-dimensional modelling is possible) and multi-group calculations are possible

  10. Reactor TRIGA at the J.Stefan institute in Ljubljana

    The TRIGA Mark II Reactor began its operation on May 1966. The power of the reactor is 250 kW. TRIGA utilizes solid fuel elements in which the zirconium hydride moderator is homogeneously mixed 20% or 70% enriched uranium. The inique featUre of these fuel - moderator elements is the prompt negative temperature coefficient of reactivity, which gives TRIGA its built-in safety. The reactor core consist of a lattice of cylindrical fuel-moderator elements and graphite (dummy) elements at the bottom of the 6 m high tank full of light water which is used for cooling and radiation protection. The reactor has the following experimental and irradiation facilities: 2 radial beam channels, 2 tangential beam channels, 2 thermal colomns, 40 position rotary specimen rack, pneumatic transfer tube and central thimble. The reactor operates about 2.500 hours per year and it is utilized for the production of isotopes, as a source of neutrons for various experiments and for the training of personnel for the nuclear power station in Krsko

  11. Neutronic analysis of the Geological Survey TRIGA Reactor

    Highlights: • We develop a detailed MCNP model of the Geological Survey TRIGA Reactor. • We present a simplified approach to considering burnup. • The model is validated against available reactor data. • We present evidence of inaccuracies in the ENDF B/VII zirconium libraries. - Abstract: The United States Geological Survey TRIGA Reactor (GSTR) is a 1 MW reactor located in Lakewood, Colorado. In support of the GSTR’s relicensing efforts, this project developed and validated a Monte Carlo N-Particle Version 5 (MCNP5) model of the GSTR reactor. The model provided estimates of the excess reactivity, power distribution and the fuel temperature, water temperature, void, and power reactivity coefficients for the current and limiting core. The MCNP5 model predicts a limiting core excess reactivity of $6.48 with a peak rod power of 22.2 kW. The fuel and void reactivity coefficients for the limiting core are strongly negative, and the core water reactivity coefficient is slightly positive, consistent with other TRIGA analyses. The average fuel temperature reactivity coefficient of the full power limiting core is −0.0135 $/K while the average core void coefficient is −0.069 $/K from 0% to 20% void. The core water temperature reactivity coefficient is +0.012 $/K

  12. Epithermal BNCT neutron beam design for a TRIGA II reactor

    In Finland a collaborative effort by Helsinki University Central Hospital, MAP Medical Technologies Inc. and VTT Reactor Laboratory has started aiming at BNCT of glioma patients. For this the capabilities of the FiR-1 TRIGA II 250 kW research reactor have been evaluated. The FiR-1 is located in the middle of the Otaniemi campus eight kilometers from the center of Helsinki and four kilometers from the Central Hospital. The power of the reactor was increased in 1965 to 250 kW and the instrumentation modernised in 1981. It is a pool reactor with graphite reflector and a core loading of 3 kg 20w% 235U in the special TRIGA uranium-zirconium hydride fuel (8-12 w% U, 91% Zr, 1% H). The advantages of using a TRIGA reactor for BNCT have already been pointed out earlier by Whittemore and have been verified in practice by the thermal neutron treatment work done at the Musashi 100 kW reactor. The advantages include a wide core face area and a wide spatial angle covered by the thermal-epithermal column system, large flux-per-Watt feature and inherent safety of the TRIGA fuel. Because of its wider applicability and less stringent requirements for clinical operation conditions, an epithermal neutron beam has been selected as the design goal. The epithermal flux should be sufficient for glioblastoma patient treatment: 109 epithermal neutrons/cm2/s with low enough fast neutron (-13Gy/epithermal n/cm2) and gamma contamination

  13. The future of spent TRIGA fuel elements from European TRIGA reactor stations

    The paper gives a summary of the information collected and presented to the General Atomics about TRIGA fuel elements available at European TRIGA stations under the initiative to solve the problem of the future of spent TRIGA fuel elements

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

    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. Reactor

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

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

    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)

  17. Significant aspects concerning RC-1 TRIGA reactor operations

    In the present work, a brief critical survey is made of the various nuclear configurations with which the reactor operated after completion of the works directed at obtaining a power increase, from 0.1 to 1 MW, and the reactor utilizations are hinted. Furthermore, through experimental measurements of the neutron flux in significant positions, the volumetric integral of the flux in the core is solved by defining, in conservative conditions of nuclear configuration and of reactor operation, the peak factor, referred to the most stressed element of the TRIGA core. (author)

  18. Fuel elements for pulsed TRIGA research reactors

    TRIGA fuel was developed around the concept of inherent safety. A core composition was sought that had a large prompt negative temperature coefficient of reactivity such that if all the available excess reactivity were suddenly inserted into the core, the resulting fuel temperature would automatically cause the power excursion to terminate before any core damage resulted. Experiments have demonstrated that zirconium hydride possesses a basic neutron-spectrum-hardening mechanism to produce the desired characteristic. Additional advantages include the facts that ZrH has a good heat capacity, that it results in relatively small core sizes and high flux values due to the high hydrogen content, that it has excellent fission-product retentivity and high chemical inertness in water at temperatures up to 1000C, and that it can be used effectively in a rugged fuel element size. Tens of thousands of routine pulses to the range of 500 to 8000C peak fuel temperatures have been performed with TRIGA fuel, and a core was pulse-heated to peak fuel temperatures in excess of 11000C for hundreds of pulses before a few elements exceeded the conservative tolerances on dimensional change

  19. Power stabilization in CREN-K TRIGA Mark II reactor

    In order to eliminate power oscillations in the TRIGA MARK II reactor at the 'Centre Regional d'Etudes Nucleaires de Kinshasa' (CREN-K), Zaire, specially made adapters were put around the control rods in the top grid plate. The paper briefly describes how investigations were made to find out the basic reason of the power oscillations and the way these adapters were conceived and installed. (author)

  20. Accident scenarios of the TRIGA Mark II reactor in Vienna

    The safety report of the TRIGA Mark II reactor in Vienna includes three accident scenarios and their deterministic dose consequences to the environment. The destruction of the cladding of the most activated fuel element, the destruction of all fuel elements and a plane crash were considered scenarios in that report. The calculations were made in 1978 with the software program named STRISK. In this paper, the program package PC Cosyma was applied on the TRIGA Mark II reactor in Vienna and the deterministic consequences of the scenarios to the environment were updated. The fission product inventories of all fuel elements were calculated with ORIGEN2. To get meteorological data of the atmospheric condition around the release area, a weather station was installed. The release parameters were taken from the safety report or were replaced by worst case parameters. This paper focuses on two accident scenarios: the destruction of the cladding of the fuel element with the highest activity content and the case of a large plane crash. The current accident scenarios show good agreement with the calculations from 1978, hence no technical modifications in the safety report of the TRIGA reactor Vienna were necessary. Even in the very worst case scenario - complete destruction of all fuel elements in a large plane crash - the expected doses in the Atominstitut's neighborhood remain moderate.

  1. Accident scenarios of the TRIGA Mark II reactor in Vienna

    Villa, Mario, E-mail: mvilla@ati.ac.a [Vienna University of Technology, Atominstitut, Stadionallee 2, 1020 Wien (Austria); Haydn, Markus [Vienna University of Technology, Atominstitut, Stadionallee 2, 1020 Wien (Austria); Steinhauser, Georg, E-mail: georg.steinhauser@ati.ac.a [Vienna University of Technology, Atominstitut, Stadionallee 2, 1020 Wien (Austria); Boeck, Helmuth [Vienna University of Technology, Atominstitut, Stadionallee 2, 1020 Wien (Austria)

    2010-12-15

    The safety report of the TRIGA Mark II reactor in Vienna includes three accident scenarios and their deterministic dose consequences to the environment. The destruction of the cladding of the most activated fuel element, the destruction of all fuel elements and a plane crash were considered scenarios in that report. The calculations were made in 1978 with the software program named STRISK. In this paper, the program package PC Cosyma was applied on the TRIGA Mark II reactor in Vienna and the deterministic consequences of the scenarios to the environment were updated. The fission product inventories of all fuel elements were calculated with ORIGEN2. To get meteorological data of the atmospheric condition around the release area, a weather station was installed. The release parameters were taken from the safety report or were replaced by worst case parameters. This paper focuses on two accident scenarios: the destruction of the cladding of the fuel element with the highest activity content and the case of a large plane crash. The current accident scenarios show good agreement with the calculations from 1978, hence no technical modifications in the safety report of the TRIGA reactor Vienna were necessary. Even in the very worst case scenario - complete destruction of all fuel elements in a large plane crash - the expected doses in the Atominstitut's neighborhood remain moderate.

  2. Treating and conditioning the radioactive wastes produced in TRIGA Reactor

    The technologies employed in treating the radioactive waste, applied at INR Pitesti are: - treating by evaporation of the liquid radioactive wastes from the TRIGA reactor and conditioning by concrete casting of the compact radioactive product. The liquid evaporation is achieved with an evaporator of 1.2 m3/h capacity supplied by PEC Engineering, France. The radioactive compact cast in concrete is finally disposed in steel barrels of 220 l capacity; - for treating and conditioning the solid wastes produced by TRIGA reactor and the Laboratory for Post-Irradiation Examination, the technology of concrete casting is used. There are two categories of solid wastes, namely, compressible, which can be compacted to a volume of upmost 5 l, and non-compressible, in which case the material is cut into pieces of 700 x 400 x 400 mm3. In the last case the compacted or broken wastes are introduced in a metallic container which is then conditioned by casting in concrete in view of final disposal in 220 l barrels; - for treating and conditioning waste ion exchangers, produced in TRIGA reactor operation, the technology of casting in bitumen in 80 l barrels which are then conditioned in 220 l barrels for final disposal

  3. Calculation of radioactive species transport in a TRIGA reactor

    Mladin, Daniela, E-mail: daniela.mladin@nuclear.ro [Institute for Nuclear Research, Campului, 1, Mioveni 115400, Arges (Romania); Mladin, Mirea, E-mail: mirea.mladin@nuclear.ro [Institute for Nuclear Research, Campului, 1, Mioveni 115400, Arges (Romania); Toma, Alexandru, E-mail: alexandru.toma@nuclear.ro [Institute for Nuclear Research, Campului, 1, Mioveni 115400, Arges (Romania); Dulama, Cristian, E-mail: cristian.dulama@nuclear.ro [Institute for Nuclear Research, Campului, 1, Mioveni 115400, Arges (Romania); Prisecaru, Ilie, E-mail: prisec@cne.pub.ro [University “Politehnica” of Bucharest, Power Engineering Faculty, Nuclear Power Plants Department, Splaiul Independentei 313, Sector 6, Bucharest (Romania); Covaci, Stefan, E-mail: bebecus@yahoo.com [Institute for Nuclear Research, Campului, 1, Mioveni 115400, Arges (Romania)

    2013-09-15

    Highlights: • We created a TRIGA facility model with CATHARE2. • We calculated a source of fission products in postulated accident conditions. • We calculated the source of Ar-41 during reactor normal operation. • We modeled the transport and release of fission products and Ar-41 at reactor stack. • Steady-state experimental Ar-41 volumetric activity is compared with the calculated activity. -- Abstract: The objective of the paper is to develop and use a model for radioactive species transport in the primary circuit and in the reactor hall of the Romanian TRIGA facility. CATHARE2 V25 code (Code for Analysis of Thermal–Hydraulics during an Accident of Reactor and Safety Evaluation) is used. CATHARE is developed by the French Atomic Energy Commission (CEA) and owned in partnership with three other French partners: EDF, AREVA-ANP and IRSN. The radio-chemical components in CATHARE2 include, besides activation products, four fission products with predefined characteristics (Kr-87, Xe-133, I-131, Cs-137). New radioactive species can be defined by the user, and the characteristics of the existing ones can be modified. The TRIGA model created comprises both the primary reactor circuit and reactor hall, involving water zones and non-condensable gas (air). Ventilation system is simulated by means of boundary conditions. Using the same facility model, two separate studies are performed with externally calculated sources: -fission product species transport and evacuation. This is done as PSA support studies, postulating core damage and volatile species release; -Ar-41 transport and evacuation. Argon activity at reactor stack is calculated for normal operation and compared to monitor readings. The paper describes also the calculation of the radioactive sources based on SCALE 4.4 in case of fission products, and using MCNP5 for Ar-41.

  4. Probabilistic safety analysis for the Triga reactor Vienna

    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)

  5. TRIGA I.N.R. Reactor liquid waste management

    During TRIGA reactor operation, primary water system cooling has been contaminated with activation products and accidentally, with fission products. The paper is a synthesis concerning the quantity and the isotopic composition of liquid waste, between 1986-1989. There are considered, in turn, the conditioned waste water as radioactive waste, and than evacuated in the river. The main isotopes generated in normal operation of INR-TRIGA reactor are Cr-51, Mn-54, Co-57, Co-58, Co-60. Cobalt isotopes activity over-stands 80% from the entire evicted activity. Computing the ratio of evicted radioactivity (A) to the consumed energy with reactor operation (E), one notes that for each operated MWh an accumulated activity in radioactive liquid waste of about 92 ± 25 μCi is generated. Attention must be paid to the interpretation of this value, because it depends on the operation power level of the reactor cooling rate, on the operation period of the ionic mixed layer filters, and the storage period of the waste in the tanks. The ratio A/E may be used for a brief characterisation of the primary cooling system corrosion and for radioactive liquid waste volume estimation which is to be evicted when the reactor operating programme is settled. Referring to the radiological impact on the environment one can state that radioactive liquid waste eviction generated in normal TRIGA reactor operation do not cause risks for population and the environment. Radioactivity measurements done for samples took from Doamnei and Arges Rivers showed that radioactivity value is lower than the detection limit of the monitor chain. For measurements it was used a CANBERRA Spectrometer, whose detection limit is 0.25x10-3 μCi/cm3 for Co-60. (authors)

  6. Fast neutron benchmark proposal at TRIGA-ACPR Reactor

    The development of fast neutron benchmarks is a historical aim of reactor physics. The dry experimental tube situated in the central region of the core in TRIGA Annular-Core Pulsing Reactor (ACPR) offers a suitable neutron source for fast neutron benchmark development. Our proposal consists in mounting a high-enriched uranium annular converter into the dry channel of the core. Preliminary computations and measurements are presented in this paper. Neutron flux computations in the dry channel and the uranium converter were performed using MCNP and WIMS codes. Also neutron flux spectrum measurements and fast and thermal neutron flux distribution measurements were performed using foil activation techniques. (authors)

  7. Operational experience data base of TRIGA Mark II reactor

    Two kinds of operational data available from operator logs: component failure-event data and abnormal event scenario information can be effectively used in PSA. Most operating data collection systems are aimed at improving the safety and availability of research reactors or commercial plants. This paper describes our failure-event data collection scheme, suitable for reliability and safety evaluations. Following the proposed data collection scheme the last five years operational experience was analysed and computerized data base for Triga Mark II reactor was developed. (orig.)

  8. Analysis of TRIGA reactor thermal power calibration method

    Analysis of thermal power method of the nuclear instrumentation of the TRIGA reactor in Ljubljana is described. Thermal power calibration was performed at different power levels and at different conditions. Different heat loss processes from the reactor pool to the surrounding are considered. It is shown that the use of proper calorimetric calibration procedure and the use of heat loss corrections improve the accuracy of the measurement. To correct the position of the control rods, perturbation factors are introduced. It is shown that the use of the perturbation factors enables power readings from nuclear instrumentation with accuracy better than without corrections.(author)

  9. The Berkeley TRIGA Mark III research reactor

    The Berkeley Research Reactor went critical on August 10, 1966, and achieved licensed operating power of 1000 kW shortly thereafter. Since then, the reactor has operated, by and large, trouble free on a one-shift basis. The major use of the reactor is in service irradiations, and many scientific programs are accommodated, both on and off campus. The principal off-campus user is the Lawrence Radiation Laboratory at Berkeley. The reactor is also an important instructional tool in the Nuclear Engineering Department reactor experiments laboratory course, and as a source of radioisotopes for two other laboratory courses given by the Department. Finally, the reactor is used in several research programs conducted within the Department, involving studies with neutron beams and in reactor kinetics

  10. Radiation sources generated by TRIGA - INR reactor operation

    The main radioisotopes occurring in TRIGA reactor and in its accessories and irradiation devices during reactor operation, that determine the radiation fields in the adjacent technological halls are presented. The source data covering, the period November 1979 to May 200, were gamma spectrometric analysis reports for the liquid radioactive waste as well as analysis reports of water, gas or refuse samples and filters for radioactive aerosols retained from installations and adjacent rooms. The main radiation sources inside the reactor building are: - fission products; - radioactive wastes; - from the reactor cooling water and water additions (intrinsic activation products); - activated products of corrosion leavings. These radiation sources are analyzed in details and their occurrence and strength interpreted as probes of reactor operation. For instance, occurrence of delayed neutrons in cooling systems indicates can failure

  11. 25 years operation experience at TRIGA reactor, Inr-Pitesti

    During the 25 years of operation of the TRIGA-2-PITESTI REACTOR no events with major impact on the nuclear safety, on personnel or on population and environment occurred, this was possible because the reactor operation was done according to, and using: a) a set of internal technical and administrative procedures, b) national regulatory body safety standards, c) quality assurance management system licensed by Lloyd's Register, d) IAEA recommendations, e) the reactor operation and the fissile material handling under safeguard of IAEA, f) a new physical protection system was built with IAEA and DOE assistance and g) personnel training and evaluation program. The main aspects of the reactor operation such as core configuration, fuel elements safety management, refueling mixed core, reactor utilization, spent fuel storage and transport are briefly reviewed. (nevyjel)

  12. Maintenance programme and periodic inspection of reactor TRIGA PUSPATI

    The PUSPATI TRIGA Reactor (RTP) at Malaysian Nuclear Agency has been safely operated since 1983. The safety of RTP requires provisions in order to facilitate maintenance and appropriate functional testing and inspection. During the operation of RTP lifetime, maintenance, periodic testing and inspection are required to ensure the adequacy of safety status of the reactor and compliance with the operational limits and conditions (OLCs). To help achieve these objectives, several measures such as preventive, corrective, annual and semi-annual maintenance and condition based maintenance have been carried out to maintain the reactor systems towards successful and safe operation of RTP. This paper will report the current practice of RTP maintenance and also to share experience and problems to fulfill safety operation of reactor. In the future, a well structured and systematic maintenance program will be needed as a strategy to shorten the unscheduled shutdown of the reactor. (Author)

  13. SANS facility at the PITESTI 14MW TRIGA reactor

    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. 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

  14. The reactor noise analysis for a TRIGA Mark-II

    For the purpose of measurement of reactor kinetic parameter, rossi-α experiment in TRIGA Mark-II reactor are performed. The past neutron noise measurement which is using HARDWARE have had defects of inaccuracy. In this study, I developed SOFTWARE to betterment of these defects and using it investigated α which is reciprocal of prompt period. To collect neutron pulses, developed data acquisition system using 16 bit personal computer (IBM-AT) and developed pascal language program to analysis neutron pulses. As a result of experiment, α is 103, 5, 155.6, 172.7, 238.7, 266.5 (1/sec) at -1, -20, -40, -60, -80, (cent) respectively, and compare it with other experiment data convinced accurate, know S/B ratio must be larger then 10% and in case of thermal reactor, low power reactor such as AGN-201 is needed to neutron noise analysis. (Author)

  15. Ageing Management in the CENM Triga Mark II Research Reactor

    Physical ageing is one of the most important factors that may reduce the safety margins calculated in the design of safety system components of a research reactor. In this context, special efforts are necessary for ensuring the safety of research reactors through appropriate ageing management actions. The paper deals with the overall aspects of the ageing management system of the Moroccan TRIGA Mark II research reactor. The management system covers among others, management of structures, critical components inspections, the control command system and nuclear instrumentation verification. The paper presents also how maintenance and periodic testing are organized and managed in the reactor module. Practical examples of ageing management actions of some systems and components during recent years are presented. (author)

  16. Refurbishment and Modernisation of PUSPATI TRIGA Reactor and Lessons Learnt

    The PUSPATI TRIGA Reactor first became critical in June 1982, and has been in operation since then. Over the years, several of the reactor systems, structures and components (SSCs) experience ageing and obsolescence problems and had to be refurbished, replaced or modernised. Initially refurbishment or replacements were carried out with SSCs of equivalent quality or capability. Subsequently SSCs were replaced with higher specification to allow for future upgrading of the reactor. Features of new SSCs should include all features of SSCs to be replaced and consider human machine interface to avoid any incidents. Lessons learnt over the years have been applied to the reactor control console modernisation project. In this project the involvement of our personnel during the design, fabrication and testing stages will enable us to have the capability to solve any associated problems with minimal vendor involvement. The close cooperation between regulators of Malaysia and vendor country was also beneficial to ensure that the project meet international safety standards

  17. The optimal control of ITU TRIGA Mark II Reactor

    In this study, optimal control of ITU TRIGA Mark-II Reactor is discussed. A new controller has been designed for ITU TRIGA Mark-II Reactor. The controller consists of main and auxiliary controllers. The form is based on Pontragyn's Maximum Principle and the latter is based on PID approach. For the desired power program, a cubic function is chosen. Integral Performance Index includes the mean square of error function and the effect of selected period on the power variation. YAVCAN2 Neutronic - Thermal -Hydraulic code is used to solve the equations, namely 11 equations, dealing with neutronic - thermal - hydraulic behavior of the reactor. For the controller design, a new code, KONTCAN, is written. In the application of the code, it is seen that the controller controls the reactor power to follow the desired power program. The overshoot value alters between 100 W and 500 W depending on the selected period. There is no undershoot. The controller rapidly increases reactivity, then decreases, after that increases it until the effect of temperature feedback is compensated. Error function varies between 0-1 kW. (author)

  18. Modification of the Core Cooling System of TRIGA 2000 Reactor

    Umar, Efrizon; Fiantini, Rosalina

    2010-06-01

    To accomplish safety requirements, a set of actions has to be performed following the recommendations of the IAEA safety series 35 applied to research reactor. Such actions are considered in modernization of the old system, improving the core cooling system and safety evaluations. Due to the complexity of the process and the difficulty in putting the apparatus in the reactor core, analytical and experimental study on the determination of flow and temperature distribution in the whole coolant channel are difficult to be done. In the present work, a numerical study of flow and temperature distribution in the coolant channel of TRIGA 2000 has been carried out using CFD package. For this study, simulations were carried out on 3-D tested model. The model consists of the reactor tank, thermal and thermalizing column, reflector, rotary specimen rack, chimney, fuel element, primary pipe, diffuser, beam tube and a part of the core are constructed by 1.50 million unstructured tetrahedral cell elements. The results show that for the initial condition (116 fuel elements in the core) and for the inlet temperature of 24°C and the primary velocity of 5.6 m/s, there no boiling phenomena occur in the coolant channel. Due to this result, it is now possible to improve the core cooling system of TRIGA 2000 reactor. Meanwhile, forced flow from the diffuser system only affected the flow pattern in the outside of chimney and put on a small effect to the fluid flow's velocity in the inside of chimney.

  19. Impact of a security event at a TRIGA reactor

    Highlights: • The fission product inventories of the TRIGA reactor were evaluated by Origen-Arp. • A security event was considered to happen in the reactor. • Atmospheric dispersion is done by RASCAL, HOTSPOT and GENII to evaluate the dose. • A significant difference among codes’ results is found. • Emergency actions for the near residential area are founded. - Abstract: The aim of this work is the study of the impact of security-related events and their consequences for research reactors, with particular emphasis on the off-site effects. The study case is done for the ENEA-Casaccia TRIGA RC-1 research reactor near Rome. The RC-1 Safety Report includes three different safety-related accident scenarios, namely the insertion of a step of positive reactivity, the uncontrolled extraction of all control rods at start-up or during a power variation and the emptying of the reactor pool. None of these scenarios imply radioactive releases at all. In this work, the focus is instead the description of the worst case scenario related to a security event and its consequences. Several possible scenarios have been analysed, however, the physical protection measures deter intrusions and, as shown by the Safety Report, even sabotage actions with major effects such as, the emptying of the reactor pool have no consequences on the fuel integrity. For these reasons, the worst security-related scenario considered in this work is a large plane crash with the complete destruction of the reactor hall and the reactor core. Propagation of the source term to the environment and the effective dose calculations have been performed using RASCAL, HOTSPOT and GENII codes following the ICRP-60 standards

  20. PUSPATI Triga Reactor - First year in operation

    First year operation of RTP reactor was mostly devoted to making in house training, setting up and testing the facilities in preparation for more routine operations. Generally the operations are categorized into 4 main purposes; experiment of research, teaching and training, demonstration, and testing and maintenance. These four purposes are elaborated in detail. Additions and modifications were performed in order to improve the safety of reactor operation. (A.J.)

  1. Dose Distribution Mapping at Reactor TRIGA PUSPATI

    Reactor has been classified as one of the radiation control area in Malaysian Nuclear Agency. Currently, to monitor the radiation and contamination level of the environment around reactor, an area radiation monitoring (ARM) system has been installed in several strategic locations. Besides that, monthly dose exposure was also measured in certain areas by Health Physics Group using thermo luminescent dosimeter (TLD). Most of these devices were installed at the reactor building wall or wide from working areas. Hence, there are possibilities that radiation workers may exposed to higher radiological risk compared to the values measured in these devices. In this study, dose rate distribution was determined in reactor hall. A 1x1 meter grid was used to locate the reading spots for the dose distribution mapping. Measurement was made using portable gamma survey meter (Ludlum Model 5) at ground level and 1 meter from ground. These data development may contribute in the planning of suitable working time in reactor hall area and reduce the chances of receiving annual dose exposure exceeding the recommended limit among the radiation workers. (author)

  2. Emergency intervention plan for 14 MW TRIGA - PITESTI Research Reactor

    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. (author)

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

    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.

  4. The SANS facility at the Pitesti 14MW TRIGA reactor

    The SANS facility existing at the Pitesti 14MW TRIGA reactor is presented. The main characteristics and the preliminary evaluation of the installation performances are given. A monochromatic neutron beam with 1.5 A ≤ λ ≤ 5 A is produced by a mechanical velocity selector with helical slots. A fruitful partnership was established between INR Pitesti (Romania) and JINR Dubna (Russia). The first step in this cooperation consists in the manufacturing in Dubna of a battery of gas-filled positional detectors devoted to the SANS instrument

  5. Thermal Hydraulic Calculation Of PUSPATI TRIGA Reactor (RTP)

    Thermal hydraulic calculation for 1 MW (thermal) PUSPATI TRIGA Reactor will be carried out using COOLOD-N2. COOLOD-N2 was developed by Japan Atomic Energy Agency and this code has capability in calculating fuel temperature distribution, coolant temperature heat flux as well as departure nucleate boiling (DNB) heat flux. For the case of RTP, the parameter such as cladding and fuel meat temperature, inlet and outlet coolant temperature was calculated in order to obtain the heat flux and DNB ratio. This paper will discuss and compare the steady state thermal hydraulic calculation for RTP and some safety parameters stated in RTP Safety Analysis Report (SAR). (author)

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

    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

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

    Ramli, Nurhayati; Maskin, Mazleha; Tom, Phongsakorn Prak; Husain, Nurfazila; Farid, Mohd Fairus Abd; Ramli, Shaharum; 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.

  8. Accident scenarios of the TRIGA Mark II reactor in Vienna

    The safety report of the TRIGA Mark II reactor in Vienna includes three accident scenarios and their deterministic dose consequence to the environment. The destruction of the most activated fuel element, the destruction of all fuel elements and a plane crash were treated scenarios in that report. The calculations were made in 1978 with the computer program STRISK. In this work, the program package PC COSYMA was applied on the TRIGA Mark II reactor in Vienna and the deterministic consequences of the scenarios to the environment were updated. The fission product inventories of all fuel elements were taken from a calculation with ORIGEN2. To get meteorological data of the atmospheric condition around the release area, a weather station was installed. The release parameters were taken from the safety report or were replaced by worst case parameters. Further on, a fourth scenario for the case of a small plane crash was added. For the sake of completeness all scenarios were calculated with different atmospheric conditions. In this paper only two accident scenarios are presented, the destruction of the fuel element with the highest activity content and the case of a large plane crash, which means a totally destruction of the reactor hall. (author)

  9. Monte Carlo analysis of Musashi TRIGA mark II reactor core

    Matsumoto, Tetsuo [Atomic Energy Research Laboratory, Musashi Institute of Technology, Kawasaki, Kanagawa (Japan)

    1999-08-01

    The analysis of the TRIGA-II core at the Musashi Institute of Technology Research Reactor (Musashi reactor, 100 kW) was performed by the three-dimensional continuous-energy Monte Carlo code (MCNP4A). Effective multiplication factors (k{sub eff}) for the several fuel-loading patterns including the initial core criticality experiment, the fuel element and control rod reactivity worth as well as the neutron flux measurements were used in the validation process of the physical model and neutron cross section data from the ENDF/B-V evaluation. The calculated k{sub eff} overestimated the experimental data by about 1.0%{delta}k/k for both the initial core and the several fuel-loading arrangements. The calculated reactivity worths of control rod and fuel element agree well the measured ones within the uncertainties. The comparison of neutron flux distribution was consistent with the experimental ones which were measured by activation methods at the sample irradiation tubes. All in all, the agreement between the MCNP predictions and the experimentally determined values is good, which indicated that the Monte Carlo model is enough to simulate the Musashi TRIGA-II reactor core. (author)

  10. Spent fuel situation at the ASTRA Seibersdorf and the TRIGA Vienna research reactors

    In the past decades Austria operated three research reactors, the 10 MW ASTRA reactor at Seibersdorf, the 250 kW TRIGA reactor at the Atomic Institut Vienna and the 1 kW Argonaut reactor at the Technical University in Graz. Since the shut down on July 31st, 1999 and decommissioning of the ASTRA reactor and the shut down of the ARGONAUT reactor Graz on July 31, 2004 only the TRIGA reactor remains operational. The MTR fuel elements of the ASTRA reactor have been shipped in spring 2001 to Savannah River and the fuel plates from the ARGONAUT reactor Graz in December 2005 under the DOE fuel return programme. (author)

  11. AFRRI TRIGA Reactor water quality monitoring program

    AFRRI has started a water quality monitoring program to provide base line data for early detection of tank leaks. This program revealed problems with growth of algae and bacteria in the pool as a result of contamination with nitrogenous matter. Steps have been taken to reduce the nitrogen levels and to kill and remove algae and bacteria from the reactor pool. (author)

  12. Decommissioning of the ICI TRIGA Mark I reactor

    Parry, D.R.; England, M.R.; Ward, A. [BNFL, Sellafield (United Kingdom); Green, D. [ICI Chemical Polymers Ltd, Billingham (United Kingdom)

    2000-07-01

    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)

  13. Decommissioning of the ICI TRIGA Mark I reactor

    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)

  14. Computer codes used during upgrading activities at MINT TRIGA reactor

    Mohammad Suhaimi Kassim; Adnan Bokhari; Mohd. Idris Taib [Malaysian Institute for Nuclear Technology Research, Kajang (Malaysia)

    1999-10-01

    MINT TRIGA Reactor is a 1-MW swimming pool nuclear research reactor commissioned in 1982. In 1993, a project was initiated to upgrade the thermal power to 2 MW. The IAEA assistance was sought to assist the various activities relevant to an upgrading exercise. For neutronics calculations, the IAEA has provided expert assistance to introduce the WIMS code, TRIGAP, and EXTERMINATOR2. For thermal-hydraulics calculations, PARET and RELAP5 were introduced. Shielding codes include ANISN and MERCURE. However, in the middle of 1997, MINT has decided to change the scope of the project to safety upgrading of the MINT Reactor. This paper describes some of the activities carried out during the upgrading process. (author)

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

    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)

  16. Neutronic Analysis of the 3 MW TRIGA MARK II Research Reactor, Part II: Benchmark Analysis of TRIGA Experiments

    The three-dimensional continuous-energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the TRIGA MARK II research reactor at AERE, Savar. Thr consistency and accuracy of both the Monte Carlo simulation and neutron transport physics was established by benchmarking the TRIGA experiments. Analysis of neutron flux and reactivity experiments comprising control rod worths, critical rod height, excess reactivity and shutdown margin were used in the validation process. Calculations of fast neutron flux, and fuel and graphite element worths distribution are also presented. Good agreement between the experiments and MCNP calculations indicate that the simulation of TRIGA reactor is treated adequately. (author)

  17. PSA application for the scram system of Romanian TRIGA Reactor

    The paper is dedicated to the fault tree analysis of the scram system in TRIGA-INR Pitesti reactor. It is a brief description of the scram system which involves instrumentation, mechanical, electrical,and control devices. The failure criteria considered is fail to drop 5 of 8 control rods. Fault tree was developed using immediate cause principle. The reliability data base used is developed in INR Pitesti based on the IAEA data available. The fault tree was analyzed by an original PC code developed for Romanian PSA program. The dominant for this fault tree appeared to be the human errors. This deserves a sensitivity analysis. If we do not consider the CCF errors contribution, the system computed unavailability is: A = 1.25 · 10-7. The failure rate is 1.087 · 10-2 eV/1000 yr. The mean time between failures is 105 years. Taking in the account roads stuck common cause failure, unavailability will increase by two magnitude orders, A = 3.02 · 10-5. We considered this number still provides a reassuring mean time between failures. This value is within the limits accepted by similar scram system studies, but is higher than the value obtained in a similar way for the TRIGA reactor of University of Texas. The reason was the taking into account in our case the human error and CCF

  18. Comparing failure rates in different research reactors including Romanian TRIGA SSR Reactor

    Probabilistic Safety Assessment (PSA) is a tool for evaluating and enhancing the safety of a nuclear reactor. In general, the PSA is used to support the system design, configuration decisions and the operational safety management of the plant. Ideally, the failure data used for safety and reliability analyses should be based on site-specific data. The paper presents failure rates for some components, both stand-alone taken from Romanian TRIGA SSR Reactor Data Collection and in comparison with other research reactors, as well as statistics pertaining to the failures and failure modes of the investigated components. Most of the work in the purpose of obtaining reliability data for Romanian TRIGA as well as for different research reactors was performed during IAEA Coordinated Research Project: 'Update and Expand Reliability database for research reactors for PSA use' (2000-2004). (authors)

  19. Operation and maintenance experiences at the C.R.E. Casaccia TRIGA reactor

    The memoir explains TRIGA RC-1 plant activities from last European TRIGA Users' Conference till today. In particular, measures following reactor exercise license renewing (March 1987) are described. Finally, difficulties and measures about shielding tank's water funguses and spores contamination, are explained. (author)

  20. On Line Measurement of Reactivity Worth of TRIGA Mark-II Research Reactor Control Rods

    Nusrat Jahan; Mamunur M. Rashid; F. Ahmed; M. G. S. Islam; M. Aliuzzaman; Islam, S.M.A

    2011-01-01

    The reactivity worth measurement system for control rods of the TRIGA MARK-II research reactor of Bangladesh has been design and developed. The theory of the kinetic technique of measuring reactivity has been used by this measurement system. The system comprises of indigenous hardware and software for online acquisition of neutron flux signals from reactor console and then computes the reactivity worth accordingly. Here for the TRIGA MARK-II research reactor, the reactivity measurement system...

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

    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

  2. STAR 3D nodal kinetics and thermal-hydraulic model for the Pennsylvania State TRIGA reactor

    A detailed three-dimensional (3D) time-dependent STAR nodal kinetics model coupled to a one-dimensional (1 D) thermal-hydraulics WIGL model has been developed to describe conservatively the peak power and pulse behavior of the Penn State University (PSU) Breazeale TRIGA reactor. This paper describes how the STAR model and its cross section data input was developed and benchmarked against actual TRIGA pulse experiments. Different core configurations (i.e., different core loading patterns, and with/without the TRIGA core next to the D20 tank) were used for several TRIGA pulse tests with different reactivity insertion worths (1.5$, 2.0$ , 2.5$). This paper shows that the STAR nodal kinetics code adequately simulates TRIGA pulses when group constants are generated from physics codes (i.e., WIMS-D4) that can accurately model the TRIGA uranium-zirconium-hydride fuel. (author)

  3. Conceptual design of epithermal neutron beam for BNCT in the thermalizing column of TRIGA reactor

    The Monte Carlo feasibility study of development of the epithermal neutron beam for BNCT clinical trials in thermalising column (TC) of TRIGA reactor is presented. The investigation of the possible use of fission converter as well as the set-up of TRIGA reactor core is performed. The optimization of the irradiation facility components is carried out and the configuration with the most favorable cost/performance ratio is proposed. The results prove, that a BNCT irradiation facility with performances, comparable to existing beams throughout the world, could be installed in TC/DC of the TRIGA reactor, quite suitable for the clinical treatments of human patients.(author)

  4. Transportation of radioactive materials from TRIGA reactors - operational considerations and regulatory problems: The situation at the Oregon State University TRIGA reactor

    General information regarding transportation of radioactive materials is presented with the idea that not everyone in our audience is fully conversant with the complexities and impact of current transportation requirements. Certain considerations and problems associated with OSU's program for the transportation of radioactive materials are briefly described. The roundtable discussion entitled: 'Transportation of Radioactive Materials from TRIGA Reactors - Operational Considerations and Regulatory Problems', at the Sixth TRIGA Owner's Conference, Corvallis, Oregon, February 27 to March 1, 1978 is attached

  5. About the safety analysis of Istanbul TRIGA Mark II reactor

    The accidents potentially related to the operation of TRIGA Mark-II reactor have been analysed in Safety Analysis Report of ITU Research Reactor, with special consideration being given to site characteristics. The maximum credible accident which can take place in a swimming pool type research reactor - accidental dropping of a fuel element into of the critical reactor core - is considered. In the safety analysis of pool type reactors BORAX accident is also included. The following events are abnormal incidents that should be taken into account: 1. Cladding rupture. 2. Reactivity accident. 3. Loss of coolant accident. Fission product release during an accident is analysed. Even though the possibility is believed to be exceedingly remote, the most unfavourable assumptions are made: the rapid insertion of the total excess' reactivity in the reactor operating at a power less then 1 kW; Coincidence of the reactivity insertion and loss of coolant accident; Cladding rupture occurring at one of the highest power density fuel elements as a consequence; Emergency ventilation system failure, leading to a vanishing filter efficiency. It is shown that, even under this most unfavourable condition, the maximum radiation to which the nearby inhabitants will be subjected, is 3.8 x 10-2 mRem per 1/2 hr. Even in the hypothetical case of the coincidence of four abnormal incidents the resulting radiation dose to the population does not exceed much the magnitude of the permissible dose of the ICRP recommendations

  6. Key role of TRIGA reactors in nuclear development in Romania

    Nuclear Power successful use depends on research and development directed toward ensuring that technology in all parts of the nuclear fuel cycle which is appropriate as well as the safety systems to protect the public and the environment. The Institute for Nuclear Research is the Romanian centre for research and development in the field of nuclear power and materials testing. To achieve the main objective - development of the scientific and technological support for the Cernavoda Nuclear Power Project - the INR main activities are directed toward: Reactor Physics - development of methods and computer codes for commissioning, in-core fuel management and safety analyses; Nuclear Fuel Cycle - performance evaluation in normal and abnormal operation conditions, irradiation tests, advanced fuels, structural materials studies; Out-of-pile Testing - fuelling machine heads and other NPS equipment testing (ultrasonic, eddy current and acoustic emission non-destructive testing); Radioactive Waste Management - low and medium level wastes treatment, conditioning and storage and decontamination methods; Radiation Protection - personnel exposure and environmental monitoring and assessment. In 1973 a contract was concluded with General Atomics to build a Research and Testing Reactor in Romania. The construction of the reactor started by the end of 1974, with a large contribution of Romanian industry. The reactor construction and pre-operational tests was completed in 1979; at the end of the same year, the reactor reached its first criticality. Since that time, the 14 MW steady state reactor was the most powerful TRIGA in the world and remains so. Twelve years of reactor operation will be soon celebrated. During this period, the reactor was operated successfully in a heavy duty cycle in order to test the experimental nuclear fuel, nuclear materials as well for irradiation of thousands of samples for electronic and electrotechnical industry needed for nuclear power plant in

  7. Environmental impact assessment around TRIGA research reactor

    Population distribution, atmospheric change, X/Q, characteristics of terrestrial ecosystem around Seoul site were surveyed. Environmental radiation and radioactivities such as grossα, grossβ, Cs-137, Sr-90 and H-3 of various environmental samples were analyzed. The values of environmental radiation dose tended to increase gradually in the light of the recent five years' results of environmental radiation monitoring around the nuclear power plants from 1980 to 1984, however, the changes were not significant. In addition, continuous assessment of environmental radiation monitoring on the roofs of main building and life science building at KAERI showed that the environmental radiation dose tended to increase a little during the night time. Judging from the above results, it is concluded that environmental contamination level by radioactive materials could be ignored in the case of radioisotope production or experiment using radioisotopes except the release of gaseous radioactive materials such as Ar-41 of short half life by the operation of nuclear reactor. (Author)

  8. Oregon State TRIGA Reactor (OSTR) console upgrading

    It was decided in the summer of 1977 to replace and upgrade part of the electronics of the OSTR console. The console was the original system installed in 1967 when the reactor first went critical. Although it was generally quite reliable, maintenance was becoming more frequent, and locating spare and replacement parts was getting very difficult. The upgrading would replace the majority of the system with new, state- of-the-art electronics. The new, upgrading package consisted of replacing the left-hand console electronics drawer; specifically: 1. The present multirange linear channel using an ion chamber was replaced by a new 9.5-decade linear channel driven by a fission chamber. No scram features are on the new linear channel. 2. The present multirange log channel with a period circuit using an ion chamber was replaced by a new 10-decade wide- range log channel, also with a period circuit, driven by a fission chamber. The same fission chamber drives the new linear and wide-range log channels. 3. The present count-rate (startup) channel using a fission chamber was removed. Its function was taken over by the new wide-range log and linear channels. 4. A new safety channel with scram capability, driven by an ion chamber, was added. 5. A new fuel element temperature circuit with scram capability was added. The upgrading package was ordered in February 1978 and installation was completed in January 1980. One of the biggest time delays in the process was the NRC review time of the Technical Specifications amendment that was requested for this change. The actual installation of the new package required five weeks, including functional testing. The linearity of the new instrument systems is excellent, and the wide-range capability of the new log and linear channels provides increased operational flexibility and accuracy, especially when a low power run immediately follows a high power run. (author)

  9. Study on neutronic safety parameters of BAEC TRIGA research reactor

    Highlights: • The control rod worth is measured using the positive period method. • Core excess reactivities are measured at different power levels. • Shut down margin of the reactor has been determined. • Fuel and water temperatures are measured at different power levels. • All the safety parameters are compared with the safety analysis report of the reactor. - Abstract: Measurement and validation of safety parameters of a nuclear reactor are required for reactor start up, normal power operation, experimental research and shutdown. The reactivity of the control rod is one of the important parameters for management of reactor operations, and is used for the prediction of control rod position at startup and the estimation of the core excess reactivity during the reactor operation. In this study, some reactor safety parameters such as control rod worth, core excess reactivity, shutdown margin, reactivity changes by fuel and control rods, and temperature effect on reactivity has been measured using digital instrumentation & control (I&C) system of the BAEC TRIGA Research Reactor (BTRR). All of these safety parameters have significant effects on the reactor control system. The measured total worth of all control rods of BTRR are 14.888 $, 14.672 $, 14.348 $ for 1.5 folding time, doubling time, 5 folding time, respectively. The measured reactivity has also been compared with the previously measured reactivity. The core excess reactivity and shutdown margin were found to be 6.38 $ and 5.20 $, respectively. The measured values were found to be within the safety limit as mentioned in the Safety Analysis Report (SAR) of the BTRR

  10. Operating history and experience at the Penn State TRIGA reactor

    Reactor operation began at Penn State with the startup of the first University Park reactor. The first core was assembled from typical MTR fuel elements, and logged 2200 MW hours during its life. In the early 1960's, a concern for the integrity of the fuel assemblies, and a desire for increased capability developed. As a result, funding was obtained for a new core, and after a study of various alternatives the TRIGA reactor was selected. The new core and control system was installed in December of 1965 and on December 31, the 'new' Penn State reactor was loaded to criticality. The present low excess reactivity is not an operational problem because the reactor is operated on a two shift basis five days a week and is often operated at low power levels for various experiments, laboratory classes and training. Problems due to activation were experienced in December of 1968, when a detailed inspection of the core structure was accomplished. The objective of this was to reveal any deterioration of the structure. It was found that such an inspection was quite feasible

  11. Upgraded reactor systems for enhanced safety at TRIGA-INR

    After almost three decades of operation of stationary TRIGA 14MW with systems provided and installed at reactor first start-up, it appeared obvious that an extended modernization program is required, both for enhancing the nuclear safety and to expand the facility lifetime. A first step has been achieved through complete HEU to LEU core conversion, meaning also core refuelling possibility for the future. Systems that have been subjected to the upgrading program are: control rods, radiation monitoring, data acquisition and processing, ventilation, irradiation devices, and above all, the outstanding modernization of the I and C system, including a brand new reactor control desk. Taking into account own and research reactors community operation experience, IAEA guides and recommendations, the basic requirement for the Instrumentation and Control System is the separation between safety and operation components, in order to decrease human error consequences and avoid common cause failures. Modernization did not cover any sensor replacement, but preserve the present scram logic and conditions (as given and approved in the Safety Report and Licensed Limits and Conditions) The entire modernization program is performed according to QA system. Out of intrinsic nuclear safety enhancement, enhanced population and environment protection is a concern and an expected result of the program. Upgrading the overall performances of the reactor and extending its operational lifetime, the Reactor Department of Institute will be able to perform competitive irradiation tests for nuclear fuel and materials, and to continue to develop nuclear investigation techniques or isotope production. (author)

  12. Perturbation analysis of the TRIGA Mark II reactor Vienna

    The safety design of a nuclear reactor needs to maintain the steady state operation at desired power level. The safe and reliable reactor operation demands the complete knowledge of the core multiplication and its changes during the reactor operation. Therefore it is frequently of interest to compute the changes in core multiplication caused by small disturbances in the field of reactor physics. These disturbances can be created either by geometry or composition changes of the core. Fortunately if these changes (or perturbations) are very small, one does not have to repeat the reactivity calculations. This article focuses the study of small perturbations created in the Central Irradiation Channel (CIC) of the TRIGA mark II core to investigate their reactivity influences on the core reactivity. For this purpose, 3 different kinds of perturbations are created by inserting 3 different samples in the CIC. The cylindrical void (air), heavy water (D2O) and Cadmium (Cd) samples are inserted into the CIC separately to determine their neutronics behavior along the length of the core. The Monte Carlo N-Particle radiation transport code (MCNP) is applied to simulate these perturbations in the CIC. The MCNP theoretical predictions are verified by the experiments performed on the current reactor core. The behavior of void in the whole core and its dependence on position and water fraction is also presented in this article. (orig.)

  13. Perturbation analysis of the TRIGA Mark II reactor Vienna

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

    2013-04-15

    The safety design of a nuclear reactor needs to maintain the steady state operation at desired power level. The safe and reliable reactor operation demands the complete knowledge of the core multiplication and its changes during the reactor operation. Therefore it is frequently of interest to compute the changes in core multiplication caused by small disturbances in the field of reactor physics. These disturbances can be created either by geometry or composition changes of the core. Fortunately if these changes (or perturbations) are very small, one does not have to repeat the reactivity calculations. This article focuses the study of small perturbations created in the Central Irradiation Channel (CIC) of the TRIGA mark II core to investigate their reactivity influences on the core reactivity. For this purpose, 3 different kinds of perturbations are created by inserting 3 different samples in the CIC. The cylindrical void (air), heavy water (D2O) and Cadmium (Cd) samples are inserted into the CIC separately to determine their neutronics behavior along the length of the core. The Monte Carlo N-Particle radiation transport code (MCNP) is applied to simulate these perturbations in the CIC. The MCNP theoretical predictions are verified by the experiments performed on the current reactor core. The behavior of void in the whole core and its dependence on position and water fraction is also presented in this article. (orig.)

  14. Dual-core TRIGA research and materials testing reactor

    General Atomic Company is under contract from the Romanian Institute for Nuclear Technologies to design, fabricate, and install a research reactor in support of the Romanian National Program for Power Reactor Development. The goal was to select a design concept that provided reasonably high neutron fluxes for long term testing of various fuel-cladding-coolant combinations and also provide high performance pulsing for transient testing of fuel specimens. An effective solution was achieved by the selection of a 14 MW steady-state TRIGA reactor for high flux endurance testing, and an Annular Core Pulsing Reactor (ACPR) for high performance pulsing testing, with both reactors mounted in the same reactor tank and operated independently. The fuel bundles for the steady-state reactor consist of 25 uranium-zirconium hydride rods clad in stainless steel arranged in a square 5 x 5 array. The steady-state core is provided with downflow cooling at a rate of approximately 275 gpm/bundle. Bundle flow tests will be performed with both heated and unheated models. The core will be optimized for peak thermal neutron flux and reactivity lifetime within the constraint of a peak fuel meat temperature of 7500C. The operation of the steady-state reactor at a power level of 14 MW will yield peak unperturbed thermal neutron fluxes in the central experiment position of 2.9 x 1014 n/cm2-sec. The corresponding fast neutron flux (less than 1.125 keV) will be 2.6 x 1014 nv. (U.S.)

  15. Fuel burnup analysis for the Moroccan TRIGA research reactor

    Highlights: ► A fuel burnup analysis of the 2 MW TRIGA MARK II Moroccan research reactor was established. ► Burnup calculations were done by means of the in-house developed burnup code BUCAL1. ► BUCAL1 uses the MCNP tallies directly in the calculation of the isotopic inventories. ► The reactor life time was found to be 3360 MW h considering full power operating conditions. ► Power factors and fluxes of the in-core irradiation positions are strongly affected by burnup. -- Abstract: The fundamental advantage and main reason to use Monte Carlo methods for burnup calculations is the possibility to generate extremely accurate burnup dependent one group cross-sections and neutron fluxes for arbitrary core and fuel geometries. Yet, a set of values determined for a material at a given position and time remains accurate only in a local region, in which neutron spectrum and flux vary weakly — and only for a limited period of time, during which changes of the local isotopic composition are minor. This paper presents the approach of fuel burnup evaluation used at the Moroccan TRIGA MARK II research reactor. The approach is essentially based upon the utilization of BUCAL1, an in-house developed burnup code. BUCAL1 is a FORTRAN computer code designed to aid in analysis, prediction, and optimization of fuel burnup performance in nuclear reactors. The code was developed to incorporate the neutron absorption reaction tally information generated directly by MCNP5 code in the calculation of fissioned or neutron-transmuted isotopes for multi-fueled regions. The fuel cycle length and changes in several core parameters such as: core excess reactivity, control rods position, fluxes at the irradiation positions, axial and radial power factors and other parameters are estimated. Besides, this study gives valuable insight into the behavior of the reactor and will ensure better utilization and operation of the reactor during its life-time and it will allow the establishment of

  16. Power spectra of stochastic signals in reactor TRIGA

    On TRIGA Mark II reactor measurements and analyses of some stochastic signals were performed to determine their reference spectra in the frequency band from 0.01 Hz to 100 Hz. Autopower spectra of neutron flux fluctuations were computed for full power and for 50 KW and 5 KW at different cooling conditions. The spectra show a significant resonance at the frequency of 2.3 Hz which dependence on the state of the cooling system. To determine the cause of the resonance vibrations of coolant water inlet pipe, ionization chamber and control rod were also investigated. Reference power spectra of these vibrations were found and only a slight correlation between the ionization chamber and control rod vibrations and the resonance were established. Since control rod vibration are most probable cause of the resonance preliminary measurements of control rod vibrations should be improved to prove this hypothesis

  17. Testing Of Secondary Cooling Component Of TRIGA Mark Reactor

    The aim of this activity is to improve the knowledge of the mechanical testing technology of the research reactor cooling pipe material. The way which was chosen is through a series of testing to know the mechanical properties of carbon steel pipe used in TRIGA-MARK II secondary cooling pipe. Scopes of these testing activities are tensile testing, hardness testing, chemical composition analysis, and metallography analysis. Visual examination shows that thickness of the pipe was reduced over the range 0.31-1.76 mm and there was scales inside the pipe about 7.1-9.1 mm. Result of the mechanical testing shows that ultimate tensile strength, yield strength, elongation and. hardness of that material are 39 kg mm2, 34 kg/mm2, 38 %, and HV161, respectively. That yield strength value is on the design range

  18. Activation of TRIGA Mark II research reactor concrete shield

    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 60Co and 152Eu and in barytes concrete samples 60Co, 152Eu and 133Ba. Measured activity density of all nuclides was found to decrease almost linearly with depth in logarithmic scale. (author)

  19. Fast Sample Transportation Systems for INAA at TRIGA Reactors

    Ismail, S.S., E-mail: ismail@ati.ac.a [Atomic Institute, Vienna University of Technology (Austria)

    2011-07-01

    The facilities of short-time neutron activation analysis at the TRIGA Mark-II (250 kW) reactor of Atomic Institute-Vienna were completely reconstructed to implement the new generation of digital gamma spectrometers, to facilitate the analysis of large samples, to enhance the sensitivity and the quality of measurements, to develop modern and fast control units, to implement moveable neutron filters for thermal-/epithermal irradiation, to implement moveable counting chambers for accurate analysis at high count rates and to develop software packages for fully-automatic analysis. The quality and performance of the facilities were tested using radioactive sources and standard reference materials. The results indicate the effective and dynamic operation of the new irradiation-counting facilities. (author)

  20. Planned Scientific programs around the Triga Mark 2 Reactor

    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

  1. TRIGA reactor relocation at the University of Texas at Austin

    The University of Texas at Austin (UT) is in the process of relocating its TRIGA reactor facilities. This undertaking includes the construction of a new reactor building with laboratories and offices and the decommissioning of the existing facility. The main campus of The University of Texas at Austin is becoming congested, and several major research projects (mostly engineering) are moving to the Balcones Research Center ∼ 8 miles from the main campus. The process of constructing a new nuclear facility in today's regulatory environment can best be described as challenging. Fortunately, research reactor licensing is not as complicated as that for commercial power facilities, although many procedures are similar. Unfortunately, the university has its own challenging procedures for building construction. Considerable time has been expended coordinating the US Nuclear Regulatory Commission (NRC) and UT licensing and construction activities. The paper summarizes the major steps and dates accomplished. The new reactor facility at the Balcones Research Center will enhance the universities ability to carry on teaching and research activities. The increased power level and the Mark II arrangement will allow us to perform new and additional projects. Considerable time and effort were devoted by the Nuclear Engineering Laboratory staff to ensure that the facility would provide educational and research flexibility over the next several years

  2. Analysis of cocked fuel elements in the AFRRI TRIGA Mark-F reactor

    The Armed Forces Radiobiology Research Institute (AFRRI) TRIGA Mark-F pulsing reactor has experienced eight cocked fuel elements during the period 5 November 1974 through 17 February 1982. Although there are no adverse health and safety consequences associated with their occurrence and there is no credible potential for system damage, cocked TRIGA fuel elements do cause inconvenience to the reactor staff and a temporary delay in operations. This paper presents the history of cocked TRIGA fuel elements at AFRRI, discusses possible mechanisms for their occurrence, and outlines a plan to isolate and ultimately determine their actual cause

  3. 44 years of operation - The successful fuel history of the TRIGA Mark II reactor Vienna

    A review is given on the fuel element situation of the TRIGA Mark II reactor Vienna after 44 years of operation. 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 11.5 g per year. Presently we have 82 TRIGA fuel elements in the core, 51 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. (author)

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

    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)

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

    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

  6. Thermal Hydraulic Analysis of 3 MW TRIGA Research Reactor of Bangladesh Considering Different Cycles of Burnup

    M. H. Altaf; N.H. Badrun

    2014-01-01

    Burnup dependent steady state thermal hydraulic analysis of TRIGA Mark-II research reactor has been carried out utilizing coupled point kinetics, neutronics and thermal hydraulics code EUREKA-2/RR. From the previous calculations of neutronics parameters including percentage burnup of individual fuel elements performed so far for 700 MWD burnt core of TRIGA reactor showed that the fuel rod predicted as hottest at the beginning of cycle (fresh core) was found to remain as the hottest until 200 ...

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

    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) 192Ir production in the core of the low power research reactor. (author)

  8. Numerical simulation of non-steady state neutron kinetics of the TRIGA Mark II reactor Vienna

    Riede, Julia; Boeck, Helmuth

    2013-01-01

    This paper presents an algorithm for numerical simulations of non-steady states of the TRIGA MARK II reactor in Vienna, Austria. The primary focus of this work has been the development of an algorithm which provides time series of integral neutron flux after reactivity changes introduced by perturbations without the usage of thermal-hydraulic / neutronic numerical code systems for the TRIGA reactor in Vienna, Austria. The algorithm presented takes into account both external reactivity changes...

  9. New practical exercises at the JSI TRIGA Mark II reactor

    Since the 1990s the Jozef Stefan Institute (JSI) TRIGA reactor has been extensively used for performing training in experimental reactor physics. In 2012 we upgraded some of the existing and introduced some new exercises. The pulse mode operation exercise was upgraded by installation of new data acquisition system. The critical experiment exercise was improved by adding a new detector inside the reactor core and changing the data acquisition system. Now we monitor neutron population with two independent fission chambers on different locations. In the past the void reactivity coefficient exercise was performed by inserting Al tube into various positions in the reactor core and measuring the corresponding reactivity changes. In order to make the exercise more realistic, we installed a pneumatic system for generating air bubbles just below the core. The aim of the exercise is to measure reactivity changes versus flow rate and air bubble position. The second new exercise was measurement of water activation. In this exercise we installed special system which pumps the water through the core at a constant flow rate to the reactor platform, where the water activity is measured. The purpose of the exercise is to measure the 16N and 19O gamma line intensity and dose rate versus reactor power. The third new exercise, named in core flux mapping, was performed by measuring the axial fission rate distribution at various radial positions in the core. We used CEA - developed mini fission chambers and a special home developed system for moving the fission chamber in axial direction and measuring the count rate versus fission chamber position. In the paper the experiments are presented together with results. (author)

  10. BNCT-Project at the Finnish TRIGA Reactor

    An epithermal neutron irradiation station for the Boron Neutron Capture Therapy (BNCT) will be constructed in the thermal column of the Finnish Triga reactor. The first target of the BNCT at FiR 1 is the treatment of malignant brain tumors. The epithermal neutrons have the capability to penetrate deep into the brain tissue thermalizing at the same time. The thermal neutrons are captured by 10B-nuclei situated ideally in the tumor cells only and thus the reaction products destroy selectively only the tumor cells. The graphite filling of the thermal column will be replaced by a special moderator material: Al+AlF3. The moderator material and its thickness has been chosen so that the system produces as much as possible epithermal neutrons with low fast neutron and gamma contamination. Both fast neutrons and gamma radiation are harmful for the patient. To reduce the gamma radiation there is a lead-bismuth gamma shield at the outer end of the moderator block. In spite of the low power (250 kW) of the reactor the needed epithermal neutron dose to destroy the tumor will be accumulated in a reasonable time e.g. 0.5 to 1.5 h. This is possible because of the rather short distance between the reactor core and the irradiation target. (author)

  11. IPR-R1 TRIGA research reactor decommissioning plan

    The International Atomic Energy Agency (IAEA) is concerning to establish or adopt standards of safety for the protection of health, life and property in the development and application of nuclear energy for peaceful purposes. In this way the IAEA recommends that decommissioning planning should be part of all radioactive installation licensing process. There are over 200 research reactors that have either not operated for a considerable period of time and may never return to operation or, are close to permanent shutdown. Many countries do not have a decommissioning policy, and like Brazil not all installations have their decommissioning plan as part of the licensing documentation. Brazil is signatory of Joint Convention on the safety of spent fuel management and on the safety of radioactive waste management, but until now there is no decommissioning policy, and specifically for research reactor there is no decommissioning guidelines in the standards. The Nuclear Technology Development Centre (CDTN/CNEN) has a TRIGA Mark I Research Reactor IPR-R1 in operation for 47 years with 3.6% average fuel burn-up. The original power was 100 k W and it is being licensed for 250 k W, and it needs the decommissioning plan as part of the licensing requirements. In the paper it is presented the basis of decommissioning plan, an overview and the end state / final goal of decommissioning activities for the IPR-R1, and the Brazilian ongoing activities about this subject. (author)

  12. New burnup calculation of TRIGA IPR-R1 reactor

    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)

  13. IPR-R1 TRIGA research reactor decommissioning plan

    The International Atomic Energy Agency (IAEA) is concerning to establish or adopt standards of safety for the protection of health, life and property in the development and application of nuclear energy for peaceful purposes. In this way the IAEA recommends that decommissioning planning should be part of all radioactive installation licensing process. There are over 200 research reactors that have either not operated for a considerable period of time and may never return to operation or, are close to permanent shutdown. Many countries do not have a decommissioning policy, and like Brazil not all installations have their decommissioning plan as part of the licensing documentation. Brazil is signatory of Joint Convention on the safety of Spent Fuel Management and on the Safety of Radioactive Waste Management, but until now there is no decommissioning policy, and specifically for research reactor there is no decommissioning guidelines in the standards. The Nuclear Technology Development Centre (CDTN/CNEN) has a TRIGA Mark I Research Reactor IPR-R1 in operation for 47 years with 3.6% average fuel burn-up. The original power was 100 kW and it is being licensed for 250 kW, and it needs the decommissioning plan as part of the licensing requirements. In the paper it is presented the basis of decommissioning plan, an overview and the end state / final goal of decommissioning activities for the IPR-R1, and the Brazilian ongoing activities about this subject. (author)

  14. New burnup calculation of TRIGA IPR-R1 reactor

    Meireles, Sincler P. de; Campolina, Daniel de A.M.; Santos, Andre A. Campagnole dos; Menezes, Maria A.B.C.; Mesquita, Amir Z., E-mail: sinclercdtn@hotmail.com.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    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)

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

    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 99mTc, which is the most widely used radionuclide in diagnostic nuclear medicine. The methods were developed for the production of the following isotopes: 18F, 85mKr, 24Na, 82Br, 64Zn, 125I. 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

  16. Analysis of JSI TRIGA MARK II reactor physical parameters calculated with TRIPOLI and MCNP

    New computational model of the JSI TRIGA Mark II research reactor was built for TRIPOLI computer code and compared with existing MCNP code model. The same modelling assumptions were used in order to check the differences of the mathematical models of both Monte Carlo codes. Differences between the TRIPOLI and MCNP predictions of keff were up to 100 pcm. Further validation was performed with analyses of the normalized reaction rates and computations of kinetic parameters for various core configurations. - Highlights: • TRIGA Benchmark keff calculated with the TRIPOLI code. • Reaction rate profiles in TRIGA calculated with TRIPOLI code. • TRIPOLI model of the JSI TRIGA was validated. • TRIGA Kinetic parameters were calculated with TRIPOLI code. • All results are in good agreement, largest discrepancies due to nuclear data

  17. INR TRIGA Research Reactors: A Neutron Source for Radioisotopes and Materials Investigation

    At the INR there are 2 high intensity neutron sources. These sources are in fact the two nuclear TRIGA reactors: TRIGA SSR 14 MW and TRIGA ACPR. TRIGA stationary reactor is provided with several in-core irradiation channels. Other several out-of-core irradiation channels are located in the vertical channels in the beryllium reflector blocks. The maximum value of the thermal neutron flux (E14 cm-2s-1 and of fast neutron flux (E>1 MeV) is 6.89×1013 cm-2s-1. For neutron activation analysis both reactors are used and k0-NAA method has been implemented. At INR Pitesti a prompt gamma ray neutron activation analysis devices has been designed, manufactured ant put into operation. For nuclear materials properties investigation neutron radiography methods was developed in INR. For these purposes two neutron radiography devices were manufacture, one of them underwater and other one dry. The neutron beams are used for investigation of materials properties and components produced or under development for applications in the energy sector (fission and fusion). At TRIGA 14 MW reactor a neutron difractormeter and a SANS devices are available for material residual stress and texture measurements. TRIGA 14 MW reactor is used for medical and industrial radioisotopes production (131I, 125I, 192Ir, etc) and a method for 99Mo-99Tc production from fission is under developing. At INR Pitesti several special programmes for new types of nuclear fuel behavior characterization are under development. (author)

  18. Neutron measurements at the TRIGA reactor Ljubljana for core inventory verification

    Safeguards inspections are periodically made in nuclear facilities as a consequence to the Nonproliferation Treaty. The inspection methods are permanently being improved and should not cause serious interference with the reactor operation. Therefore, Core Inventory Verifier (CIVR) is being developed as an indirect quantitative method for verification of the core inventory and detection of the declared operation of research reactors. The CIVR method measures the kinetic behavior of the reactor as well as the neutron flux and its energy distribution at several points inside or outside the core. Measured data taken during inspection is compared with the set of reference data determined previously. The inspection result ''nothing has changed'' indicates that all declared nuclear material really exists in the core. TRIGA reactors are one of the target groups for the CIVR method. The TRIGA reactor Ljubljana was chosen as a reference facility. First results of test series at the TRIGA reactor Ljubljana will be presented in this paper.(author)

  19. Qualitative Analysis on Void Fraction of TRIGA 2000 Reactor in Bandung

    A qualitative analysis concerning the void fraction of TRIGA 2000 reactor has been done. That analysis is performed by studying the void phenomenon theoretically, followed by studying the cooling system performance, measuring the fuel element and cooling temperature, and visually observing the operation of reactor system. TRIGA 2000 reactor is a TRIGA Mark II reactor, which originally has 1000 kW thermal power, and then is upgraded up to 2000 kW. During reactor operation, voids are observed beginning at 1000 kW power and increased at higher power. The are several probability on where the voids come from. They might be caused by boiling process, water radiolysis, pump leakage, or cavitation. From the analysis performed, the voids might be caused by nucleate boiling, which do not affect the safety of reactor operation at certain margin. (author)

  20. Determination of neutronic fluxes in research nuclear reactor of Triga Mark I and WWRS types

    In this paper is presented the determination of the thermal, epithermal and fast neutron fluxes, using neutron activation analysis technique, for two research nuclear reactors of different design: the Triga Mark I reactor was designed by Gulf General Atomic Co in USA and the WWRS reactor was designed in the URSS, both in the 50's years. (Author)

  1. Experience in the operation and maintenance of the Austrian TRIGA Mark II reactor

    The Austrian TRIGA Mark II reactor ia in operation since March 1962. The reactor instrumentation, core design and irradiation facilities and operation are described. Besides steady state power and pulse operation, square wave operation has been installed 1968, allowing power squares up to 750 kW. A Survey of reactor operation and experiments is given

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

    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 1.1013ncm2/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)

  3. Different microprocessor controlled devices for ITU TRIGA Mark II reactor

    In this paper the design of a period meter and multichannel thermometer, which are controlled by a microprocessor, in order to be used at ITU TRIGA Mark-II Reactor is presented. The system works as a simple microcomputer, which includes a CPU, a EPROM, a RAM, a CTC, a PIO, a PIA a keyboard and displays, using the assembly language. The period meter can work either with pulse signal or with analog signal depending on demand of the user. The period is calculated by software and its range is -99,9 sec, to +2.1 sec. When the period drops +3 sec, the system gives alarm illuminating a LED. The multichannel thermometer has eight temperature channels. Temperature channels can manually or automatically be selected. The channel selection time can be adjusted. The thermometer gives alarm illuminating a LED, when the temperature rises to 600 C. Temperature data is stored in the RAM and is shown on a display. This system provides us to use four spare thermocouples in the reactor. (orig.)

  4. SANS facility at the Pitesti 14MW Triga reactor

    The SANS configuration with mechanical monochromator has significantly increased luminosity while the spatial extension of the instrument is quite reasonable. Tacking account that TRIGA reactor existing at INR Pitesti is a medium flux reactor and that the available dimension for the SANS instrument is severely limited by the dimensions of the room where the instrument has to be installed, this experimental configuration has been chosen as the most suited for the situation existing in our institute. For usual collimation values of about 30 minutes, and for an inclination angle of the monochromator axis of about 2-3 degrees, Dl/l is about 20-30%, i.e. quite reasonable value. The sample width may be fixed between 10 mm and 20 mm. The minimum value of the scattering vector is Qmin = 0.005 A-0-1 while the maximal value is Qmax = 0.5 A-0-1. The relative error is ΔQ/Qmin = 0.5. In the case of our SANS instrument a monochromatic neutron beam with 1.5 A-0 ≤ λ ≤ 5 A-0 is produced by a mechanical velocity selector with helical slots. The distance between sample and detectors plane is (5.2 m). (author)

  5. Inspection of PUSPATI TRIGA Reactor (RTP) core and control rod

    The 1 MW PUSPATI TRIGA Reactor (RTP), located at Malaysian Nuclear Agency has been operated since its first criticality on 28 June 1982. The RTP uses uranium zirconium hydride fuel enriched to about 20% of U-235. The RTP has four control rods made up of boron carbide where three are fuel-followers and one is an air-follower. The aluminium cylindrical core can accommodate up to 127 fuel elements while the reflector surrounding it is made from high purity graphite. Since, the reactor power is relatively small, natural convection is used for cooling. Light water is used both as a coolant and as well as a moderator. Visual inspection of the core, fuel and control rods are carried out routinely to ascertain their integrity. An underwater camera and boroscope was used to visually inspect the top grid plate of the core as well as the control rods. No visible defect was detected at the top grid plate however, two of the fuel-follower control rods had blemishes on its surface. This paper will describe the findings of the visual inspection as well as corrective actions taken. (author)

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

    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)

  7. Current activities at the FiR 1 TRIGA reactor

    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+AlF3 (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

  8. 14 MW INR-TRIGA research reactor core conversion - emergency preparedness challenges

    INR-Pitesti TRIGA research reactor is basically a pool type reactor with a special design in order to fulfil the requirements for material testing, power reactor fuel and nuclear safety studies. The safety evaluation involved a several design basis accidents. For training purposes, and to exercise our ability to conduct Level-3 PSA studies, a severe accident scenario involving 14-MW INR-TRIGA research reactor has been developed. In this scenario is assumed that a large part of the reactor hall roof or a heavy object escaped from the crane hook is dropped over the 14-MW TRIGA-SSR core, resulting in mechanical damage of the core. It is assumed, also, that no core melting is occurring, but only fuel-cladding rupture being involved for several 25-pins fuel bundles. The paper evaluates the radiological consequences, both early and late consequences, from the emergency preparedness point of view. (author)

  9. The research reactor TRIGA Mark II of the Johannes Gutenberg-University Mainz

    Hampel, Gabriele; Eberhardt, Klaus [Mainz Univ. (Germany). Inst. of Nuclear Chemistry

    2012-10-15

    The TRIGA Mark II research reactor of the University of Mainz was built in the 1960ies on the initiative of Fritz Strassmann, co-discoverer of the fission, at that time the director of the Institute for Inorganic and Nuclear Chemistry. On August 3{sup rd}, 1965 the TRIGA Mainz reached first criticality with the insertion of the 57{sup th} fuel element in the reactor core. Two years later, in April 1967, the Nobel Prize laureate Otto Hahn initiated the first of now more than 18,000 pulses at the official inauguration. Since then, the TRIGA Mainz has operated without failure about 200 days per year. The TRIGA Mainz can be operated in the steady state mode at power levels ranging up to 100 kW{sub th}, depending on the requirements of the different experiments. Pulse-mode operation is also possible. (orig.)

  10. The research reactor TRIGA Mark II of the Johannes Gutenberg-University Mainz

    The TRIGA Mark II research reactor of the University of Mainz was built in the 1960ies on the initiative of Fritz Strassmann, co-discoverer of the fission, at that time the director of the Institute for Inorganic and Nuclear Chemistry. On August 3rd, 1965 the TRIGA Mainz reached first criticality with the insertion of the 57th fuel element in the reactor core. Two years later, in April 1967, the Nobel Prize laureate Otto Hahn initiated the first of now more than 18,000 pulses at the official inauguration. Since then, the TRIGA Mainz has operated without failure about 200 days per year. The TRIGA Mainz can be operated in the steady state mode at power levels ranging up to 100 kWth, depending on the requirements of the different experiments. Pulse-mode operation is also possible. (orig.)

  11. Safe operation of TRIGA reactor in the situation of LEU-HEU core conversion

    Romanian TRIGA reactor was commissioned in 1980. The location of the research institute is Pitesti, 100 Km west of Bucharest. In fact there are two independent cores sharing the same pool. There are a 14 MW Steady State Reactor (SSR), high flux, and materials testing reactor and an Annular Core Pulsing Reactor (ACPR). The SSR reactor is a forced convection reactor cooled via a primary circuit with 4 pumps and 3 heat exchangers. The ACPR is natural convection reactor cooled by the pool water. The characteristics of the two reactors are presented. The reactor core configuration is shown as well as the original start-up core configuration. Fuel management of TRIGA steady state core allows obtaining the requested fluxes for experimental purposes in safe operation condition. One can firmly state that the present operation of the reactor and the HEU-LEU (High Enriched Uranium - Low Enriched Uranium), core conversion fully respect the provisions of the National Regulatory Body and IAEA. (authors)

  12. United States Domestic Research Reactor Infrastructure TRIGA Reactor Fuel Support

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  13. United States Domestic Research Reactor Infrastrucutre TRIGA Reactor Fuel Support

    Douglas Morrell

    2011-03-01

    The United State Domestic Research Reactor Infrastructure Program at the Idaho National Laboratory manages and provides project management, technical, quality engineering, quality inspection and nuclear material support for the United States Department of Energy sponsored University Reactor Fuels Program. This program provides fresh, unirradiated nuclear fuel to Domestic University Research Reactor Facilities and is responsible for the return of the DOE-owned, irradiated nuclear fuel over the life of the program. This presentation will introduce the program management team, the universities supported by the program, the status of the program and focus on the return process of irradiated nuclear fuel for long term storage at DOE managed receipt facilities. It will include lessons learned from research reactor facilities that have successfully shipped spent fuel elements to DOE receipt facilities.

  14. Computer code for the thermal-hydraulic analysis of ITU TRIGA Mark-II reactor

    Istanbul Technical University (ITU) TRIGA Mark-II reactor core consists of ninety vertical cylindrical elements located in five rings. Sixty-nine of them are fuel elements. The reactor is operated and cooled with natural convection by pool water, which is also cooled and purified in external coolant circuits by forced convection. This characteristic leads to consider both the natural and forced convection heat transfer in a 'porous-medium analysis'. The safety analysis of the reactor requires a thermal-hydraulic model of the reactor to determine the thermal-hydraulic parameters in each mode of operation. In this study, a computer code cooled TRIGA-PM (TRIGA - Porous Medium) for the thermal-hydraulic analysis of ITU is considered. TRIGA Mark-II reactor code has been developed to obtain velocity, pressure and temperature distributions in the reactor pool as a function of core design parameters and pool configuration. The code is a transient, thermal-hydraulic code and requires geometric and physical modelling parameters. In the model, although the reactor is considered as only porous medium, the other part of the reactor pool is considered partly as continuum and partly as porous medium. COMMIX-1C code is used for the benchmark purpose of TRIGA-PM code. For the normal operating conditions of the reactor, estimations of TRIGA-PM are in good agreement with those of COMMIX-1C. After some more improvements, this code will be employed for the estimation of LOCA scenario, which can not be analyses by COMMIX-1C and the other multi-purpose codes, considering a break at one of the beam tubes of the reactor

  15. Thermal hydraulics modeling of the US Geological Survey TRIGA reactor

    Alkaabi, Ahmed K.

    The Geological Survey TRIGA reactor (GSTR) is a 1 MW Mark I TRIGA reactor located in Lakewood, Colorado. Single channel GSTR thermal hydraulics models built using RELAP5/MOD3.3, RELAP5-3D, TRACE, and COMSOL Multiphysics predict the fuel, outer clad, and coolant temperatures as a function of position in the core. The results from the RELAP5/MOD3.3, RELAP5-3D, and COMSOL models are similar. The TRACE model predicts significantly higher temperatures, potentially resulting from inappropriate convection correlations. To more accurately study the complex fluid flow patterns within the core, this research develops detailed RELAP5/MOD3.3 and COMSOL multichannel models of the GSTR core. The multichannel models predict lower fuel, outer clad, and coolant temperatures compared to the single channel models by up to 16.7°C, 4.8°C, and 9.6°C, respectively, as a result of the higher mass flow rates predicted by these models. The single channel models and the RELAP5/MOD3.3 multichannel model predict that the coolant temperatures in all fuel rings rise axially with core height, as the coolant in these models flows predominantly in the axial direction. The coolant temperatures predicted by the COMSOL multichannel model rise with core height in the B-, C-, and D-rings and peak and then decrease in the E-, F-, and G-rings, as the coolant tends to flow from the bottom sides of the core to the center of the core in this model. Experiments at the GSTR measured coolant temperatures in the GSTR core to validate the developed models. The axial temperature profiles measured in the GSTR show that the flow patterns predicted by the COMSOL multichannel model are consistent with the actual conditions in the core. Adjusting the RELAP5/MOD3.3 single and multichannel models by modifying the axial and cross-flow areas allow them to better predict the GSTR coolant temperatures; however, the adjusted models still fail to predict accurate axial temperature profiles in the E-, F-, and G-rings.

  16. An analysis of decommissioning costs for the AFRRI TRIGA reactor facility

    A decommissioning cost analysis for the AFRRI TRIGA Reactor Facility was made. AFRRI is not at this time suggesting that the AFRRI TRIGA Reactor Facility be decommissioned. This report was prepared to be in compliance with paragraph 50.33 of Title 10, Code of Federal Regulations which requires the assurance of availability of future decommissioning funding. The planned method of decommissioning is the immediate decontamination of the AFRRI TRIGA Reactor site to allow for restoration of the site to full public access - this is called DECON. The cost of DECON for the AFRRI TRIGA Reactor Facility in 1990 dollars is estimated to be $3,200,000. The anticipated ancillary costs of facility site demobilization and spent fuel shipment is an additional $600,000. Thus the total cost of terminating reactor operations at AFRRI will be about $3,800,000. The primary basis for this cost estimate is a study of the decommissioning costs of a similar reactor facility that was performed by Battelle Pacific Northwest Laboratory (PNL) as provided in USNRC publication NUREG/CR-1756. The data in this study were adapted to reflect the decommissioning requirements of the AFRRI TRIGA. (author)

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

    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 235U or 252Cf 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

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

    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)

  19. RIA and LOCA simulating tests on experimental fuel elements in TRIGA MT reactor of INR Pitesti

    Full text: One of the main objectives of Institute for Nuclear Research (INR), Pitesti R and D Program is to investigate thermal and mechanical behaviour of fuel elements, thresholds and mechanisms of cladding failure during RIA and LOCA tests. Dual core TRIGA Material Testing Reactor of INR Pitesti (TRIGA SS MTR and TRIGA ACPR) is utilized extensively for studies of fuel behaviour under normal and postulated accident condition. A total of 39 test fuel elements have been irradiated in the TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti under RIA conditions. The ACPR tests program is still in progress and new experiments are foreseen to be performed in the following period. The test fuel elements are instrumented with CrAl thermocouples for cladding surface temperature measurement and every test fuel element has a pressure sensor for the internal pressure measurement. An experimental database of fuel behaviour parameters including fission - gas release, sheath strain, power - burnup history, etc. has been obtained using in-pile measurements and PIE results of test fuel elements irradiated in the TRIGA Steady State Material Testing Reactor (TRIGA SS MTR) of INR Pitesti. More than 100 test fuel elements have been irradiated in TRIGA SS MTR in different power history conditions. LOCA simulating tests are planned to be performed in C2 LOCA tests capsule and in Loop A of TRIGA SS MTR of INR Pitesti. The LOCA tests in capsule C2 are instrumented to measure fuel, sheath and coolant temperature, internal element and coolant pressure during the entire irradiation period. In the second phase of the experiment the C2 capsule will be connected to the sweep gas system with the on-line gamma ray spectrometer included. RIA type tests are planned in C6 capsule of TRIGA ACPR on test fuel elements with pre-hydrided claddings in order to investigate the influence of the precipitated hydride on fuel element cladding failure at high burnups in RIA conditions. This paper

  20. The Application of Estimator Module for Controlling of TRIGA Mark II Reactor

    The estimator module application for control TRIGA Mark II reactor have been done. This application have purpose to help operator quickly and exactly when they control reactor reactivity. Which this module, if in the reactor will do experiment ( neutron activation, radioisotope production ect.) so the operator not need to calculate probability of reactivity changes. The result of estimator is close to measurements result (< 7 sec.), it is cause estimator can be used as equipment that can be used to help operation of TRIGA Mark II. (author)

  1. Thermal hydraulic analysis of the IPR-R1 TRIGA reactor

    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)

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

    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)

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

    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)

  4. The new 10 MW(t) multipurpose TRIGA reactor in Thailand

    The Office of Atomic Energy for Peace (OAEP) in Thailand awarded a turnkey contract to General Atomics (GA) to design, build and commission the Ongkharak Nuclear Research Center (ONRC), which is to include a 10 MW(t) multipurpose TRIGA research reactor. This paper describes the basic design features of this 10 MW(t) multipurpose TRIGA reactor, including features of the reactor fuel, core and related structures. The results of analyses performed in support of the Preliminary Safety Analysis Report (PSAR) are also described. (author)

  5. Design of a 250 kW mini-TRIGA reactor

    General Atomic Company has designed an inexpensive, compact core for a below-ground research reactor capable of up to 250 kW steady-state operation. This mini-TRIGA reactor is a simplified version of the familiar TRIGA Mark I. Created as a tool for small laboratories, the mini-TRIGA is intended for applications such as neutron activation analysis, neutron radiography, isotope production, and training. Like the Mark I, the mini-TRIGA is a below ground, graphite reflected, natural convection cooled reactor utilizing uranium-zirconium-hydride fuel. The principal difference between the mini-TRIGA and the Mark I is the U-235 enrichment in the fuel. In order to reduce core size, uranium enriched to 93 percent has been used instead of the standard 20 percent enrichment. Aluminum cladding is used for mini-TRIGA elements instead of stainless steel in order to reduce parasitic neutron capture and thus further reduce core size. A standard seventeen-element core results from these features which provides a thermal neutron flux of approximately 8.7 x 1012 n/cm2-sec at 250 kW in the central thimble. The standard instrumentation and control system has been greatly simplified for the mini-TRIGA. Included in the mini-TRIGA instrumentation package are a wide range log power instrument with period and safety channel and a multi-range linear power channel. The rod control chassis includes one rod position indicator with a selector switch, magnet power supply, annunciators, and scram switches. Experimental facilities available include an in-core thimble and an optional pneumatic transfer system and rotary specimen rack. (U.S.)

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

    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)

  7. Thermal spectra of the TRIGA Mark III reactor

    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)

  8. 78 FR 5840 - Notice of License Termination for University of Illinois Advanced TRIGA Reactor, License No. R-115

    2013-01-28

    ... COMMISSION Notice of License Termination for University of Illinois Advanced TRIGA Reactor, License No. R-115... No. R-115, for the University of Illinois Advanced TRIGA Reactor (ATR). The NRC has terminated the..., Facility Operating License No. R-115 is terminated. The above referenced documents may be examined,...

  9. 77 FR 7613 - Dow Chemical Company; Dow Chemical TRIGA Research Reactor; Facility Operating License No. R-108

    2012-02-13

    ... COMMISSION Dow Chemical Company; Dow Chemical TRIGA Research Reactor; Facility Operating License No. R-108... Chemical Company (the licensee) to operate the Dow Chemical TRIGA Research Reactor (DTRR) at a maximum... accordance with the NRC E-Filing rule (72 FR 49139, August 28, 2007). The E-Filing process...

  10. SANS facility at the Pitesti 14 MW Triga reactor

    Ionita, I.; Anghel, E.; Mincu, M.; Datcu, A. [Institute for Nuclear Research - Pitesti (Romania); Grabcev, B.; Todireanu, S. [National Institute of Materials Physics (NIMP) Bucharest (Romania); Constantin, F. [National Institute of Physics and Nuclear Engineering Bucharest (Romania); Shvetsov, V. [Joint Institute for Nuclear Research Dubna (Russian Federation); Popescu, G. [National College Al. Odobescu Pitesti (Romania)

    2006-07-01

    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 {<=} {lambda} {<=} 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{sub min} = 0.005 Angstrom{sup -1} while the maximal value is Q{sub max} = 0.5 Angstrom{sup -1}. The relative error is {delta}Q/Q{sub 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

  11. SANS facility at the Pitesti 14 MW Triga reactor

    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 Qmin = 0.005 Angstrom-1 while the maximal value is Qmax = 0.5 Angstrom-1. The relative error is ΔQ/Qmin = 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 consists

  12. SANS Facility at the Pitesti 14 MW TRIGA Reactor

    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 (nanocrystals, 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 Research in Pitesti, Romania becomes actual especially after the shutting down of the WWR-S Reactor from Bucharest. A monochromatic neutron beam with 1.5 A ≤ λ ≤ 5 A 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 Qmin = 0.005 A-1 while the maximal value is Qmax = 0.5 A-1. The relative error is ΔQ/Qmin = 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 consists in the manufacturing at Dubna of a battery of gas filled

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

    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

  14. Examples of the work of the Health Physics Division of the Austrian TRIGA reactor

    It will be reported about some problems of radiation protection which arise during the operation of the Austrian TRIGA reactor. Determination of noble gas concentration in the gaseous effluent. Determination of aerosol activity in the gaseous effluent. Levels of dose- equivalent rate on the shieldings of the beam holes. Cases of contamination during the reactor operation. (author)

  15. Event Investigation at Research Reactor: Case Study at PUSPATI TRIGA Reactor

    Any events or incidents at research reactors have to be investigated thoroughly to determine the direct and root causes so that corrective or preventive actions could be taken to minimise or eliminate such events or incidents from occurring in the future. These events or incidents should then be reported to the Incident Reporting System for Research Reactors (IRSRR) managed by the International Atomic Energy Agency IAEA) as lessons learnt for the research reactor community. This paper will present methodology for event investigation and several events at the PUSPATI TRIGA Reactor (RTP) that has been reported to the national regulator. However, no nuclear safety related event or incident has occurred since RTP first reached criticality in June 1982. (author)

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

    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

  17. Design of epithermal neutron beam for clinical BNCT treatment at Slovenian TRIGA research reactor

    The Monte Carlo feasibility study of development of epithermal neutron beam for BNCT clinical trials on Jozef Stefan Institute (JSI) TRIGA reactor is presented. The investigation of the possible use of fission converter for the purpose of enhancement of neutron beam, as well as the set-up of TRIGA reactor core is performed. The optimization of the irradiation facility components is carried out and the configuration with the most favorable cost/performance ratio is proposed. The simulation results prove that a BNCT irradiation facility with performances, comparable to existing beams throughout the world, could be installed in the thermalizing column of the TRIGA reactor, quite suitable for the clinical treatments of human patients. (author)

  18. Design of epithermal neutron beam for clinical BNCT treatment at Slovenian TRIGA research reactor

    Maucec, Marko [Jozef Stefan Institute, Reactor Physics Division, Lubljana (Slovenia). E-mail: marko.mauce@ijs.si

    1999-07-01

    The Monte Carlo feasibility study of development of epithermal neutron beam for BNCT clinical trials on Jozef Stefan Institute (JSI) TRIGA reactor is presented. The investigation of the possible use of fission converter for the purpose of enhancement of neutron beam, as well as the set-up of TRIGA reactor core is performed. The optimization of the irradiation facility components is carried out and the configuration with the most favorable cost/performance ratio is proposed. The simulation results prove that a BNCT irradiation facility with performances, comparable to existing beams throughout the world, could be installed in the thermalizing column of the TRIGA reactor, quite suitable for the clinical treatments of human patients. (author)

  19. Numerical simulation of non-steady state neutron kinetics of the TRIGA Mark II reactor Vienna

    Riede, Julia

    2013-01-01

    This paper presents an algorithm for numerical simulations of non-steady states of the TRIGA MARK II reactor in Vienna, Austria. The primary focus of this work has been the development of an algorithm which provides time series of integral neutron flux after reactivity changes introduced by perturbations without the usage of thermal-hydraulic / neutronic numerical code systems for the TRIGA reactor in Vienna, Austria. The algorithm presented takes into account both external reactivity changes as well as internal reactivity changes caused by feedback mechanisms like effects caused by temperature changes of the fuel and poisoning effects. The resulting time series have been compared to experimental results.

  20. Validation of the Monteburns code for criticality calculation of TRIGA reactors

    Dalle, Hugo Moura [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Jeraj, Robert [Jozef Stafan Institute, Ljubljana (Slovenia)

    2002-07-01

    Use of Monte Carlo methods in burnup calculations of nuclear fuel has become practical due to increased speed of computers. Monteburns is an automated computational tool that links the Monte Carlo code MCNP with the burnup and decay code ORIGEN2.1. This code system was used to simulate a criticality benchmark experiment with burned fuel on a TRIGA Mark II research reactor. Two core configurations were simulated and k{sub eff} values calculated. The comparison between the calculated and experimental values shows good agreement, which indicates that the MCNP/Monteburns/ORIGEN2.1 system gives reliable results for neutronic simulations of TRIGA reactors. (author)

  1. Operational safety experience at 14 MW TRIGA research reactor from INR Pitesti, Romania

    The safe operation of TRIGA-14 MW Core and Annular Pulsed TRIGA Core in the assembly of Research Reactor in Pitesti, Romania for 27 years is presented from historical perspective as well in the light of evolving safety experience. The accomplishment of safety objectives and responsibilities of operating organization is described and sustained with practical examples including management responsibilities, resources of management, performance indicators, measurement analysis and monitoring. Further improvement of safety of Research Reactor trough a large refurbishment and modernization program under way is also presented in the paper. (author)

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

    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.

  3. Neutron spectra at two beam ports of a TRIGA Mark III reactor loaded with HEU fuel

    The neutron spectra have been measured in two beam ports, one radial and another tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research in Mexico. Measurements were carried out with the reactor core loaded with high enriched uranium fuel. Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a 6LiI(Eu) scintillator and 2, 3, 5, 8, 10 and 12 in.-diameter high-density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code. For each spectrum total flux, mean energy and ambient dose equivalent were determined. Measured spectra show fission, epithermal and thermal neutrons, being harder in the radial beam port. - Highlights: • Neutron spectra of a TRIGA reactor were measured. • The reactor core is loaded with HEU. • The spectra were measured at two reactor beam ports. • Measurements were carried out at 5 and 10 W

  4. Neutronic and thermal-hydraulic experimental program in the IPR-R1 TRIGA reactor at CDTN

    The IPR-R1 TRIGA reactor, located at CDTN (Belo Horizonte/Brazil), is a typical 100 kW Mark I light-water reactor cooled by assisted natural convection with an annular graphite reflector. In order to study the safety aspects connected with the increase of the maximum steady state power of the IPR-R1 TRIGA reactor, experimental measures were taken. This paper summarizes the experimental program and some recent results and procedures of the neutronic and thermalhydraulic experiments carried out in the IPR-R1 TRIGA reactor. (authors)

  5. 30 years of reactor operation of the TRIGA reactor in Vienna

    The TRIGA reactor Vienna first went critical on March 7th 1962 at 12.04 p.m. with 57 Al-clad fuel elements. Since this time the reactor operated without major undesired shut down as students training and education reactor. Until May 1 1992 it has produced 340 MWd of power. About 90 % of the operation schedule is at maximum power level of 250 kW. Pulsing is possible up to 250 MW, until now totally 1186 pulses have been shot. Since about ten years pulse operation has diminished. Only 200 pulses have been carried out since 1980. During the last 30 years 3 reactor instrumentations have been installed, the latest one in July 1992. Two rotary specimen racks have been used during this period; presently the reactor is operated without rotary specimen rack. The experimental facilities are intensively used, all four beam tubes and the thermal column are used for solid state-, neutron- and low temperature physics experiments. Two neutron radiography facilities are installed in the thermal column and in the former experimental tank which has been converted to an irradiation channel. The core is composed of 55 old Al-clad fuel elements together with 14 SST elements and 9 FLIP elements, totally 78 fuel elements. This shows that 70 % of the core is composed of 30 year old Al-clad elements. Only 10 fuel elements have been permanently removed due to damage and are stored in a wet storage facility. Throughout the past three decades the TRIGA reactor Vienna operated without any major problems. Some components had to be replaced due to failure or ageing, but none of these problems can be considered as unacceptable for further operation. On the other hand, the operation of the TRIGA reactor Vienna together with other associated facilities like two accelerators, a helium-liquefaction plant, several x-ray facilities etc. resulted in approximately 600 engineering diploma 200 Ph.D. doctorate 3000 scientific publications in international journals or at conferences The institute's staff is

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

    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)

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

    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)

  8. Data base formation for important components of reactor TRIGA MARK II

    The paper represents specific data base formation for reactor TRIGA MARK II in Podgorica. Reactor operation data from year 1985 to 1990 were collected. Two groups of collected data were formed. The first group includes components data and the second group covers data of reactor scrams. Time related and demand related models were used for data evaluation. Parameters were estimated by classical method. Similar data bases are useful everywhere where components unavailabilities may have severe drawback. (author)

  9. Power pulse tests on CANDU type fuel elements in TRIGA reactor of INR Pitesti

    Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensor for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280cal g-1 UO2. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti. (author)

  10. Behavior of CANDU fuel under power pulse conditions at the TRIGA reactor of INR Pitesti

    Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensors for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280 cal g-1 UO2. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti. (orig.)

  11. Behavior of CANDU fuel under power pulse conditions at the TRIGA reactor of INR Pitesti

    Horhoianu, G.; Dobrea, D.; Parvan, M.; Stefan, V. [Institute for Nuclear Research, Pitesti (Romania)

    2009-04-15

    Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensors for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280 cal g{sup -1} UO{sub 2}. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti. (orig.)

  12. Power pulse tests on CANDU type fuel elements in TRIGA reactor of INR Pitesti

    Horhoianu, G.; Ionescu, D.; Olteanu, G. [Inst. for Nuclear Research, Pitesti (Romania)

    2008-07-01

    Pulse irradiation tests on short fuel elements have been carried out in TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti to investigate aspects related to the thermal and mechanical behavior of CANDU type fuel elements under short duration and large amplitude power pulse conditions. Short test fuel elements were instrumented with thermocouples for cladding surface temperature measurements and pressure sensor for element internal pressure measurement. Transient histories of reactor power, cooling water pressure, fuel element internal pressure and cladding temperature were recorded during tests. The fuel elements were subjected to total energy deposition from 70 to 280cal g{sup -1} UO{sub 2}. Rapid fuel pellet expansion due to a power excursion caused radial and longitudinal deformation of the cladding. Cladding failure mechanism and the failure threshold have been established. This paper presents some recent results obtained from these power pulse tests performed in TRIGA ACPR of INR Pitesti. (author)

  13. Power and neutron flux calculation for the PUSPATI TRIGA Reactor using MCNP

    The Malaysian 1 MW TRIGA MARK II research reactor at Malaysian Nuclear Agency 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 calculation of neutron flux and power distribution in PUSPATI TRIGA REACTOR (RTP) 14th core configuration. The 3-D continuous energy Monte Carlo code MCNP was used to develop a versatile and accurate full model of the TRIGA core and fuels. 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 as well as S (α, β) thermal neutron scattering functions distributed with the MCNP code were used. Results of calculations are analyzed and discussed. (author)

  14. Thermal Hydraulic Characteristic of TRIGA 2000 Reactor for The 110 Percent Normal Power

    In order to accomplish the Safety Analysis Report (SAR) of TRIGA 2000 reactor according to International Atomic Energy Agency (IAEA) recommendation, the thermal hydraulic aspect’s analysis of TRIGA 2000 reactor for the 110 percent normal power had been carried out by using STAT computer code. STAT code was made by General Atomic and used specifically for analysing the characteristic of TRIGA 2000 reactor. The purpose of the thermal hydraulic analysis is to considerably study the safety problems to achieve thermal hydraulic parameters in the TRIGA 2000 reactor’s core in order to convince that the reactor was not operating unless in the safety condition. Result of this analysis indicated that the film boiling does not occur in the reactor core and DNBR for 2200 kW power with the inlet temperature range between 34 °C – 40 °C is about 2.5 – 2.8 for Mc Adams correlation and about 1.6 – 1.8 for Bernath correlation. (author)

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

    Ketelaer, J; Beck, D; Blaum, K; Block, M; Eberhardt, K; Eitel, G; Ferrer, R; Geppert, C; George, S; Herfurth, F; Ketter, J; Nagy, Sz; Neidherr, D; Neugart, R; Nörtershäuser, W; Repp, J; Smorra, C; Trautmann, N; Weber, C

    2008-01-01

    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 beam line 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 235-U or 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. This pub...

  16. The Management of TRIGA Spent Fuel at ENEA RC-1 Research Reactor

    TRIGA Mark II reactor of ENEA's Casaccia research Center (in Italy named RC-1) reached first criticality in 1960. Reactor core was realized with 61 standard TRIGA fuel elements, aluminium clad. In this condition, the reactor was operated until August 1965 at a steady state power level of 100 kW. In the summer of 1965, a programme was established to increase the reactor power to 1 MW. After significant plant modifications (in order both to adapt the reactor to the new operative circumstances, including safety regulations, and to extend reactor flexibility in the widest research areas), the new criticality was reached in July 1967. The 1 MW reactor operative configuration was initially obtained with 76 standard TRIGA fuel elements, but stainless steel clad. The RC-1 Reactor is still operational and during these years, many fuel elements were used. In this paper we describe the facility, the infrastructure available for spent fuel storage, and the operative experience accumulated during these years in the management of RC-1 Spent Nuclear Fuel (SNF). The activities and the incumbencies during SNF shipment that was carried out in 1999, in the frame of the USA Return of Foreign Research Reactors Spent Fuel Programme, are also described. (author)

  17. Results of MCNP analysis for Moroccan TRIGA Mark-II Reactor

    The construction work on the Moroccan Triga Mark II research reactor has already started and the first criticality is planned for the near future. The main objective of this study is to ensure that the calculations tools available at CNESTEN as the operator of this reactor are sufficiently adequate for the prediction of the neutronic and the operating characteristics of the first Moroccan research reactor. In this work, we have analyzed the 2 MW Triga Mark II reactor using the Monte Carlo code MCNP. In order to reduce possible errors due to inexact core geometry specification, a complete and exact 3D model of this reactor was developed. The parameters of interest in this study are the core excess reactivity, the critical size of the cold and clean core, the total reactivity worth of the control rods and the verification of the shutdown margin. (author)

  18. Sipping test update device for fuel elements cladding inspections in IPR-r1 TRIGA reactor

    It is in progress at the Centro de Desenvolvimento da Tecnologia Nuclear - CDTN (Nuclear Technology Development Center), a research project that aims to investigate possible leaks in the fuel elements of the TRIGA reactor, located in this research center. This paper presents the final form of sipping test device for TRIGA reactor, and results of the first experiments setup. Mechanical support strength tests were made by knotting device on the crane, charged with water from the conventional water supply, and tests outside the reactor pool with the use of new non-irradiated fuel elements encapsulated in stainless steel, and available safe stored in this unit. It is expected that tests with graphite elements from reactor pool are done soon after and also the test experiment with the first fuel elements in service positioned in the B ring (central ring) of the reactor core in the coming months. (author)

  19. Neutronic performance of a 14 MW TRIGA reactor: LEU vs HEU fuel

    A primary objective of the US Reduced Enrichment Research and Test Reactor (RERTR) Program is to develop means for replacing, wherever possible, currently used highly-enriched uranium (HEU) fuel (235U enrichment > 90%) with low-enriched uranium (LEU) fuel (235U enrichment < 20%) without significantly degrading the performance of research and test reactors. The General Atomic Company has developed a low-enriched but high uranium content Er-U-ZrH/sub 1.6/ fuel to enable the conversion of TRIGA reactors (and others) from HEU to LEU. One possible application is to the water-moderated 14 MW TRIGA Steady State Reactor (SSR) at the Romanian Institute for Nuclear Power Reactors. The work reported here was undertaken for the purpose of comparing the neutronic performance of the SSR for HEU fuel with that for LEU fuel. In order to make these relative comparisons as valid as possible, identical methods and models were used for the neutronic calculations

  20. Experience in operation and maintenance of the TRIGA Mark II reactor at the University of Pavia

    Experience in the operation and maintenance of the 250 kW steady state/250 MW pulsed TRIGA Mark II Reactor of the University of Pavia in the past two years is reported. Data for the reactor utilization and of Health Physics activity are also presented. Since the Second European Conference of TRIGA Reactor Users in 1972, reactor operation continued normally. No major troubles occurred during this time except for rotary specimen rack rotation. Maintenance of reactor facilities, including the substitution of the rotary specimen rack with a new one manufactured on-site is described. In June 1974 measurements of fluxes in the thermal column, with most of the graphite elements removed, were carried out in order to install a neutron converter in thermal column. Some results of fluxes and cadmium ratio values are reported. A description of the converter facility set up is given. (U.S.)

  1. Natural and mixed convection in the cylindrical pool of TRIGA reactor

    Henry, R.; Tiselj, I.; Matkovič, M.

    2016-05-01

    Temperature fields within the pool of the JSI TRIGA MARK II nuclear research reactor were measured to collect data for validation of the thermal hydraulics computational model of the reactor tank. In this context temperature of the coolant was measured simultaneously at sixty different positions within the pool during steady state operation and two transients. The obtained data revealed local peculiarities of the cooling water dynamics inside the pool and were used to estimate the coolant bulk velocity above the reactor core. Mixed natural and forced convection in the pool were simulated with a Computational Fluid Dynamics code. A relatively simple CFD model based on Unsteady RANS turbulence model was found to be sufficient for accurate prediction of the temperature fields in the pool during the reactor operation. Our results show that the simple geometry of the TRIGA pool reactor makes it a suitable candidate for a simple natural circulation benchmark in cylindrical geometry.

  2. Experience with service and maintenance of a TRIGA Mark II reactor after 24 years of operation

    The maintenance work and the inspection program carried out at the TRIGA Mark II reactor Vienna after more than two decades of reactor operation is described. With the help of a special underwater telescope all surfaces inside the reactor tank were inspected visually and two beam tubes were inspected with an endoscope. A new water purification loop was installed in 1985, which was followed by a new primary coolant circuit in 1986. The reactor bridge was dismantled, all control rod drives were serviced and some components replaced. As a result of this program it was observed that a TRIGA reactor can be serviced, improved and backfitted even after 24 years of operation with minor efforts. (author)

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

    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. PMID:26141293

  4. Full conversion of materials and nuclear fuel in TRIGA SSR 14 MW research and test reactor

    During 1952-2005 General Atomics built and commissioned worldwide 62 TRIGA research reactors. Almost all reactors built by General Atomics use Low Enriched Uranium (19,9%). One exception was the TRIGA reactor from ICN Pitesti. The transition from HEU to LEU utilization, called a core conversion, is supported by Department of Energy - USA and Member States in the project 'Reduced Enrichment in Research and Testing Reactors (RERTR)' and by Member States and IAEA through Technical Cooperation programmes. The activities that are related to core conversion are managed and reviewed as refueling operations. This type of activities was performed at least six times from reactor commissioning stage until now. The implied personnel in this type of activities is licensed by CNCAN, the Romanian Regulatory Body and periodically trained. (authors)

  5. The neutron radiography facility designed for TRIGA reactors and its results

    The two TRIGA reactors of INR, the Steady State Reactor (SSR) having a power of 14 MW and Annular Core Pulsing Reactor (ACPR) having in steady state a power of 500 kW and being capable of a pulse to the peak power of 20000 MW, are placed in the same pool. The neutron flux ranging at the edges of those reactors cores is suitable for neutron radiography. The neutron radiography facility is placed in the pool of the TRIGA reactors. Till now as neutron source only the ACPR, in steady state or pulsing mode has been used. For the future one intends to use also the neutron flux of SSR. The aim of this facility is to achieve neutron radiographs of the nuclear fuel elements. (authors)

  6. Overview of probabilistic safety assessment activity for Romanian TRIGA SSR 14 MW reactor

    Current international approach of Probabilistic Safety Assessment (PSA) is not only for Nuclear Power Plants but also for Research Reactors. Recently commissioned Research Reactors use PSA in the process of licensing. In case of the Institute for Nuclear Research Pitesti, PSA for TRIGA Research Reactor was developed with a scientific aim joint to the Deterministic Analysis, covering the qualitative and quantitative approach of safety. The paper presents the PSA activity related to TRIGA SSR 14 MW reactor starting with raw data collection for obtaining a historical view of the reactor operation and to obtain reliability data used in PSA. Further on, an overall presentation of the PSA model (Initiating Events, Event Trees and Fault Trees) is made. (authors)

  7. Sipping test update device for fuel elements cladding inspections in IPR-r1 TRIGA reactor

    Rodrigues, R.R.; Mesquita, A.Z.; Andrade, E.P.D.; Gual, Maritza R., E-mail: rrr@cdtn.br, E-mail: amir@cdtn.br, E-mail: edson@cdtn.br, E-mail: maritzargual@gmail.com [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    It is in progress at the Centro de Desenvolvimento da Tecnologia Nuclear - CDTN (Nuclear Technology Development Center), a research project that aims to investigate possible leaks in the fuel elements of the TRIGA reactor, located in this research center. This paper presents the final form of sipping test device for TRIGA reactor, and results of the first experiments setup. Mechanical support strength tests were made by knotting device on the crane, charged with water from the conventional water supply, and tests outside the reactor pool with the use of new non-irradiated fuel elements encapsulated in stainless steel, and available safe stored in this unit. It is expected that tests with graphite elements from reactor pool are done soon after and also the test experiment with the first fuel elements in service positioned in the B ring (central ring) of the reactor core in the coming months. (author)

  8. Irradiation tests in TRIGA MT reactor of INR Pitesti related to safety of nuclear fuel

    Horhoianu, Grigore; Olteanu, Gheorghe [Institute for Nuclear Research, INR, PO Box 78, 1 Campului Street, RO-115400 Mioveni, Jud. Arges (Romania); Makihara, Yoshiaki [International Atomic Energy Agency Wagramerstr. 5, A-1400 Vienna (Austria)

    2006-07-01

    The design of modern power reactors reflects the close attention paid to improve safety and reliability of nuclear fuel. With the evolution of fuel design and the possibilities for more stringent operational conditions it is of concern to determine if the present safety criteria are adequate as most of them were established 15 to 20 years ago most of the time on un-irradiated materials. One of the main objectives of Institute for Nuclear Research (INR), Pitesti R and D Program is to investigate thermal and mechanical behaviour of fuel elements, thresholds and mechanisms of cladding failure during RIA and LOCA tests. Dual core TRIGA Material Testing Reactor of INR Pitesti (TRIGA SS MTR and TRIGA ACPR) is utilized extensively for studies of fuel behaviour under normal and postulated accident conditions such as reactivity - initiated accident (RIA) and loss-of-coolant accident (LOCA). A total of 40 test fuel elements have been irradiated in the TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti under RIA conditions. The ACPR tests program is still in progress and new experiments are foreseen to be performed in the following period. The test fuel elements are instrumented with CrAl thermocouples for cladding surface temperature measurement and every test fuel element has a pressure sensor for the internal pressure measurement. New RIA type tests are planned in C6 capsule of TRIGA ACPR on test fuel elements with pre-hydrided claddings in order to investigate the influence of the precipitated hydride on fuel element cladding failure at high burnups in RIA conditions. An experimental database of fuel behaviour parameters concerning fission - gas release, sheath strain, power - burnup history, etc. has been obtained using in-pile measurements and PIE results of test fuel elements irradiated in the TRIGA Steady State Material Testing Reactor (TRIGA SS MTR) of INR Pitesti. More than 110 test fuel elements have been irradiated in TRIGA SS MTR in different power

  9. Irradiation tests in TRIGA MT reactor of INR Pitesti related to safety of nuclear fuel

    The design of modern power reactors reflects the close attention paid to improve safety and reliability of nuclear fuel. With the evolution of fuel design and the possibilities for more stringent operational conditions it is of concern to determine if the present safety criteria are adequate as most of them were established 15 to 20 years ago most of the time on un-irradiated materials. One of the main objectives of Institute for Nuclear Research (INR), Pitesti R and D Program is to investigate thermal and mechanical behaviour of fuel elements, thresholds and mechanisms of cladding failure during RIA and LOCA tests. Dual core TRIGA Material Testing Reactor of INR Pitesti (TRIGA SS MTR and TRIGA ACPR) is utilized extensively for studies of fuel behaviour under normal and postulated accident conditions such as reactivity - initiated accident (RIA) and loss-of-coolant accident (LOCA). A total of 40 test fuel elements have been irradiated in the TRIGA Annular Core Pulse Reactor (TRIGA ACPR) of INR Pitesti under RIA conditions. The ACPR tests program is still in progress and new experiments are foreseen to be performed in the following period. The test fuel elements are instrumented with CrAl thermocouples for cladding surface temperature measurement and every test fuel element has a pressure sensor for the internal pressure measurement. New RIA type tests are planned in C6 capsule of TRIGA ACPR on test fuel elements with pre-hydrided claddings in order to investigate the influence of the precipitated hydride on fuel element cladding failure at high burnups in RIA conditions. An experimental database of fuel behaviour parameters concerning fission - gas release, sheath strain, power - burnup history, etc. has been obtained using in-pile measurements and PIE results of test fuel elements irradiated in the TRIGA Steady State Material Testing Reactor (TRIGA SS MTR) of INR Pitesti. More than 110 test fuel elements have been irradiated in TRIGA SS MTR in different power

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

    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)

  11. Over Twenty Years Of Experience In ITU TRIGA MARK-II Reactor

    I.T.U. TRIGA MARK-II Training and Research Reactor, rated at 250 kW steady-state and 1200 MW pulsing power is the only research and training reactor owned and operated by a university in Turkey. Reactor has been operating since March 11, 1979; therefore the reactor has been operating successfully for more than twenty years. Over the twenty years of operation: - The tangential beam tube was equipped with a neutron radiography facility, which consists of a divergent collimator and exposure room; - A computerized data acquisition system was designed and installed such that all parameters of the reactor, which are observed from the console, could be monitored both in normal and pulse operations; - An electrical power calibration system was built for the thermal power calibration of the reactor; - Publications related with I.T.U. TRIGA MARK-II Training and Research Reactor are listed in Appendix; - Two majors undesired shutdown occurred; - The I.T.U. TRIGA MARK-II Training and Research Reactor is still in operation at the moment. (authors)

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

    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)

  13. Methods used in burn-up determination of the irradiated fuel rods at TRIGA reactor

    A short presentation of the methods used at INR TRIGA reactor for the burn-up determination is given together with some considerations on ORIGEN 2 computer code used for calculating fission products activities and nuclide concentration. Burn-up is determined by gamma spectroscopy and thermal power monitoring. (Author)

  14. Planning the procedure for dismantling the TRIGA Reactor at the Medical University of Hannover

    The aim of this contribution is to provide an overview of the procedure for dismantling the TRIGA reactor at the Medical University of Hannover (MHH). In particular, the dismantling concept and techniques, the amount and the handling of radioactive material, the measurements for release, radiation protection for persons and the environment are presented. (author)

  15. Spent fuel situation at the TRIGA reactor Vienna (1.7.2000)

    The report discussed the present situation of fresh and spent fuel at the TRIGA reactor Vienna. It includes the various types of fuel elements in use, the overall inventory and a discussion of the available storage facilities. This report is updated every two years. (author)

  16. Modeling of thermal hydraulics behaviour in reactor core of reactor TRIGA PUSPATI (RTP)

    Reactor TRIGA PUSPATI (RTP) in Malaysian Nuclear Agency (Nuclear Malaysia) is the one and only research reactor in Malaysia and had been used exclusively for research and development (R and D), training for reactor operators and education purposes. The RTP is a 1 MWt pool type reactor with natural convection cooling system and pulsing capability up to 1200 MWt. It went critical on 28 June 1982 and the core configuration has been changed twelve times to date. The core is a mixed type using 20% enriched U-ZrH fuel element containing 8.5, 12 and 20wt% uranium. This paper will discuss the modeling of thermal-hydraulics behaviour in reactor core of RTP using computer code namely PARET. The results of the calculation that were carried out at RTP are modelled and temperature profiles of the thermal hydraulics data at different locations and power levels are developed. s a comparison to the thermal hydraulics calculation using PARET, an experiment were carried out at several different locations and power levels in the reactor core for temperature profile in the core to compare the result obtained from PARET. Finally, an overall analysis of the result of PARET calculation and experimental measurement were exhibited in this paper. (author)

  17. Validation of WIMS-SNAP code systems for calculations in TRIGA-MARK II type reactors

    The following paper contributes to validate the Nuclear Engineering Department methods to carry out calculations in TRIGA reactors solving a Benchmark. The benchmark is analyzed with the WIMS-D/4-SNAP/3D code system and using the cross section library WIMS-TRIGA. A brief description of the DSN method is presented used in WIMS/d4 code and also the SNAP-3d code is shortly explained. The results are presented and compared with the experimental values. In other hand the possible error sources are analyzed. (author)

  18. Experimental and analytic investigation of the ITU TRIGA Mark-II reactor core

    Experimental and analytical studies have been performed to determine the temperature distribution as a function of reactor power in the TRIGA Mark-II reactor at the Istanbul Technical University (ITU). The lumped parameter model with four governing equations was used in the analytical model. Based on the mathematical model, a computer code has been developed for calculating fuel and coolant temperatures in the reactor core. The calculated results for fuel and coolant temperature in the reactor core for different reactor power levels have been compared with the experimental data. Agreements between experiment and results from the computer code are fairly good. (orig.)

  19. Fluid Flow Characteristic Simulation of the Original TRIGA 2000 Reactor Design Using Computational Fluid Dynamics Code

    Fiantini, Rosalina; Umar, Efrizon

    2010-06-01

    Common energy crisis has modified the national energy policy which is in the beginning based on natural resources becoming based on technology, therefore the capability to understanding the basic and applied science is needed to supporting those policies. National energy policy which aims at new energy exploitation, such as nuclear energy is including many efforts to increase the safety reactor core condition and optimize the related aspects and the ability to build new research reactor with properly design. The previous analysis of the modification TRIGA 2000 Reactor design indicates that forced convection of the primary coolant system put on an effect to the flow characteristic in the reactor core, but relatively insignificant effect to the flow velocity in the reactor core. In this analysis, the lid of reactor core is closed. However the forced convection effect is still presented. This analysis shows the fluid flow velocity vector in the model area without exception. Result of this analysis indicates that in the original design of TRIGA 2000 reactor, there is still forced convection effects occur but less than in the modified TRIGA 2000 design.

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

    Currently, four research reactors with a thermal power ranging from 0.1 to 23 MWth are in operation in Germany and one new reactor (20 MWth) 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 kWth and in the pulse mode with a peak power of 250 MWth. 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.)

  1. Research programs carried out at the TRIGA Mark II reactor Vienna

    Research programs carried out at the TRIGA Mark II reactor Vienna are reported in the presentation. Many of the research programs presented at the previous TRIGA Conference in Istambul have been completed and a number of new research programs have been started some of them in cooperation or with support of the International Atomic Energy Agency. The most important project titles are: (1) Development of a laser surveillance system for spent fuel pools, (2) Identification of LWR fuel bundeles by magnetic scanning, and (3) Test of fission chambers in intense gamma fields. A damaged TRIGA fuel rod which was stored for more than 20 years has been cut in October 1983 into several pieces. The U-Zr-H samples are now being used for burn-up calibration as they contain only Cs-137. (orig.)

  2. Decontamination and decommissioning project status of the TRIGA Mark-II and III reactors in Korea

    The decontamination and decommissioning (D and D) project of the TRIGA Mark-II and Mark-III was started in January 1997, after their shutdown in 1995 due to their life and the operation of a new research reactor, HANARO, at the KAERI site in Taejon. Preparation of the decommissioning plan and environmental impact assessment, and setting up of licensing procedure and documentation for the project were performed in 1997. At the end of 1997, Hyundai Engineering Company (HEC) was selected as the main contractor to do design and licensing documentation for the D and D of both reactors. British Nuclear Fuels Plc. (BNFL) was the technical assisting partner to Heck. Licensing documents were submitted to the Ministry of Science and Technology (MOST) at the end of 1998. And the Korea Institute of Nuclear Safety (KINS) is reviewing the documents. Practical work of the D and D will start at the end of 1999 upon the government issues the license. In the meantime, July 1998, all spent fuels from the TRIGA Mark-II and III were safely transported to the US. The foremost part of the D and D work will be the TRIGA Mark-III reactor hall that will be used as a temporary storage of radioactive waste produced during the D and D work, and followed by the TRIGA Mark-II and auxiliary facilities. This paper summarizes the current status and future plans for the D and D work. (author)

  3. Calculation of fuel element temperature TRIGA 2000 reactor in sipping test tubes using 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 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)

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

    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)

  5. Conversion and evaluation of the THOR reactor core to TRIGA fuel elements

    The THOR reactor is a pool type 1 MW research reactor and has been operated since 1961. The original MTR fuel elements have been gradually replaced by TRIGA fuel elements since 1977 and the conversion completed in 1987. The calculations were performed for various core configurations by using computer codes, WIMS/CITATION. The computing results have been evaluated and compared with the core measurements after the fuel conversion. The analysis results are in good correspondence with the measurements. (author)

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

    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)

  7. Characterization of the TRIGA Mark II reactor full-power steady state

    Cammi, Antonio; Zanetti, Matteo; Chiesa, Davide; Clemenza, Massimiliano; Pozzi, Stefano; Previtali, Ezio; Sisti, Monica; Magrotti, Giovanni; Prata, Michele; Salvini, Andrea

    2015-01-01

    In this work, the characterization of the full-power steady state of the TRIGA Mark II nuclear reactor of the University of Pavia is performed by coupling Monte Carlo (MC) simulation for neutronics with "Multiphysics" model for thermal-hydraulics. Neutronic analyses have been performed starting from a MC model of the entire reactor system, based on the MCNP5 code, that was already validated in fresh fuel and zero-power configuration (in which thermal effects are negligible) using the availabl...

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

    Muhamad, Shalina Sheik; Hamzah, Mohd Arif Arif B.

    2014-02-01

    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).

  9. Analysis of tritium production in TRIGA Mark II reactor at JSI for the needs of fusion research reactors

    Jazbec, Anze; Zerovnik, Gasper; Snoj, Luka; Trkov, Andrej [Jozef Stefan Institute, Ljubljana (Slovenia)

    2013-12-15

    In future, electricity could be produced in fusion power plants. One of the steps towards development of fusion power plants is the construction of an experimental fusion reactor ITER where deuterium (D) and tritium (T) will be fused and energy will be released. As natural concentrations of T are extremely low, the T as fusion fuel will have to be produced artificially. A series of calculations were made to investigate the possibility of producing small quantities of T for experimental fusion reactors such as JET and ITER in a small research reactor like the TRIGA Mark II reactor at the Jozef Stefan Institute (JSI). The T production is the largest if all irradiation channels in reactor's reflector are filled with LiAlO{sub 2} samples. When samples are inserted, the excess reactivity decreases by around 200 pcm. In the second part of the work an estimate was made of how long the reactor can operate with current fuel supplies. Calculations were made with the TRIGLAV computer code. TRIGA can operate at full power for at least 2,860 days, during which 152 mg of T could be produced. We conclude that small TRIGA reactors can not produce any significant quantities of T for the needs of the future experimental fusion reactors. (orig.)

  10. Analysis of tritium production in TRIGA Mark II reactor at JSI for the needs of fusion research reactors

    In future, electricity could be produced in fusion power plants. One of the steps towards development of fusion power plants is the construction of an experimental fusion reactor ITER where deuterium (D) and tritium (T) will be fused and energy will be released. As natural concentrations of T are extremely low, the T as fusion fuel will have to be produced artificially. A series of calculations were made to investigate the possibility of producing small quantities of T for experimental fusion reactors such as JET and ITER in a small research reactor like the TRIGA Mark II reactor at the Jozef Stefan Institute (JSI). The T production is the largest if all irradiation channels in reactor's reflector are filled with LiAlO2 samples. When samples are inserted, the excess reactivity decreases by around 200 pcm. In the second part of the work an estimate was made of how long the reactor can operate with current fuel supplies. Calculations were made with the TRIGLAV computer code. TRIGA can operate at full power for at least 2,860 days, during which 152 mg of T could be produced. We conclude that small TRIGA reactors can not produce any significant quantities of T for the needs of the future experimental fusion reactors. (orig.)

  11. Evaluation of nuclear safety measurements in ITU TRIGA Mark-II Reactor

    For the evaluation of the radiation measurements all the records made during over 20 years of operation of ITU TRIGA Mark-II Training and Research Reactor which has 250 kW full power are considered. In addition to the routine measurements, monitoring of the radiation levels in special places in the reactor are evaluated also which can be important for special working conditions. For the evaluation of the personnel monitoring, all the records are investigated for personnel exposed to radiation working at the ITU TRIGA Mark-II Training and Research Reactor. Determinations in air and water samples are tabulated for the reactor. Water samples have been taken from two cooling systems and the cooling tower. Air samples have been taken from the filter of ventilation system. Results of all the radiation measurements are evaluated according to the maximum permissible levels from the point of view of nuclear safety and public safety. One can conclude that ITU TRIGA Mark-II Training and Research Reactor has been operated in safe conditions since the reactor criticality date on 11 March 1979. (authors)

  12. Assessment of a RELAP5 model for the IPR-R1 Triga research reactor

    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. However, several current investigations have shown that the RELAP5 code can be also applied for thermal hydraulic analysis of nuclear research systems with good predictions. In this way, as a contribution to the assessment of RELAP5/3.3 for research reactors analysis, this work presents steady-state and transient calculation results performed by a RELAP5 model to simulate the IPR-R1 TRIGA research reactor conditions operating at 50 kW. The reactor is located at the Nuclear Technology Development Centre (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 and also calculation data from the STHIRP-1 (Research Reactors Thermal Hydraulic Simulation) code were considered in the process of the model validation. The results obtained have shown that the RELAP5 model for the IPR-R1 TRIGA reproduces the actual reactor behavior in good agreement with the available data. (author)

  13. ENEA TRIGA RC-1 reactor spent fuel elements shipment to the USA

    TRIGA Mark II reactor of ENEA's Casaccia research Center (in Italy named RC-1) reached first criticality in 1960. In more than thirty years of operation, 1 MW reactor core has been modified many times for fuel elements burn-up optimization. Till now, because of achieved maximum burn-up, 146 fuel elements have been definitively removed from reactor core and transferred to the hot storages in reactor pool (5 racks around reactor vessel) and in the reactor room (pits). The activities planning, the organizing aspect study, the analysis and valuations both nuclear safety and radioprotection have been suitable for the TRIGA RC-1 fuel element shipment. Infact, no operative anomaly is appeared respect the approved procedures. Personnel engagement has been as expectations and the personnel absorbed gamma dose resulted negligible. Finally, the NAC disposable narrow time (only one week at the end of July) has not produced heavy organization problems but it has been a strong goad per all operative structures involved in the TRIGA RC-1 elements shipment

  14. Pulsed TRIGA reactor as substitute for long pulse spallation neutron source

    TRIGA reactor cores have been used to demonstrate various pulsing applications. The TRIGA reactor fuel (U-ZrHx) is very robust especially in pulsing applications. The features required to produce 50 pulses per second have been successfully demonstrated individually, including pulse tests with small diameter fuel rods. A partially optimized core has been evaluated for pulses at 50 Hz with peak pulsed power up to 100 MW and an average power up to 10 MW. Depending on the design, the full width at half power of the individual pulses can range between 2000 μsec to 3000 μsec. Until recently, the relatively long pulses (2000 μsec to 3000 μsec) from a pulsed thermal reactor or a long pulse spallation source (LPSS) have been considered unsuitable for time-of-flight measurements of neutron scattering. More recently considerable attention has been devoted to evaluating the performance of long pulse (1000 to 4000 μs) spallation sources for the same type of neutron measurements originally performed only with short pulses from spallation sources (SPSS). Adequate information is available to permit meaningful comparisons between CW, SPSS, and LPSS neutron sources. Except where extremely high resolution is required (fraction of a percent), which does require short pulses, it is demonstrated that the LPSS source with a 1000 msec or longer pulse length and a repetition rate of 50 to 60 Hz gives results comparable to those from the 60 MW ILL (CW) source. For many of these applications the shorter pulse is not necessarily a disadvantage, but it is not an advantage over the long pulse system. In one study, the conclusion is that a 5 MW 2000 μsec LPSS source improves the capability for structural biology studies of macromolecules by at least a factor of 5 over that achievable with a high flux reactor. Recent studies have identified the advantages and usefulness of long pulse neutron sources. It is evident that the multiple pulse TRIGA reactor can produce pulses comparable to those

  15. Reactor physics tests of TRIGA Mark-II Reactor in Ljubljana

    TRIGA Mark-II Reactor in Ljubljana was recently reconstructed. The reconstruction consisted mainly of replacing the grid plates, the control rod mechanisms and the control unit. The standard type control rods were replaced by the fuelled follower type, the central grid location (A ring) was adapted for fuel element insertion, the triangular cutouts were introduced in the upper plate design. However, the main novelty in reactor physics and operational features of the reactor was the installation of a pulse rod. Having no previous operational experience in pulsing, a detailed and systematic sequence of tests was defined in order to check the predicted design parameters of the reactor with measurements. The following experiments are treated in this paper: initial criticality, excess reactivity measurements, control rod worth measurement, fuel temperature distribution, fuel temperature reactivity coefficient, pulse parameters measurement (peak power, prompt energy, peak temperature). Flux distributions in steady state and pulse mode were measured as well, however, they are treated only briefly due to the volume of the results. The experiments were performed with completely fresh fuel of 12 w% enriched Standard Stainless Steel type. The core configuration was uniform (one fuel element type, including fuelled followers) and compact (no irradiation channels or gaps), as such being particularly convenient for testing the computer codes for TRIGA reactor calculations. Comparison of analytical predictions, obtained with WIMS, SLXTUS, TRIGAP and PULSTRI codes to measured values showed agreement within the error of the measurement and calculation. The paper has the following contents: 1. Introduction; 2. Steady State Experiments; 2.1. Core loading and critical experiment; 2.2. Flux range determination for tests at zero power; 2.3. Digital reactivity meter checkout; 2.4. Control rod worth measurements; 2.5. Excess reactivity measurement; 2.6. Thermal power calibration; 2

  16. Operating experience with the Cornell University TRIGA reactor

    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

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

    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

  18. Predictive maintenance and its use in TRIGA-Pitesti reactor facilities

    The Pitesti TRIGA reactor is a research and material testing reactor situated on the bottom of an open pool of 300m3, whose steady state nominal power is 14 MW. It is cooled by a primary cooling system which comprises: 4 pumps (2 in operation, 2 in standby) and 3 heat exchangers. The generated heat in the reactor core is removed by a secondary circuit with forced convection towers (provided with 6 ventilators). The reactor was used for complete CANDU fuel testing, structural material (steel, zircaloy) testing and isotope production. The TRIGA Material Test and Research reactor was commissioned at the beginning of 1980. Since that there were made extensive tests on CANDU type fuel and structural materials. It is needed the increase the reliability of equipment's and demanded an improved performance of our facilities. Good maintenance is seen as one of the main keys to improve the performance of TRIGA reactors. For a better operation we are obliged to find and use each up to date methods and strategies. Among these new techniques we could quote the probabilistic assessments, and some of predictive maintenance's techniques. Probabilistic safety and statistical analysis provided useful insights for our reactor operation. During the reactor operation there were unexpected shutdowns, reactor components failures. The data collected were statistically processed in order to obtain a reliability data base. This paper does, indifferently the cause, analysis the failures. The study emphasizes that the most reactor's scrams took place on the first year of work. The scrams number began to lower thereafter and at the end of eighties began to increase again. The greatest number of scrams were caused by the reactor electrical control and instrumentation. An important number of scrams were caused by the irradiation devices. The main conclusion of this study is that the insights are very useful to our operational procedures, to improve the maintenance strategy and the logbook

  19. Optimization of a Potential New Core of the TRIGA Mark II Reactor Vienna

    Khan, R.; Villa, M.; Bock, H.; Abele, H.; Steinhauser, G. [Vienna University of Technology-Atominstitut, Vienna (Austria)

    2011-07-01

    The TRIGA Mark II Vienna is one of the last TRIGA reactors utilizing a mixed core with High Enrich Uranium (HEU) fuel. Due to the US Fuel Return Program, the Vienna University of Technology/Atominstitut (ATI) is obliged to return its HEU fuel by 2019. There is no final decision on any further utilization of the Vienna research reactor beyond that point. However, of all possible scenarios of the future, the conversion of the current core into Low Enriched Uranium (LEU) fuel and the complete replacement of all existing 83 burned FE(s) by new fresh FE(s) are investigated herein. This paper presents detailed reactor design calculations for three different reactor cores. The core 1 employs 104-type, core 2 uses 108-type and core 3 is loaded with mixed TRIGA fuels (i.e. 104 and 108). The combination of the Monte Carlo based neutronics code MCNP5, Oak Ridge Isotope Generation and depletion code ORIGEN2 and diffusion theory based reactor physics program TRIGLAV is used for this study. On the basis of this neutronics study, the amount of fuel required for a possible future reactor operation and its cost minimization is presented in this paper. The criticality, core excess reactivity, length of initial life cycle and thermal flux density distribution is simulated for three different cores. Keeping the utilization of existing fourteen 104-type FE(s) (i.e. six burned and eight fresh FE(s)) in view, the core 3 is found the most economical, enduring and safe option for future of the TRIGA Mark II reactor in Vienna. (author)

  20. Applications of a gas-jet transport system at the research reactor TRIGA Mainz

    Research reactors of the TRIGA-type are light water cooled reactors using uranium-zirconium-hydride (UZrH) alloy fuel-moderator elements with 20% enrichment in 235U. The TRIGA Mark II-reactor at the Johannes Gutenberg-Universitat Mainz became first critical in 1965 and since then the reactor was operated failure-free during about 200 days per year. In the steady state mode the TRIGA-Mainz can be operated at power levels ranging from about 100 mWth up to 100 kWth, depending on the requirements of the different experiments. Pulse-mode operation is also possible, corresponding to a maximum pulse peak power of up to 250 MWth, a neutron flux in the order of 1015 cm-2 per pulse and a pulse width (FWHM) of about 30 ms. For irradiation experiments the TRIGA Mainz is equipped with a central experimental tube, a rotary specimen rack and three pneumatic transfer systems. In addition, four horizontal beam ports penetrate the biological shield and extend inside the pool towards the reflector surrounding the reactor core. The TRlGA-SPEC experiment currently being installed at beam port B of the TRlGA Mainz research reactor consists of two branches: (i) the Penning-trap mass spectrometer TRlGA-TRAP and (ii) the collinear laser spectroscopy setup TRlGA-LASER. At TRlGA-SPEC a gas-jet system is connected to a high-temperature ion source and a subsequent mass analyzing magnet. The nuclides of interest are then guided either to TRlGA-TRAP or to TRlGA-LASER. Currently, TRlGA-SPEC is the only facility world-wide that is installed at a nuclear research reactor

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

    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.

  2. Optimization of a Potential New Core of the TRIGA Mark II Reactor Vienna

    The TRIGA Mark II Vienna is one of the last TRIGA reactors utilizing a mixed core with High Enrich Uranium (HEU) fuel. Due to the US Fuel Return Program, the Vienna University of Technology/Atominstitut (ATI) is obliged to return its HEU fuel by 2019. There is no final decision on any further utilization of the Vienna research reactor beyond that point. However, of all possible scenarios of the future, the conversion of the current core into Low Enriched Uranium (LEU) fuel and the complete replacement of all existing 83 burned FE(s) by new fresh FE(s) are investigated herein. This paper presents detailed reactor design calculations for three different reactor cores. The core 1 employs 104-type, core 2 uses 108-type and core 3 is loaded with mixed TRIGA fuels (i.e. 104 and 108). The combination of the Monte Carlo based neutronics code MCNP5, Oak Ridge Isotope Generation and depletion code ORIGEN2 and diffusion theory based reactor physics program TRIGLAV is used for this study. On the basis of this neutronics study, the amount of fuel required for a possible future reactor operation and its cost minimization is presented in this paper. The criticality, core excess reactivity, length of initial life cycle and thermal flux density distribution is simulated for three different cores. Keeping the utilization of existing fourteen 104-type FE(s) (i.e. six burned and eight fresh FE(s)) in view, the core 3 is found the most economical, enduring and safe option for future of the TRIGA Mark II reactor in Vienna. (author)

  3. Operation, maintenance, and utilization of 250 kW TRIGA Mark II reactor at the Institute Jozef Stefan, Ljubljana (Yugoslavia))

    At the Institute 'Jozef Stefan' in Ljubljana 250 kW TRIGA Mark II Reactor has been in operation since May 31, 1966. It is the steady state operated reactor without pulsing capabilities. In the paper the operational data, maintenance and utilization of the reactor are summarized for the first four years of reactor operation. (author)

  4. Computational analysis of thermo-hydraulic behavior of TRIGA research reactor

    Highlights: ► Key thermal hydraulic parameters of the 3 MW TRIGA Mark-II research reactor were investigated under steady-state conditions. ► The thermal hydraulic codes NCTRIGA, PARET and COOLOD-N2 were employed for investigation. ► The NCTRIGA, PARET and COOLOD-N2 model calculations were benchmarked through the TRIGA experimental and operational data. ► The result obtained in this investigation can be used for upgrading the current core configuration of the TRIGA reactor. -- Abstract: Key thermal hydraulic parameters of the 3 MW TRIGA Mark-II research reactor operating under steady-state conditions were investigated using the thermal hydraulic codes NCTRIGA, PARET and COOLOD-N2. Results of the neutronic analysis performed by 3-D Monte Carlo code MCNP4C were used in NCTRIGA and coupled output of neutronic analysis carried out by using 3-D diffusion code CITATION and 3-D Monte Carlo code MCNP4B2 were used in the PARET to study the steady-state thermal hydraulic behavior of the reactor. To benchmark the NCTRIGA, PARET and COOLOD-N2 models, data were obtained from different measurements executed by thermocouples in the instrumented fuel elements (C1 and D2) and the hottest fuel element (C4) during the steady-state operation both under forced and natural convection mode and compared with the calculation found to be quite consistent. The mass flow rates needed for input to PARET and COOLOD-N2 were taken from final safety analysis report (FSAR) for a downward forced coolant flow equivalent to 3500 gpm. For natural convection cooling of reactor, mass flow rate was generated using NCTRIGA code. The testing of the NCTRIGA, PARET and COOLOD-N2 model calculations through benchmarking the available TRIGA experimental and operational data showed that NCTRIGA, PARET and COOLOD-N2 codes can successfully be used to analyze the thermal hydraulic behavior of the reactor for the steady-state operation under both natural and forced convection mode of coolant flow to predict

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

    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)

  6. Operation experiences of the Kartini reactor using Bandung Triga Mark II spent fuels

    The operating history and improvements of the Kartini research reactor are presented. The Kartini reactor is operated during office hours: 5 days a week and 6-7 hours a day, except in particular cases. For 15 years since 1979 the Kartini reactor has been operated using spent fuels and used core from the Bandung Triga Mark II. Since 1994, however the Kartini reactor has been operated using the 104 SS type of fuel elements. Several difficulties and anomalies were encountered during its operation. A brief explanation of the maintenance, quality control and quality assurance programme during its operation are also discussed. (orig.)

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

    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)

  8. Planning and implementation of Istanbul Technical University TRIGA research reactor program

    The Istanbul Technical University TRIGA Research Reactor at the Institute for Nuclear Energy, which went critical on March 11, 1979 is basically a pulsing type TRIGA Mark - II reactor. Completion of the ITU-TRR contributed to broaden the role of the Institute for Nuclear Energy of the Technical University in Istanbul in the nuclear field by providing for the first time adequate on-campus experimental facilities for nuclear engineering studies to ITU students. The research program which is currently under planning at ITU-NEE encompasses: a) Neutron activation analysis studies by techniques and applications to chemistry, mining, materials research, archaeological and biomedical studies; b) applications of Radioisotopes; c) Radiography with reactor neutron beams; d) Radiation Pulsing

  9. Evaluation Of Reactor Coolant System Of Design Of Bandung TRIGA Mark II 2 MW Reactor

    An evaluation of reactor coolant system of Bandung TRIGA Mark II has been carried out. The evaluation is conducted for primary and secondary system, both for steady state and transient conditions. The evaluation is based on the analysis results done by the operator. In the steady state (i.e. normal operation), the maximum temperature of fuel element is 569.7C. A series of analysis covering various accident scenarios of LOPA and LOCA shows that the coolant system and ECCS able to maintain the fuel temperature less then 970C, then the fuel integrity is kept safe. However, the detail analysis using validated codes is still needed to support the actual safety analysis

  10. Neutronic calculation to the TRIGA Ipr-R1 reactor using the WIMSD4 and CITATION codes

    The WIMSD4 and CITATION codes are used to calculate neutronic parameters of a TRIGA reactor. The results are compared with experimental values. Five configurations are analysed and the excess reactivity worth, the fuel temperature reactivity coefficient, the control reactivity worth, safety and regulation rod of the TRIGA IPR-R1 reactor are calculated. The idea is to obtain the systematic error for k∞ for this methodology comparing the calculated and the experimental results

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

    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.

  12. On Line Measurement of Reactivity Worth of TRIGA Mark-II Research Reactor Control Rods

    Nusrat Jahan

    2011-09-01

    Full Text Available The reactivity worth measurement system for control rods of the TRIGA MARK-II research reactor of Bangladesh has been design and developed. The theory of the kinetic technique of measuring reactivity has been used by this measurement system. The system comprises of indigenous hardware and software for online acquisition of neutron flux signals from reactor console and then computes the reactivity worth accordingly. Here for the TRIGA MARK-II research reactor, the reactivity measurement system was implemented with a dedicated circuit assembly and a conventional personal computer. A high-level Visual Basic real-time programming has been developed for data acquisition, reactivity calculation, online display (numerically as well as graphically, saving data, etc. To measure reactivity worth of TRIGA reactor control rods the rod drop experimental technique has been adopted. The results of tests experiments, carried out with the rod drop method for measuring various reactivity worth of control rods have been presented in the paper. A comparison between this results with the results using period method and that of computation method, demonstrated that the response of this reactivity measurement system is fast enough to monitor and measure the safety-related reactivity and power excursions in the reactor.

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

    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

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

    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

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

    Full-text: 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. (author)

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

    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.

  17. Benchmark analysis of the TRIGA MARK II research reactor using Monte Carlo techniques

    This study deals with the neutronic analysis of the current core configuration of a 3-MW TRIGA MARK II research reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh and validation of the results by benchmarking with the experimental, operational and available Final Safety Analysis Report (FSAR) values. The 3-D continuous-energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the TRIGA core. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Continuous energy cross-section data from ENDF/B-VI and ENDF/B-V and S(α,β) scattering functions from the ENDF/B-VI library were used. The consistency and accuracy of both the Monte Carlo simulation and neutron transport physics was established by benchmarking the TRIGA experiments. The effective multiplication factor, power distribution and peaking factors, neutron flux distribution, and reactivity experiments comprising control rod worths, critical rod height, excess reactivity and shutdown margin were used in the validation process. The MCNP predictions and the experimentally determined values are found to be in very good agreement, which indicates that the simulation of TRIGA reactor is treated adequately

  18. Use of the TRIGA Mark III as a simulator for the Tokamak Fusion Test Reactor (TFTR)

    The Exposure-Room feature on the TRIGA Mark III reactor offers the possibility for experiments which simulate the expected dose to components produced in pulses from the TFTR fusion reactor. Although the TRIGA pulse is considerably shorter and the TRIGA fast-neutron spectrum is considerably softer, the fast-neutron fluence represents a good match to that expected from TFTR, and the thermal-neutron fluence and gamma-ray dose from the TRIGA constitute a considerable overtest compared to that expected from TFTR. An experiment is underway which involves irradiating a prototype TFTR cyropump in the Exposure-Room facility. The cryopump is pulsed twice, once before and once after deuterium is admitted to the pump. The object is to determine whether the radiation has any desorptive effect on the deuterium in the pump. Care must be taken to prevent conditions under which the deuterium might explode, or under which oxygen condensed in the presence of the liquid nitrogen or liquid helium might constitute a combustion hazard. (author)

  19. Characteristics and uses of a 250 kW TRIGA reactor

    The 250 kW TRIGA Mark II reactor is a light water reactor with solid fuel elements in which the zirconium hydride moderator is homogeneously distributed between enriched uranium. Therefore the reactor has the large prompt negative temperature coefficient of reactivity, the fuel also has very high retention of radioactive fission products. The reactor core is a cylindrical configuration with an annular graphite reflector. The experimental facilities include a rotary specimen rack, a central incore radiation thimble, a pneumatic transfer system, and pulsing capability. Other experimental facilities include two radial and two tangential beam tubes, a graphite thermal column, and a graphite thermalizing column. At the steady state power of 250 kW the peak flux is 1x1013n/cm2s in the central test position. In addition, pulsing to about 2000 MW is usually provided giving peak fluxes of about 2x1016n/cm2sec. All TRIGA reactors produce a core-average thermal neutron flux of about 107n.v per watt. Only with very large accelerators could such a high neutron flux be achieved. In order to give an appreciation for the research conducted at research reactors, the types of research could be summarized as follows: thermal neutron scattering, neutron radiography, neutron and nuclear physics, activation analysis, radiochemistry, biology and medicine, and teaching and training. Typical applied research with a 250 kW reactor has been conducted in medicine in biology, archeology, metallurgy and materials science, engineering and criminology. It is well known that research reactors have been used routinely to produce isotopes for industry and medicine. In some instances, reactors are the preferred method of isotope production. We can conclude that the 250 kW TRIGA research reactor is a useful and wide ranging source of radiation for basic and applied research. The operation cost for this instrument is relatively low. (author)

  20. Vaporization Rate Analysis of Primary Cooling Water from Reactor PUSPATI TRIGA (RTP) Tank

    Primary cooling system consists of pumps, heat exchangers, probes, a nitrogen-16 diffuser and associated valves is connected to the reactor TRIGA PUSPATI (RTP) tank by aluminium pipes. Both the primary cooling system and the reactor tank is filled with demineralized light water (H2O), which serves as a coolant, moderator as well as shielding. During reactor operation, vaporization in the reactor tank will reduce the primary water and contribute to the formation of vapor in the reactor hall. The vaporization may influence the function of the water subsequently may affect the safety of the reactor operation. It is essential to know the vaporization rate of the primary water to ensure its functionality. This paper will present the vaporization rate of the primary cooling water from the reactor tank and the influence of temperature of the water in the reactor tank to the vaporization rate. (author)

  1. Conversion of the IAN-R1 reactor from MTR fuel to TRIGA LEU fuel

    The Institute of Nuclear Sciences and Alternative Energies (INEA) in Bogota, Colombia, has operated since 1965, a small 10 kW(t) research reactor, known as the IAN-R1 reactor, which was upgraded to 30 kW(t) in 1980. This reactor was provided to the Republic of Colombia under the U.S. Atoms for Peace Program, and which has been fueled with MTR HEU fuel, enriched nominally to 93% U-235. With the cooperation of the International Atomic Energy Agency (IAEA), a gradual reactor upgrade program has been undertaken beginning in 1987. The first step in this program was the upgrade of reactor instrumentation and control systems. In December, 1994, the IAEA and INEA entered into a tripartite contract with General Atomics (GA) to prepare a new safety analysis report for performing an HEU to LEU conversion of the R-1 reactor, manufacture TRIGA type LEU (19.7% enriched) fuel to replace the original MTR-HEU fuel plate assemblies, upgrade the reactor power to 100 kW(t), carry out additional upgrades of auxiliary reactor systems and commission the reactor with TRIGA fuel. (author)

  2. ENEA TRIGA RC -1 research reactor and trade project: An important contribution to the ADS road map

    TRIGA Mark II reactor of ENEA's Casaccia research Center (in Italy named RC-1) reached first criticality in 1960 and it is still running at 1 MW power level, mainly for short mean life time radioisotopes production (for medical purposes) and neutron radiography. Since 2001, plant personnel and other national/international scientist, were involved in the TRADE (TRiga Accelerator Driven Experiment) project. TRADE experiment, that consists in the coupling of an external proton accelerator to a target to be installed in the central channel of the TRIGA core scrammed to sub-criticality, was based on an original idea of Prof. Carlo Rubbia, presented at CEA in October 2000 and was aimed at a global demonstration of the ADS concept. The TRADE layout, the studies about Target, Target Cooling System, Shielding and other matters that were investigated will be described in order to evidence their impact on the Triga reactor and reactor activity. (author)

  3. 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

    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)

  4. Decontamination and decommissioning project status of the TRIGA mark-2±3 research reactors

    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 at the Korea Atomic Energy Research Institute (KAERI) in Taejeon. Decontamination and decommissioning (D and D) project of the TRIGA Mark-II and Mark-III was started in January 1997 and will be completed in December 2002. In 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). In 1998, 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 technical assisting partner of HEC. The decommissioning plan document was submitted to the Ministry of Science and Technology (MOST) for the decommissioning license in December 1998, and it expecting to be issued a license at the end of September 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

  5. Analysis of potential common cause failure events for Romania-TRIGA 14 MW reactor

    Radu, Gabriela, E-mail: gabriela.radu@nuclear.ro [Institute for Nuclear Research, Campului 1, 115400 Mioveni, Arges (Romania); Mladin, Daniela, E-mail: daniela.mladin@nuclear.ro [Institute for Nuclear Research, Campului 1, 115400 Mioveni, Arges (Romania); Prisecaru, Ilie, E-mail: prisec@cne.pub.ro [Power Engineering Department, University “Politehnica” of Bucharest, 313 Splaiul Independentei, Bucharest (Romania)

    2013-12-15

    Highlights: • Presents an overview of the general definitions and models for treating CCFs. • Includes a qualitative analysis of the CCFs events for the Romanian TRIGA reactor. • Performs a quantitative CCF analysis for the primary circulating pumps. - Abstract: Common cause failure (CCF) events can have significant impact in the availability of safety systems in nuclear power plants or research reactors. For these reasons CCF data are being collected and analyzed in many countries. The paper presents first the methods for CCF parameters estimation and mathematical models for estimation of the probability of occurrence of common cause events. Then, the paper describes data collection and statistics referring to the potential CCF for components of Romania-TRIGA SSR 14 MW reactor. Potential CCF's components for Romania-TRIGA reactor involves pumps, control rods and control rods mechanisms, valves, fans, ejectors. Every group of components susceptible to CCF is discussed taking into account the event attributes like root cause, coupling factor, detection method and corrective action taken. An example of quantitative CCF analysis using different mathematical models for main circulating pumps is also considered.

  6. Comparison of decommissioning options for the example of 2 research reactors of type TRIGA

    For decommissioning of nuclear facilities usually the two decommissioning strategies 'immediate dismantling' or 'deferred dismantling (safe enclosure)' are applied. In general, immediate dismantling is regarded as the more advantageous and more preferable option. Accordingly, immediate dismantling is the mostly selected option. Nevertheless, only in a case by case analysis it can be shown, which decommissioning option is the better one e. g. with respect to technical aspects or to a use of the facility / remaining facility. For two real decommissioning projects of two similar research reactors of TRIGA type GRS with support of the operator, German Cancer Research Center Heidelberg, performed a study on possible advantages of the two different strategies selected. While the first research reactor, TRIGA HD I, was dismantled immediately, the second research reactor, TRIGA HD II, was dismantled after a 20 years period of safe enclosure. As a result, it could be shown, that the selected different decommissioning strategies reflected the special conditions of each both research reactor in best way, so that a clear preference for one of the two decommissioning strategies can not be deduced. The slides of the presentation have been added at the end of the paper. (authors)

  7. Analysis of potential common cause failure events for Romania-TRIGA 14 MW reactor

    Highlights: • Presents an overview of the general definitions and models for treating CCFs. • Includes a qualitative analysis of the CCFs events for the Romanian TRIGA reactor. • Performs a quantitative CCF analysis for the primary circulating pumps. - Abstract: Common cause failure (CCF) events can have significant impact in the availability of safety systems in nuclear power plants or research reactors. For these reasons CCF data are being collected and analyzed in many countries. The paper presents first the methods for CCF parameters estimation and mathematical models for estimation of the probability of occurrence of common cause events. Then, the paper describes data collection and statistics referring to the potential CCF for components of Romania-TRIGA SSR 14 MW reactor. Potential CCF's components for Romania-TRIGA reactor involves pumps, control rods and control rods mechanisms, valves, fans, ejectors. Every group of components susceptible to CCF is discussed taking into account the event attributes like root cause, coupling factor, detection method and corrective action taken. An example of quantitative CCF analysis using different mathematical models for main circulating pumps is also considered

  8. Modeling TRIGA reactor pulses using the STAR 3D nodal kinetics and WIMS-D4 codes

    A detailed three-dimensional (3D) time-dependent STAR nodal kinetics model coupled to a one-dimensional (1D) thermal-hydraulics WIGL model has been developed to describe and benchmark the peak power and pulse behavior of the Penn State University (PSU) Breazeale TRIGA reactor. Different core loading patterns were used for several TRIGA pulse tests with different reactivity insertion worths (1.5 dollar, 2.0 dollar, 2.5 dollar). The STAR nodal kinetics code and TRIGA model adequately simulates TRIGA pulses when group constants are generated from physics codes (i.e., WIMS-D4) that can accurately model the TRIGA uranium-zirconium-hydride fuel

  9. Neutron Imaging Using Neutrons From TRIGA MARK II PUSPATI Reactor

    This article reports about the implementation of neutron imaging work utilizing neutron beam from TRIGA MARK II PUSPATI collimation channels. Two methods have been implemented namely radiography and tomography. Advantage of these methods is the fact that, radiograms are obtained from normal radiographic imaging methodology and they are the projections used for tomographic image reconstruction. Therefore, both radiogram and tomogram are obtained consecutively. The method has been implemented on the round robin test sample for contrast and resolution measurement and also to some archaeological objects. (author)

  10. Safe operation of a TRIGA reactor in the situation of LEU-HEU core conversion

    Romanian TRIGA reactor was commissioned in 1980. The location of the research institute is Pitesti, 100 Km west of Bucharest. In fact there are two independent cores sharing the same pool. There are a 14 MW Steady State Reactor (SSR), high flux, and materials testing reactor and an Annular Core Pulsing Reactor (ACPR). The SSR reactor is a forced convection reactor cooled via a primary circuit with 4 pumps and 3 heat exchangers. The ACPR is natural convection cooled by the pool water. Modifications performed concerning core configuration resulted in the following. Removal the central pin from the bundle leads to slightly temperature increase of approximately 1% for the corner and edge pins, for the same pin power density. Also, the temperature slightly decreases for the 4 pins adjacent to the water hole. This is caused by the coolant flow redistribution. But, according to preliminary neutronic computations, PPF-s are decreasing, the edge and corner temperatures changes are no more detectable. DNB are decreasing, leading to a safer operation. Fuel management of TRIGA steady state core allows to obtain the requested fluxes for experimental reasons in the safer operation conditions. We can firmly state that the present operation of the reactor and the HEU-LEU core conversion fully respect the provisions of the National Regulatory Body and the IAEA. On the other side, we have to mention the common fact that research reactors cannot sustain themselves in the financial domain. The lack of sufficient financial support leads to shortage of the maintenance programs and to reduce of activities and personnel member; this is a real danger in maintaining the actual standards of nuclear safety. During this transition period, the Romanian TRIGA reactor is used much its capability in the frame of international cooperation this facility can ensure support for various research programmes in the fields of interest

  11. 10th European TRIGA users conference

    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)

  12. Fuel element burn-up calculation in ITU TRIGA Mark-II reactor

    The reactivity defect of fuel elements in ITU TRIGA Mark-II reactor core at 250 kW power have been calculated by considering the reactor operation history. A two-dimensional, four-group diffusion computer code TRIGLAV is used for the calculations. The unit-cell macroscopic cross sections and diffusion coefficients are generated with the WIMS-D/4 code. Two dimensional effects like vicinity of control rods, water gaps, dummy graphite elements, void channels are considered. The calculated reactivity worth of the fuel elements at known burn up are in agreement with experimental values of the fuel elements located in the reactor core without two dimensional effects. (author)

  13. Unique applications of research reactors with TRIGA UZrHx fuel

    The TRIGA reactor fuel (UZrHx) in research reactors provides significant safety features that have permitted varied and unique applications. The safety features include a very large, prompt, negative temperature coefficient of reactivity; very high safety limit for fuel temperature (1150 degrees C); and large fission product retention even for unclad fuel. The recognized safety of these reactors has permitted them to be located as appropriate on university campuses in buildings housing lecture halls and in hospitals. It has also facilitated installation of in-core or near-core experiments and facilities, including liquid hydrogen or other cryogenic neutron sources

  14. Operating experience of TRIGA MK-II Research Reactor in Bangladesh

    A 3 MW TRIGA MK II Research Reactor was installed in Bangladesh in 1986. The reactor is being utilized for research, training and for production of radioisotopes. Recently two faults were detected, one in the Emergency Core Cooling System and the other in the Primary Coolant Loop, which hindered the operation of the reactor partially. The faults were investigated by a team of local experts. Results of analyses of possible initiating events of the faults and the remedial steps are briefly discussed in the paper. (author)

  15. Investigations of cracks in the shielding concrete of a TRIGA Mark II reactor

    Cracks in the reactor shielding concrete of the TRIGA Mark II reactor, Vienna, caused an experimental and theoretical program to investigate the crack reason. After the investigation of the mechanical concrete data, the crack motion was measured as a function of various environmental temperatures. The temperature stress in the concrete was calculated analytically and with the finite-elements method and good accordance with the actual crack distribution was found. Finally some possibilities to avoid concrete cracks in future research reactor shielding construction are outlined. (orig.)

  16. Sensitivity analysis of the RELAP5 nodalization to IPR-R1 TRIGA research reactor

    The main aim of this work is to identify how much the code results are affected by 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 previous 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 model. The results found highlight the necessity of sensitivity analysis to obtain the ideal simulation model of a system. (author)

  17. Retrofitting the instrumentation and control system of primary cooling circuit from TRIGA INR 14 MW reactor

    Activities of retrofitting the instrumentation and control system from TRIGA INR primary cooling circuit consists in replacement of actual system for: - parameter measurement; - safety; - reactor external scramming; - protection, command and supply for electrical elements of the system. This retrofitting project is designed to ensure the necessary features of reactor external safety and for technological parameter measurement. The new safety system of main cooling circuit is completely separated from its operating system and is arranged in a panel assembly in reactor control room. The operating system has the following features: - data acquisition; - parameter value and state of command elements displaying; - command elements on hierarchical levels; - operator information through visual and acoustic alarm. (authors)

  18. Proposed design for the PGAA facility at the TRIGA IPR-R1 research reactor

    Guerra, Bruno T; Jacimovic, Radojko; Menezes, Maria Angela BC; Leal, Alexandre S.

    2013-01-01

    Background This work presents an initial proposed design of a Prompt Gamma Activation Analysis (PGAA) facility to be installed at the TRIGA IPR-R1, a 60 years old research reactor of the Centre of Development of Nuclear Technology (CDTN) in Brazil. The basic characteristics of the facility and the results of the neutron flux are presented and discussed. Findings The proposed design is based on a quasi vertical tube as a neutron guide from the reactor core, inside the reactor pool, 6 m below t...

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

    Rivero-Gutiérrez, Tonatiuh; Benítez-Read, Jorge S.; Segovia-De-los-Ríos, Armando; Longoria-Gándara, Luis C.; Palacios-Hernández, Javier C.

    2012-01-01

    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 ...

  20. Experimental power density distribution benchmark in the TRIGA Mark II reactor

    Snoj, L.; Stancar, Z.; Radulovic, V.; Podvratnik, M.; Zerovnik, G.; Trkov, A. [Josef Stefan Inst., Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Barbot, L.; Domergue, C.; Destouches, C. [CEA DEN, DER, Instrumentation Sensors and Dosimetry laboratory Cadarache, F-13108 Saint-Paul-Lez-Durance (France)

    2012-07-01

    In order to improve the power calibration process and to benchmark the existing computational model of the TRIGA Mark II reactor at the Josef Stefan Inst. (JSI), a bilateral project was started as part of the agreement between the French Commissariat a l'energie atomique et aux energies alternatives (CEA) and the Ministry of higher education, science and technology of Slovenia. One of the objectives of the project was to analyze and improve the power calibration process of the JSI TRIGA reactor (procedural improvement and uncertainty reduction) by using absolutely calibrated CEA fission chambers (FCs). This is one of the few available power density distribution benchmarks for testing not only the fission rate distribution but also the absolute values of the fission rates. Our preliminary calculations indicate that the total experimental uncertainty of the measured reaction rate is sufficiently low that the experiments could be considered as benchmark experiments. (authors)

  1. Modification of the radial beam port of ITU TRIGA Mark II research reactor for BNCT applications

    This paper aims to describe the modification of the radial beam port of ITU (İstanbul Technical University) TRIGA Mark II research reactor for BNCT applications. Radial beam port is modified with Polyethylene and Cerrobend collimators. Neutron flux values are measured by neutron activation analysis (Au–Cd foils). Experimental results are verified with Monte Carlo results. The results of neutron/photon spectrum, thermal/epithermal neutron flux, fast group photon fluence and change of the neutron fluxes with the beam port length are presented. - Highlights: • Using MCNP5, radial beam port of ITU TRIGA Mark II research reactor is modified. • Polyethylene and Cerrobend collimators are used to modify the beam port. • Results of two-group neutron/photon flux are presented. • Monte Carlo results are compared with experimental results

  2. Experimental power density distribution benchmark in the TRIGA Mark II reactor

    In order to improve the power calibration process and to benchmark the existing computational model of the TRIGA Mark II reactor at the Josef Stefan Inst. (JSI), a bilateral project was started as part of the agreement between the French Commissariat a l'energie atomique et aux energies alternatives (CEA) and the Ministry of higher education, science and technology of Slovenia. One of the objectives of the project was to analyze and improve the power calibration process of the JSI TRIGA reactor (procedural improvement and uncertainty reduction) by using absolutely calibrated CEA fission chambers (FCs). This is one of the few available power density distribution benchmarks for testing not only the fission rate distribution but also the absolute values of the fission rates. Our preliminary calculations indicate that the total experimental uncertainty of the measured reaction rate is sufficiently low that the experiments could be considered as benchmark experiments. (authors)

  3. A parametric thermal-hydraulic analysis of I.T.U. TRIGA Mark-II reactor

    In this study, a transient, one-dimensional thermal-hydraulic subchannel analysis for I.T.U. TRIGA Mark-II reactor was employed. The cooling of this reactor is based on natural convection; however, mixed convection is considered in modeling in order to enhance the capability of the computer code. After the continuity, conservation of energy, momentum balance equations for coolant in axial direction and heat conduction equation for fuel rod in radial direction had been written, they were discretized by using the control volume approach to obtain a set of algebraic equations. By the aid of discretized continuity and momentum balance equations, a pressure correction equation was derived. Then, a FORTRAN program called TRIGATH (TRIGA Thermal-Hydraulics) has been developed to solve this set of algebraic equations by using SIMPLE algorithm. As a result, the temperature distributions of the coolant and fuel rods as well as the velocity and pressure distributions of the coolant have been estimated. (authors)

  4. A study on site release criterion assessment of nuclear power facilites for TRIGA research reactor decommissioning

    The process of establishing the site release criterion in MARSSIM is a guide which makes a decision if the contamination level of the building in the site meets guide level, so it is able to classify the contamination site with the expected contamination level in facility site as the process to raise the working efficiency with applying to the site facility building of TRIGA research reactor on the progress of the internal decommissioning plan. It is unreasonable to establish the criterion for site recycling so far due to the lack of survey because the decommissioning plan of TRIGA research reactor is still on the progress. But it is able to design process to establish the site recycling criterion according to survey result with using the method to decide survey quantity and location in MARSSIM process guide

  5. 15. European TRIGA Conference

    The 15th European TRIGA Conference was organised by the VTT Chemical Technology and held in June 15-17, 1998, in Espoo, Finland. The topics of the conference included: reactor operation and maintenance experience, developments and improvements of TRIGA components, safety aspects, licensing, radiation protection, fuel management, personnel, training programmes, and research programmes at TRIGA stations. The special topic of the conference was TRIGA reactors and the Boron Neutron Capture Therapy (BNCT)

  6. 15. European TRIGA Conference

    Salmenhaara, S. (ed.)

    1999-12-15

    The 15th European TRIGA Conference was organised by the VTT Chemical Technology and held in June 15-17, 1998, in Espoo, Finland. The topics of the conference included: reactor operation and maintenance experience, developments and improvements of TRIGA components, safety aspects, licensing, radiation protection, fuel management, personnel, training programmes, and research programmes at TRIGA stations. The special topic of the conference was TRIGA reactors and the Boron Neutron Capture Therapy (BNCT)

  7. Contribution of the TRIGA - INR Pitesti reactor to implementation of National Program for Nuclear Power

    The TRIGA reactor of INR Pitesti, designed for nuclear fuel and structural materials testing, was commissioned in 1979. It has two reactor cores completely independent that share the same pool, which are practically two distinct nuclear reactors, a 14 MW steady-state unit and a second pulse reactor, working at 20,000 MW/pulse. The last unit may also be operated in a steady state regime of 500 KW power. Being of pool type it allows easy handling of the irradiation devices. The two TRIGA reactors are described. These reactors are used mainly for irradiation testing, particularly, for Cernavoda NPP fuel elements and production of radioisotopes as, for instance, Ir-131, Mo-99 for non-destructive industrial analyses and Co-60 for cobalt therapy. Also, programs for experimental physical research were developed as for instance crystallographic studies by means of neutron diffraction, prompt gamma spectrometry for isotopic composition of Ga poison at Cernavoda NPP and neutron activation analyses. A program RERTR is now undergoing for converting the highly enriched-fuel research reactors into slightly enriched-fuel reactors. This project is developed in the frame of a cooperation with IAEA-Vienna and DOE - USA. Also a program of power cycling testing for the study of CANDU fuel power followup in collaboration with AECL - Canada is currently implemented. Several research programs were established aiming at testing slightly enriched fuel for Cernavoda NPP, testing of CANDU fuel in LOCA accident conditions, preparation of radiopharmaceuticals, etc

  8. TRIGA reactor spent fuel pool under severe earthquake conditions

    Supplemental criticality safety analysis of a pool type storage for TRIGA spent fuel at 'Jozef Stefan' Institute in Ljubljana, Slovenia, is presented. Previous results (Ravnik, M, Glumac, B., 1996) have shown that subcriticality is not guaranteed for some postulated accidents. To mitigate this deficiency, a study was made about replacing a certain number of fuel elements in the rack with absorber rods (Glumac, B., Ravnik, M., Logar, M., 1997) to lower the supercriticality probability, when the pitch is decreased to contact (as a consequence of a severe earthquake) in a square arrangement. The criticality analysis for the hexagonal contact pitch is presented in this paper, following the same scenario as outlined above. The Monte Carlo computer code MCNP4B with ENDF-B/VI library and detailed three dimensional geometry was used. First, the analysis about the influence of the number of triangular fuel piles on the bottom that could appear, if the fuel rack, made of three segments, disintegrates, is presented. Next, the number of uniformly mixed absorber rods in the lattice needed to sustain the subcriticality of the storage for hexagonal contact pitch is studied. Because of supercriticality possibility due to random mixing of the absorber rods in the case of lattice compaction, a probabilistic study was made in order to sample the probability density functions for random lattice loadings of the absorber rods. The results show that reasonably low probabilities for supercriticality can be achieved even when fresh 12 wt.% standard TRIGA fuel is stored in the spent fuel pool. (orig.)

  9. ENEA TRIGA RC-1 Reactor spent fuel elements shipment in USA

    In the second half of July 1999, 140 spent fuel elements of ENEA TRIGA RC-1 reactor were shipped in the United State for the final storage. DOE, NAC, TRANSNUCLEAIRE, MIT and ENEA competencies for own questions are described. The activities planning, the organizing aspect study, the analysis and valuations of both nuclear safety and radioprotection have been suitable for the TRIGA RC-1 fuel element shipment. In fact, no operative anomaly appeared with respect to the approved procedures. Personnel engagement has been as expectations and the personnel absorbed gamma dose resulted to be negligible. Lastly, the NAC disposable narrow time (only one week at the end of July) has not produced heavy organizational problems but it has been a strong goad per all operative structures involved in the TRIGA RC-1 elements shipment. Further, the report describes the Italian Regulatory Body (ANPA) requests in order to obtain the maximum safety conditions. A brief description of the safety analysis prepared by TRIGA staff is also presented. (authors)

  10. Using TRIGA Mark II research reactor for irradiation with thermal neutrons

    Kolšek, Aljaž, E-mail: aljaz.kolsek@gmail.com; Radulović, Vladimir, E-mail: vladimir.radulovic@ijs.si; Trkov, Andrej, E-mail: andrej.trkov@ijs.si; Snoj, Luka, E-mail: luka.snoj@ijs.si

    2015-03-15

    Highlights: • Monte Carlo N-Particle Transport Code was used to design and perform calculations. • Characterization of the TRIGA Mark II ex-core irradiation facilities was performed. • The irradiation device was designed in the TRIGA irradiation channel. • The use of the device improves the fraction of thermal neutron flux by 390%. - Abstract: Recently a series of test irradiations was performed at the JSI TRIGA Mark II reactor for the Fission Track-Thermoionization Mass Spectrometry (FT-TIMS) method, which requires a well thermalized neutron spectrum for sample irradiation. For this purpose the Monte Carlo N-Particle Transport Code (MCNP5) was used to computationally support the design of an irradiation device inside the TRIGA model and to support the actual measurements by calculating the neutron fluxes inside the major ex-core irradiation facilities. The irradiation device, filled with heavy water, was designed and optimized inside the Thermal Column and the additional moderation was placed inside the Elevated Piercing Port. The use of the device improves the ratio of thermal neutron flux to the sum of epithermal and fast neutron flux inside the Thermal Column Port by 390% and achieves the desired thermal neutron fluence of 10{sup 15} neutrons/cm{sup 2} in irradiation time of 20 h.

  11. Using TRIGA Mark II research reactor for irradiation with thermal neutrons

    Highlights: • Monte Carlo N-Particle Transport Code was used to design and perform calculations. • Characterization of the TRIGA Mark II ex-core irradiation facilities was performed. • The irradiation device was designed in the TRIGA irradiation channel. • The use of the device improves the fraction of thermal neutron flux by 390%. - Abstract: Recently a series of test irradiations was performed at the JSI TRIGA Mark II reactor for the Fission Track-Thermoionization Mass Spectrometry (FT-TIMS) method, which requires a well thermalized neutron spectrum for sample irradiation. For this purpose the Monte Carlo N-Particle Transport Code (MCNP5) was used to computationally support the design of an irradiation device inside the TRIGA model and to support the actual measurements by calculating the neutron fluxes inside the major ex-core irradiation facilities. The irradiation device, filled with heavy water, was designed and optimized inside the Thermal Column and the additional moderation was placed inside the Elevated Piercing Port. The use of the device improves the ratio of thermal neutron flux to the sum of epithermal and fast neutron flux inside the Thermal Column Port by 390% and achieves the desired thermal neutron fluence of 1015 neutrons/cm2 in irradiation time of 20 h

  12. Development of U Zr alloy for the TRIGA/IPR-R1 reactor fuel

    This paper reports the fabrication development at CDTN of the UZr alloy for the TRIGA/IPR-R1 reactor fuel. A comparative study of the melting of UZr alloy by using vacuum consumable-electrode arc (VAR) and vacuum induction melting (VIM) process, it was necessary to remelt the ingot to homogenize the alloy. The influence of the observed contamination by c in the vim process on the alloy neutronic and mechanical properties is a case for further studies. (author)

  13. Steady-State Thermal-Hydraulic Analysis of TRIGA Research Reactor

    Mohammad Mizanur Rahman; Mohammad Abdur R. Akond; Mohammad Khairul Basher; Md. Quamrul Huda

    2014-01-01

    The COOLOD-N2 and PARET computer codes were used for a steady-state thermal hydraulic and safety analysis of the 3 MW TRIGA Mark-II research reactor located at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh. The objective of the present study is to ensure that all important safety related thermal hydraulic parameters uphold margins far below the safety limits by steady-state calculations at full power. We, therefore, have calculated the hot channel fuel centreline ...

  14. Analysis of JSI TRIGA MARK II reactor physical parameters calculated with TRIPOLI and MCNP.

    Henry, R; Tiselj, I; Snoj, L

    2015-03-01

    New computational model of the JSI TRIGA Mark II research reactor was built for TRIPOLI computer code and compared with existing MCNP code model. The same modelling assumptions were used in order to check the differences of the mathematical models of both Monte Carlo codes. Differences between the TRIPOLI and MCNP predictions of keff were up to 100pcm. Further validation was performed with analyses of the normalized reaction rates and computations of kinetic parameters for various core configurations. PMID:25576735

  15. Optical inspection and maintenance of the triga mark-II reactor in Pavia/Italy

    The TRIGA reactor Pavia was optically inspected using the underwater endoscope. Problems which required this inspection were a loose control rod guide tube connection to the lower grid plate and a deformed central irradiation thimble which prevented the removal of this tube through the upper grid plate. Both problems were resolved by optically inspecting the inner core area. Using special tools both pieces were repaired. In addition the tank was cleaned and debris was removed from the tank. (author)

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

    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

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

    C. A. M. Silva; J. A. D. Salomé; B. T. Guerra; Pereira, C; Costa, A. L.; Veloso, M. A. F.; M. A. B. C. Menezes; Dalle, H. M.

    2014-01-01

    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 sma...

  18. Determination of the neutron fluxes in the research nuclear reactors: the Triga Mark I and the WRS

    In this paper is presented the determination of the thermal, epithermal and fast neutron fluxes, using neutron activation analysis technique, for two research nuclear reactors of different design: the TRIGA Mark I reactor was designed by Gulf General Atomic Co in US and the W WRS reactor was designed in the URSS, both in the 50's years. (Author)

  19. Research Reactor Spent Fuel Transfer/Storage Cask with Application to TRIGA Fuel - Designed Cask Shielding Independent Evaluation

    Margeanu, C.A.; Iorgulis, C. [Reactor Physics, Nuclear Fuel Performances and Nuclear Safety Department, Institute for Nuclear Research Pitesti, Campului Street, no.1, 115400 Mioveni (Romania); Ciocanescu, M. [Institute for Nuclear Research Pitesti, Campului Street, no.1, 115400 Mioveni (Romania); Prava, M. [Design Department, Institute for Nuclear Research Pitesti, Campului Str, no.1, 115400 Mioveni (Romania); Margeanu, S. [Radiation Protection Department, Institute for Nuclear Research Pitesti, Campului Street, no.1, 115400 Mioveni (Romania)

    2011-07-01

    Institute for Nuclear Research (INR) Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies (dual-core concept involves independent operation of TRIGA 14 MW Steady-State Reactor and TRIGA Annular-Core Pulsing Reactor at each end of a large pool). In May 2006, TRIGA 14 MW SSR core was fully converted to Low Enriched Uranium (LEU 20 wt% {sup 235}U) fuel, according to Reduced Enrichment for Research and Test Reactors agreements and current worldwide non-proliferation efforts. Paper presents a shielding independent evaluation applied to designed transfer/ storage cask for TRIGA INR spent fuel, a mandatory step in preparation of the documentation required for spent fuel transfer/storage cask authorisation process. Fuel elements irradiation was modelled by assuming constant power for entire residence time inside reactor core, for 14 MW reactor operation power and two different scenarios characteristic for accident calculations according to TRIGA 14 MW SSR safety report and reactor operation experience. The discharged spent LEU fuel was cooled down for 2 and 5 years, respectively. Source term assessment and spent fuel characteristic parameters estimation were done by means of ORIGEN-S burn-up code (included in Oak Ridge National Laboratory's SCALE6 package) with specific cross-sections libraries, updating data for each burn-up step. For the transfer/storage cask shielding analysis, two different cases have been considered, the main difference residing in TRIGA fuel elements loading. The radiation dose rates to the transfer/storage cask wall and in air at different distances from the cask have been estimated by means of MAVRIC/Monaco shielding 3D Monte Carlo code included in ORNL's SCALE6 package. (author)

  20. Research Reactor Spent Fuel Transfer/Storage Cask with Application to TRIGA Fuel - Designed Cask Shielding Independent Evaluation

    Institute for Nuclear Research (INR) Pitesti owns and operates a TRIGA dual-core Research Reactor for material testing, power reactor fuel and nuclear safety studies (dual-core concept involves independent operation of TRIGA 14 MW Steady-State Reactor and TRIGA Annular-Core Pulsing Reactor at each end of a large pool). In May 2006, TRIGA 14 MW SSR core was fully converted to Low Enriched Uranium (LEU 20 wt% 235U) fuel, according to Reduced Enrichment for Research and Test Reactors agreements and current worldwide non-proliferation efforts. Paper presents a shielding independent evaluation applied to designed transfer/ storage cask for TRIGA INR spent fuel, a mandatory step in preparation of the documentation required for spent fuel transfer/storage cask authorisation process. Fuel elements irradiation was modelled by assuming constant power for entire residence time inside reactor core, for 14 MW reactor operation power and two different scenarios characteristic for accident calculations according to TRIGA 14 MW SSR safety report and reactor operation experience. The discharged spent LEU fuel was cooled down for 2 and 5 years, respectively. Source term assessment and spent fuel characteristic parameters estimation were done by means of ORIGEN-S burn-up code (included in Oak Ridge National Laboratory's SCALE6 package) with specific cross-sections libraries, updating data for each burn-up step. For the transfer/storage cask shielding analysis, two different cases have been considered, the main difference residing in TRIGA fuel elements loading. The radiation dose rates to the transfer/storage cask wall and in air at different distances from the cask have been estimated by means of MAVRIC/Monaco shielding 3D Monte Carlo code included in ORNL's SCALE6 package. (author)

  1. Treatment and conditioning of radioactive wastes resulting from the TRIGA reactor

    The technologies utilized by ICN Pitesti in treating radioactive wastes are the following: - The treatment of liquid radioactive wastes resulting from the TRIGA reactor by evaporation and the conditioning by embedding in concrete of the radioactive concentrate; - The evaporation of liquid effluents is done in a 1.2 m3/h capacity evaporator, supplied by PEC Engineering, France; - The radioactive concentrate is conditioned, in view of its final disposal, by concrete-embedding in 220 l capacity metallic drums; - For treating and conditioning solid radioactive wastes resulting from the TRIGA reactor and from the Post-Irradiation Examination Labs, their embedding in concrete is utilized, in view of their final disposal. The wastes are separated into two classes, i.e. compact solid wastes, pressed in a die to reduce volume thus obtaining max. 5 l compacts, and non-compact solid wastes, cut in pieces of 700x400x400 mm. The compact and/or crushed wastes are put into a metal basket which is conditioned by embedding in concrete for the final disposal in 220 l drums; - Bituminization is used for treating and conditioning of spent ion exchangers resulting from the operation of the TRIGA reactor. This is done in a 80 l drum which is conditioned in the 220 l drum for the final disposal. (authors)

  2. Feasibility analysis of I-131 production in the Moroccan TRIGA research reactor

    Highlights: • A feasibility analysis for I-131 production at the Moroccan TRIGA MARK II research reactor was conducted. • Two production scenarios were discussed with several TeO2 target masses. • The MCNPX v2.7 computer code with its depletion capabilities was used. • A production activity of about 4.63 Ci per 80 MWh irradiation period is obtained. - Abstract: Since the commissioning of the Moroccan 2 MW TRIGA MARK II research reactor hosted by the Centre National de l’Energie des Sciences et des Techniques Nucléaires (CNESTEN), the latter institution has established a radioisotope production program to supply radiopharmaceuticals for use in nuclear medicine. This paper presents a feasibility analysis for I-131 production using two in-core irradiation positions within the Moroccan TRIGA MARK II research reactor. The MCNPX v2.7 code, with its depletion capabilities, was used for the evaluation of two different production scenarios using several masses of TeO2 target samples. The maximum achievable activities were found to be 3.90 Ci/week for scenario 1 and 4.63 Ci/week for scenario 2. Thermal analysis shows that safety limits of capsules used for these experiments were not violated

  3. A complete fuel development facility utilizing a dual core TRIGA reactor system

    A TRIGA Dual Core Reactor System has been chosen by the Romanian Government as the heart of a new fuel development facility which will be operated by the Romanian Institute for Nuclear Technologies. The Facility, which will be operational in 1976, is an integral part of the Romanian National Program for Power Reactor Development, with particular emphasis being placed on fuel development. The unique combination of a new 14 MW steady state TRIGA reactor, and the well-proven TRIGA Annular Core Pulsing Reactor (ACPR) in one below-ground reactor pool resulted in a substantial construction cost savings and gives the facility remarkable experimental flexibility. The inherent safety of the TRIGA fuel elements in both reactor cores means that a secondary containment building is not necessary, resulting in further construction cost savings. The 14 MW steady state reactor gives acceptably high neutron fluxes for long- term testing of various prototype fuel-cladding-coolant combinations; and the TRIGA ACPR high pulse capability allows transient testing of fuel specimens, which is so important for accurate prediction of the performance of power reactor fuel elements under postulated failure conditions. The 14 MW steady state reactor has one large and three small in-core irradiation loop positions, two large irradiation loop positions adjacent to the core face, and twenty small holes in the beryllium reflector for small capsule irradiation. The power level of 14 MW will yield peak unperturbed thermal neutron fluxes in the central experiment position approaching 3.0 x 1014 n/cm2-sec. The ACPR has one large dry central experimental cavity which can be loaded at pool level through a shielded offset loading tube; a small diameter in-core flux trap; and an in-core pneumatically-operated capsule irradiation position. A peak pulse of 15,000 MW will yield a peak fast neutron flux in the central experimental cavity of about 1.5 x 1017 n/cm2-sec. The pulse width at half maximum during a

  4. Renewal and upgrading of the TRIGA Mark II research reactor in Ljubljana

    At the 250 kW TRIGA Mark II research reactor in Ljubljana, ever since the beginning of operation in 1966, gradual modification and modernization have been taking place. In 1991 the reactor has been almost completely reconstructed and upgraded. The reconstruction consisted mainly of replacing the grid plates, the control rod mechanisms and the control unit. A new PC based system to collect the operational radiation data of the reactor was developed. A new spent fuel storage facility was built in the basement of the reactor building with a capacity of 630 spent fuel elements. The main novelty in the reactor physics and operational features of the reactor was installation of the pulse rod. The following experiments were conducted: initial criticality, excess reactivity measurement, control rod worth measurement, fuel temperature distribution, fuel temperature reactivity coefficient, pulse parameter measurements (peak power, prompt energy, peak temperature). Flux distributions in steady state and pulse mode were measured as well. The experiments were performed with completely fresh fuel of 12 w% Standard Stainless Steel type. The core configuration was uniform (one fuel element type, including fuelled followers) and compact (no irradiation channels or gaps), as such an array is particularly convenient for testing computer codes for TRIGA reactor calculations

  5. Operation experience with the TRIGA Mark II reactor Vienna in the years 1972 through 1974

    Since the last TRIGA Users Conference in Pavia 1972 the TRIGA reactor Vienna was in operation without any larger undesired shut-down. The integral thermal power production by Sept. 1, 1974 was 3420 MWh. The principal work carried out during the last two years on the reactor system was the installation of a new heat exchanger and primary pump both designed for 1 MW steady state operation. Permission was also obtained from the local authority to withdraw up to 90 m3/h secondary cooling water from the well. Some troubles were observed with the pulse rod. After nearly 12 years of operation the connection between the piston rod and control rod broke off just below the water surface. Therefore the piston was shot out without withdrawing the pulse rod itself. After locating the trouble the damage was repaired within one day. The SST fuel elements type 110 were received by the end of 1972 for the purpose of power upgrading. All other fuel elements except one are still located in the reactor core and shifted periodically in order to obtain an optimal burnup. A new alarm system was ordered from Hartmann and Braun and is under installation at the moment. In order to facilitate cooperation with the reactor operation personnel and the experimenters in the reactor hall an accurate power indicator has been installed in the reactor hall which allows all experimenters to read the reactor power as accurately as in the control room itself. (U.S.)

  6. EVALUATION OF COOLING INSTRUMENTATION SYSTEM OF TRIGA MARK II REACTOR OF BANDUNG

    Evaluation of cooling instrumentation system of Triga Mark II reactor has been done. The reactor has been upgraded from 1 MW to 2 MW. The increasing of power is performed by changing the reactor components and systems. The reactor cooling system has important role in reactor operation, the system transfers heat produced in the core. The operation of the cooling system needed to be back up with qualified instrumentation. Evaluation has been done by doing analysis and observing the equipment design, type and clarification, performance study of instrumentation and system related to cooling system. It is known that the performance and system of Triga mark II reactor included the cooling system. It is also obtained the characteristic data of primary and secondary cooling system, piping diagram and instrumentation, emergency core cooling system. The cooling system has 4 measurement, i.e. flow rate, input and output temperature to heat exchanger, and electricity conductivity of water. The measurement can be observed from the reactor console. From this evaluation it is concluded that cooling system instrumentation followed the required criteria

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

    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

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

    Ciocaanescu, M.; Ionescu, M. [Inst. for Nuclear Research, Pitesti (Romania). TRIGA Reactor Dept.

    1996-08-01

    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{sub 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{sub 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.

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

    The cooperation between Romania and the USA in the field of technologic transfer of nuclear research reactor technology began with the steady state 14 MWt 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 MWt 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

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

    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)

  11. PUSPATI TRIGA Reactor Upgrading: Towards the Safe Operation and Feasibility of Neutronic Approach

    The PUSPATI TRIGA Reactor (RTP) has been safely operated for the last 29 years with no incidents as listed in the unusual event reporting categories being reported. However, in order to maintain the reactor integrity and safety, several reactor components and system were refurbished or replaced over the years. The latest approach to enhance safety was the replacement of the heat exchanger from previously shell and tube heat exchanger to plate-type heat exchanger. Prior to have this reactor extensively and safely utilised, the feasibility of neutronic approach to upgrade the reactor have been carried out and reported. This paper will describe the strategies for ensuring prolonged and continuously safe operations of the reactor and also discussed on the feasible of the neutronics approach towards the safe operation. (author)

  12. TRIGA reactor spent fuel pool under severe earthquake conditions

    Logar, M. [Univ. of Maribor (Slovenia). Fac. of Elec. Eng.; Glumac, B.; Maucec, M. [`Jozef Stefan` Institute, Jamova 39, POB 100, 1111 Ljubljana (Slovenia)

    1998-07-01

    Supplemental criticality safety analysis of a pool type storage for TRIGA spent fuel at `Jozef Stefan` Institute in Ljubljana, Slovenia, is presented. Previous results (Ravnik, M, Glumac, B., 1996) have shown that subcriticality is not guaranteed for some postulated accidents. To mitigate this deficiency, a study was made about replacing a certain number of fuel elements in the rack with absorber rods (Glumac, B., Ravnik, M., Logar, M., 1997) to lower the supercriticality probability, when the pitch is decreased to contact (as a consequence of a severe earthquake) in a square arrangement. The criticality analysis for the hexagonal contact pitch is presented in this paper, following the same scenario as outlined above. The Monte Carlo computer code MCNP4B with ENDF-B/VI library and detailed three dimensional geometry was used. First, the analysis about the influence of the number of triangular fuel piles on the bottom that could appear, if the fuel rack, made of three segments, disintegrates, is presented. Next, the number of uniformly mixed absorber rods in the lattice needed to sustain the subcriticality of the storage for hexagonal contact pitch is studied. Because of supercriticality possibility due to random mixing of the absorber rods in the case of lattice compaction, a probabilistic study was made in order to sample the probability density functions for random lattice loadings of the absorber rods. The results show that reasonably low probabilities for supercriticality can be achieved even when fresh 12 wt.% standard TRIGA fuel is stored in the spent fuel pool. (orig.) 7 refs.

  13. Validating the Serpent Model of FiR 1 Triga Mk-II Reactor by Means of Reactor Dosimetry

    Viitanen Tuomas; Leppänen Jaakko

    2016-01-01

    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 ...

  14. Analysis of safety limits of the Moroccan TRIGA MARK II research reactor

    A 2 MW TRIGA MARK II research reactor has been designed by General Atomics (GA) for the National Centre for Energy and Nuclear Sciences and Techniques (CNESTEN) in Morocco. This TRIGA reactor has the particularity of being the only TRIGA reactor designed to operate at the power level of 2 MW with the use of natural convection cooling. The main objective of this study is to check the ability of the reactor to operate at its nominal power with sufficient safety margins. The analysis of the reactor core starts from the basic reactor cells calculations which were performed for all the reactor cells using the LEOPARD code. The zone averaged group constants provided by cell calculations are used to compute the multiplication factor keff of the cold and clean core using the diffusion theory code Mcrac which is a recent version of the earlier code EXTERMINATOR-2. The main objective of the core calculations is to predict the core excess reactivity in cold zero power conditions and the power peaking factors which are very important data for the thermal hydraulic analysis of the core. For the maximum power peaking factors, our results agree with the values given by the reactor designer. Concerning the core excess reactivity, our results from both XY and RZ core calculations models lead to higher values than the results given by GA (about +2000 pcm). However, we should mention that GA results correspond probably to the minimum core excess reactivity which is guaranteed. The thermal hydraulic analysis of the TRIGA core was performed using the French code FLICA developed in CEA/Saclay for transient and study state thermal hydraulic analysis of a large variety of reactor cores. The objective of this analysis is to evaluate the main safety related parameters of the core and to ensure that they are within the safety limits in any operating conditions. The parameters considered in our study are: maximum fuel temperature, minimum DNBR and maximum void fraction. The obtained results

  15. Measured and calculated effective delayed neutron fraction of the IPR-R1 Triga reactor

    Souza, Rose Mary G.P.; Dalle, Hugo M.; Campolina, Daniel A.M., E-mail: souzarm@cdtn.b, E-mail: dallehm@cdtn.b, E-mail: campolina@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The effective delayed neutron fraction, {beta}{sub eff}, one of the most important parameter in reactor kinetics, was measured for the 100 kW IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil. The current reactor core has 63 fuel elements, containing about 8.5% and 8% by weight of uranium enriched to 20% in U{sup 235}. The core has cylindrical configuration with an annular graphite reflector. Since the first criticality of the reactor in November 1960, the core configuration and the number of fuel elements have been changed several times. At that time, the reactor power was 30 kW, there were 56 fuel elements in the core, and the {beta}{sub eff} value for the reactor recommended by General Atomic (manufacturer of TRIGA) was 790 pcm. The current {beta}{sub eff} parameter was determined from experimental methods based on inhour equation and on the control rod drops. The estimated values obtained were (774 {+-} 38) pcm and (744 {+-} 20) pcm, respectively. The {beta}{sub eff} was calculated by Monte Carlo transport code MCNP5 and it was obtained 747 pcm. The calculated and measured values are in good agreement, and the relative percentage error is -3.6% for the first case, and 0.4% for the second one. (author)

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

    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)

  17. Follow-up the commissioning of CENM TRIGA Mark II research reactor on safety level

    The follow-up of the commissioning of the CENM-TRIGA Mark II Reactor has been performed in conformance with national regulation and the IAEA standards. For this purpose, the CNESTEN established a safety committee to review all safety aspects during reactor commissioning and operation. A set of hold points was established in the commissioning program, typically at the end of each stage to ensure that (i) test results have been reviewed by the safety committee and meet acceptance criterion, and (ii) requirements for the performance of the following stage of the commissioning program reviewed and understood by all the parties

  18. TRIGA-III research reactor pool inspection using an underwater vehicle

    Song, T. K.; Lee, J. R.; Kim, S. H.; Yoon, J. S.; Lee, B. J. [KAERI, Taejon (Korea, Republic of)

    1999-10-01

    For the inspection of radioactivity at the nuclear reactor and spent fuel storage pool, an underwater vehicle system has been developed. This underwater vehicle is navigated freely by five thruster which are controlled by developed control system and has a faculty of radiation detection at the inner wall and special point in pool using the radiation detector which is attached to the bottom of the vehicle. In this paper, the developed underwater vehicle and its components are described in detail. Also, the field test result in TRIGA-III research reactor pool is described.

  19. Radioactive liquid waste treatment for decontamination and decommissioning of TRIGA research reactors

    Park, Seung Kook; Chung, K.H

    1999-04-01

    All of operated radioactive liquid waste will be stored by using existing collection tank and temporally transfer piping system before dismantle the TRIGA research reactors. In this paper, there are presented and discussed as follows; 1.The status of operated radioactive liquid waste. 2. The radioactive liquid waste during dismantle the reactor. 3. Radiological status of radioactive liquid waste. 4. The classification criteria and method radioactive liquid waste. 6. The collection and transportation of radioactive liquid waste. (Author). 13 refs., 13 tabs., 8 figs.

  20. Neutron spectrum and radial power distribution measurements in a TRIGA reactor fuel element

    The neutron spectrum in the Illinois Advanced TRIGA Reactor was measured by a crystal spectrometer utilizing an LiF(1, 1, 1) crystal monochromator whose reflectivity was determined experimentally. The fission heat source distribution in a fuel element was also determined as a function of the fuel element temperature. These two measurements were used to investigate the effects of fuel element temperature and the local core loading on the thermal diffusion length in a fuel element. Changes in the thermal diffusion lengths during a reactor pulse underlie the proposed temperature feedback mechanism for the ZrH fuel material. The results of the measurements confirm, in part, this proposed temperature feedback mechanism

  1. Characteristics and facilities of a 3MW LEU fuelled TRIGA reactor

    A 3 MW TRIGA reactor fuelled with low enriched uranium having 19.7 % U-235 and 20 wt% Uranium and Zirconium Hydride, has been installed and recently made critical at a research laboratory of the Bangladesh Atomic Energy Commission. This paper describes the basic design, low and high power test results and the facilities of the reactor. The details of the core configuration of the initial criticality with 50 elements and the final core with 100 elements at 3 MW power are discussed. The available experimental facilities are also described briefly. (author)

  2. Production and use of 18F by TRIGA nuclear reactor: a first report

    The irradiation and radiochemical facilities at public research centre can contribute to the start up of the regional PET centre. In particular, the TRIGA reactor of Casaccia Research Centre could produce a sufficient amount of 18F to start up a PET centre and successively integrated the cyclotron production. This report establishes, in the light of the preliminary experimental works, a guideline to the reactor's production and extraction of 18F in a convenient form for the synthesis of the most representative PET radiopharmaceutical: 18F-FDG

  3. Fuel burnup analysis of the TRIGA Mark II Reactor at the University of Pavia

    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

    2015-01-01

    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 analyse neutron fluxes in the reactor core. The software that took care of time evolution, completely designed in-house, used the n...

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

    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 108 n/ cm2/ s. According to IAEA (2001) flux of 1.00 x 109 n/ cm2/ s requires three hours of treatment. Few modifications were needed to get higher flux. (Author)

  5. Epithermal neutron flux characterization of the TRIGA MARK II reactor, Ljubljana, Yugoslavia, for use in NAA

    The nonideality of the epithermal neutron flux distribution at a reactor site can be described by a 1/E1+α spectrum representation, with parameter α as a measure of nonideality. α-values were determined in 3 typical irradiation positions of the TRIGA MARK II reactor, Ljubljana, Yugoslavia, using the 'Cd-ratio for multi-monitor' method. The simpler 'Cd-ratio for dual monitor' method also yielded reliable results. This characterization is useful in the ko-method of NAA. (author) 18 refs.; 3 figs

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

    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)

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

    The 69 groups constants of H in ZrH, 166Er and 167Er 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

  8. Radioactive liquid waste treatment for decontamination and decommissioning of TRIGA research reactors

    All of operated radioactive liquid waste will be stored by using existing collection tank and temporally transfer piping system before dismantle the TRIGA research reactors. In this paper, there are presented and discussed as follows; 1.The status of operated radioactive liquid waste. 2. The radioactive liquid waste during dismantle the reactor. 3. Radiological status of radioactive liquid waste. 4. The classification criteria and method radioactive liquid waste. 6. The collection and transportation of radioactive liquid waste. (Author). 13 refs., 13 tabs., 8 figs

  9. Monte Carlo design for a new neutron collimator at the ENEA Casaccia TRIGA reactor.

    Burgio, N; Rosa, R

    2004-10-01

    The TRIGA RC-1 1MW reactor operating at ENEA Casaccia Center is currently being developed as a second neutron imaging facility that shall be devoted to computed tomography as well as neutron tomography. In order to reduce the gamma-ray content in the neutron beam, the reactor tangential piercing channel was selected. A set of Monte Carlo simulation was used to design the neutron collimator, to determine the preliminary choice of the materials to be employed in the collimator design. PMID:15246415

  10. Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code

    2015-01-01

    Cases of core coolant flow blockage transient have been simulated and analysed for the TRIGA IPR-R1 research reactor using the RELAP5-MOD3.3 code. The transients are related to partial and to total obstruction of the core coolant channels. The reactor behaviour after the loss of flow was analysed as well as the changes in the coolant and fuel temperatures. The behaviour of the thermal hydraulic parameters from the transient simulations was analysed. For a partial blockage, it was observed tha...

  11. Transient behavior studies of TRIGA core for variations in reactor kinetics

    Highlights: • Fast reactivity insertion analysis was done for TRIGA reactor. • Clad temperature remains below the design limit for 2$ reactivity at full power. • Peak power seems taken higher values for low initial power level. • Reactor parameters are more sensitive to variation of βeff than lp. • Inherent safe properties of TRIGA plays vital role in reactor safety. - Abstract: This paper illustrates the transient characteristics of TRIGA core at different kinetics conditions that arise from variations of externally inserted reactivity together with variations in other kinetic parameters such as prompt neutron life time, lp and effective delayed neutron fraction, βeff provided the reactor scram system is disabled. From the concern of severity of fast reactivity accident which could lead to most dangerous consequences, the inserted reactivity considered herein was fast reactivity which was a step type within the range 1$–2$ with insertion time 0.5 s. The initial power was 100 W and full power, 3 MW whereby the values of lp and βeff had been kept fixed at their recommended values 30 μs and 0.007, respectively. The observed parameters were reactor peak power and maximum clad temperature of the hottest channel. The analysis infers that clad temperature remains within its design limit value even for the maximum inserted reactivity, 2$ at full power operation of reactor. Also, the peak power took relatively higher values for transients at low power level. For each inserted reactivity, values of lp and βeff were varied individually within certain ranges. In this case, although both power and clad temperature are more sensitive to the variation in effective delayed neutron fraction than to the variation in prompt neutron life time, however, clad temperature remained within its design limit even for the maximum value of inserted reactivity with minimum βeff value considered. Prompt negative temperature coefficient of reactivity that stems from U-ZrH fuel

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

    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. PMID:25958412

  13. Neutronic Analysis of the 3 MW TRIGA MARK II Research Reactor, Part I: Monte Carlo Simulation

    This study deals with the neutronic analysis of the current core configuration of a 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh and validation of the results by benchmarking with the experimental, operational and available Final Safety Analysis Report (FSAR) values. The three-dimensional continuous-energy Monte Carlo code MCNP4C was used to develop a versatile and accurate full-core model of the TRIGA core. The model represents in detail all components of the core with literally no physical approximation. All fresh fuel and control elements as well as the vicinity of the core were precisely described. Continuous energy cross-section data from ENDF/B-VI and S(α, β) scattering functions from the ENDF/B-V library were used. The validation of the model against benchmark experimental results is presented. The MCNP predictions and the experimentally determined values are found to be in very good agreement, which indicates that the Monte Carlo model is correctly simulating the TRIGA reactor. (author)

  14. Transition from HEU to LEU fuel in Romania`s 14-MW TRIGA reactor

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

    1991-12-31

    The 14-MW TRIGA steady state reactor (SSR) located in Pitesti, Romania, first went critical in the fall of 1979. Initially, the core configuration for full power operation used 29 fuel clusters each containing a 5 {times} 5 square array of HEU (10 wt%) -- ZrH -- Er (2.8 wt%) fuel-moderator rods (1.295 cm o.d.) clad in Incology. With a total inventory of 35 HEU fuel clusters, burnup considerations required a gradual expansion of the core from 29 to 32 and finally to 35 clusters before the reactor was shut down because of insufficient excess reactivity. At this time each of the original 29 fuel clusters had an overage {sup 235}U burnup in the range from 50 to 62%. Because of the US policy regarding the export of highly enriched uranium, fresh HEU TRIGA replacement fuel is not available. After a number of safety-related measurements, the SSR is expected to resume full power operation in the near future using a mixed core containing five LEU TRIGA clusters of the same geometry as the original fuel but with fuel-moderator rods containing 45 wt% U (19.7% {sup 235}U enrichment) and 1.1 wt% Er. Rods for 14 additional LEU fuel clusters will be fabricated by General Atomics. In support of the SSR mixed core operation numerous neutronic calculations have been performed. This paper presents some of the results of those calculations.

  15. Transition from HEU to LEU fuel in Romania's 14-MW TRIGA reactor

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

    1991-01-01

    The 14-MW TRIGA steady state reactor (SSR) located in Pitesti, Romania, first went critical in the fall of 1979. Initially, the core configuration for full power operation used 29 fuel clusters each containing a 5 {times} 5 square array of HEU (10 wt%) -- ZrH -- Er (2.8 wt%) fuel-moderator rods (1.295 cm o.d.) clad in Incology. With a total inventory of 35 HEU fuel clusters, burnup considerations required a gradual expansion of the core from 29 to 32 and finally to 35 clusters before the reactor was shut down because of insufficient excess reactivity. At this time each of the original 29 fuel clusters had an overage {sup 235}U burnup in the range from 50 to 62%. Because of the US policy regarding the export of highly enriched uranium, fresh HEU TRIGA replacement fuel is not available. After a number of safety-related measurements, the SSR is expected to resume full power operation in the near future using a mixed core containing five LEU TRIGA clusters of the same geometry as the original fuel but with fuel-moderator rods containing 45 wt% U (19.7% {sup 235}U enrichment) and 1.1 wt% Er. Rods for 14 additional LEU fuel clusters will be fabricated by General Atomics. In support of the SSR mixed core operation numerous neutronic calculations have been performed. This paper presents some of the results of those calculations.

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

    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. (authors)

  17. The TRIGA reactor Frankfurt FRF 2, construction and present status

    The reactor FR2 was planned to replace the old FR1 reactor at the Frankfurt University, which was shutdown due to a failure. The FR2 is a tank reactor; the core is divided into 2 parts. The 1 MW operational core was calculated to consist of about 70 fuel elements. The grid plates contain 110 fuel element positions, surplus positions are occupied by dummy elements filled with water. The core has 5 control rods, 4 of which have fuel element followers. The reactor plant including all protection facilities was completed in 1977. The fuel elements were delivered in 1978. The reactor has been ready for critical experiments, but the permission for the start-up was not granted. In 1980 was decided that the reactor was not going in operation. Preparations are being made to dismantle the reactor components and to re-use them elsewhere

  18. Analysis of safety limits of the Moroccan TRIGA MARK II research reactor

    Erradi, L.; Essadki, H.

    2001-06-01

    The main objective of this study is to check the ability of the Moroccan TRIGA MARK II research reactor, designed to use natural convection cooling, to operate at its nominal power (2 MW) with sufficient safety margins. The neutronic analysis of the core has been performed using Leopard and Mcrac codes and the parameters of interest were the power distributions, the power peaking factors and the core excess reactivity. The thermal hydraulic analysis of the TRIGA core was performed using the French code FLICA designed for transient and study state situations. The main safety related parameters of the core have been evaluated with special emphasises on the following: maximum fuel temperature, minimum DNBR and maximum void fraction. The obtained results confirm the designer predictions except for the void fraction.

  19. The fuel element situation at the TRIGA mark II reactor Vienna

    The fuel history, spent fuel storage situation and recent problems covering the period from 1962 until 1.6.2001 were reviewed. After almost 40 years of TRIGA MARK II reactor Vienna operation, it must be mentioned that the experience with TRIGA fuel elements was and is excellent. During this period only 9 fuel elements had to be permanently be removed from the core and 57 fuel elements from the initial start-up are still used in the core. A careful fuel management and a frequent fuel inspection is of most importance, fuel elements should be moved at least two-times a year from their core position to check free movement and a 180 deg. rotation of the fuel element is also recommended (nevyjel)

  20. Analysis of safety limits of the Moroccan TRIGA MARK II research reactor

    The main objective of this study is to check the ability of the Moroccan TRIGA MARK II research reactor, designed to use natural convection cooling, to operate at its nominal power (2 MW) with sufficient safety margins. The neutronic analysis of the core has been performed using Leopard and Mcrac codes and the parameters of interest were the power distributions, the power peaking factors and the core excess reactivity. The thermal hydraulic analysis of the TRIGA core was performed using the French code FLICA designed for transient and study state situations. The main safety related parameters of the core have been evaluated with special emphasises on the following: maximum fuel temperature, minimum DNBR and maximum void fraction. The obtained results confirm the designer predictions except for the void fraction.

  1. Performance of the solid deuterium ultra-cold neutron source at the pulsed reactor TRIGA Mainz

    Karch, J.; Sobolev, Yu.; Beck, M.; Eberhardt, K.; Hampel, G.; Heil, W.; Kieser, R.; Reich, T.; Trautmann, N.; Ziegner, M.

    2014-04-01

    The performance of the solid deuterium ultra-cold neutron (UCN) source at the pulsed reactor TRIGA Mainz with a maximum peak energy of 10MJ is described. The solid deuterium converter with a volume of cm3 (8mol), which is exposed to a thermal neutron fluence of n/cm2, delivers up to 240000 UCN ( m/s) per pulse outside the biological shield at the experimental area. UCN densities of 10 cm3 are obtained in stainless-steel bottles of 10 L. The measured UCN yields compare well with the predictions from a Monte Carlo simulation developed to model the source and to optimize its performance for the upcoming upgrade of the TRIGA Mainz into a user facility for UCN physics.

  2. Performance of the solid deuterium ultra-cold neutron source at the pulsed reactor TRIGA Mainz

    The performance of the solid deuterium ultra-cold neutron (UCN) source at the pulsed reactor TRIGA Mainz with a maximum peak energy of 10MJ is described. The solid deuterium converter with a volume of V=160 cm3 (8mol), which is exposed to a thermal neutron fluence of 4.5 x 1013 n/cm2, delivers up to 240000 UCN (v ≤ 6 m/s) per pulse outside the biological shield at the experimental area. UCN densities of ∼ 10 cm3 are obtained in stainless-steel bottles of V ∼ 10 L. The measured UCN yields compare well with the predictions from a Monte Carlo simulation developed to model the source and to optimize its performance for the upcoming upgrade of the TRIGA Mainz into a user facility for UCN physics. (orig.)

  3. Performance of the solid deuterium ultra-cold neutron source at the pulsed reactor TRIGA Mainz

    Karch, J; Beck, M; Eberhardt, K; Hampel, G; Heil, W; Kieser, R; Reich, T; Trautmann, N; Ziegner, M

    2013-01-01

    The performance of the solid deuterium ultra-cold neutron source at the pulsed reactor TRIGA Mainz with a maximum peak energy of 10 MJ is described. The solid deuterium converter with a volume of V=160 cm3 (8 mol), which is exposed to a thermal neutron fluence of 4.5x10^13 n/cm2, delivers up to 550 000 UCN per pulse outside of the biological shield at the experimental area. UCN densities of ~ 10/cm3 are obtained in stainless steel bottles of V ~ 10 L resulting in a storage efficiency of ~20%. The measured UCN yields compare well with the predictions from a Monte Carlo simulation developed to model the source and to optimize its performance for the upcoming upgrade of the TRIGA Mainz into a user facility for UCN physics.

  4. Basic research using the 250 KW research reactor triga in Ljubljana, Yugoslavia

    The 25 KW Triga Mark II reactor of J. 'Stefan Institute' was commissioned on May 1966. During the last two years, it has been operated for about 4200 hr/year. According to experience gained with the reactor, most of the cost of reactor operation will be earned through isotope production for local hospitals and industries, performing low cost applied experiments and organizing training courses. The rest was provided through the Research Communities of the Republic of Slovenia. The reactor has been operated for 15 years without major problems and many basic research programmes have been performed. The research is being conducted in the following mainfields: solid state physics, neutron dosimetry, neutron radiography and autoradiography, reactor physics, examination of nuclear fuel using gamma scanning, irradiation of semiconducting materials and neutron activation analysis. (A.J)

  5. Corrosion of aluminium alloy test coupons in the TRIGA Mark III Research Reactor of Mexico

    The results of corrosion studies developed in the Instituto Nacional de Investigaciones Nucleares (ININ) are presented. The extent of corrosion of the aluminium alloy coupons exposed to the water of ININ TRIGA reactor pool was not significant. Few pits and oxides were observed on the coupon surfaces immersed for different times. This reduced extent of corrosion was similar to those on coupons exposed at other sites as per data obtained by visual inspection, metallographic analysis and image analysis. The water chemistry in the reactor pool was monitored throughout the duration of the project. The main parameters that influence the corrosion of Al alloy fuel cladding were measured. The conductivity of the water in the reactor pool was 1-3 μS/cm, within recommended values to avoid corrosion. The chloride ion concentration was maintained below 1 ppm. Others ions (sulphates, calcium, nitrates) were also below 1 ppm. Another parameter that was measured was the amount of settled solids on coupon surfaces and their influence on corrosion. The sedimentation rate in the TRIGA Reactor pool was 17.66 μg/cm2 and the sediment composition indicated iron oxides, aluminium-silicon compounds and some calcium carbonates. The sedimentation rate was similar in magnitude to that at other storage sites. However, the corrosion racks in the ININ TRIGA Reactor were exposed to high water flow rates, 1324.5 l/min. This high flow rate is considered to reduce the amount of deposited solids on coupon surfaces. The particles deposited on the coupon surfaces influenced pit initiation. (author)

  6. Predictive maintenance and its use in TRIGA-Pitesti reactor facilities

    Preda, M.; Barbalata, E.; Sabau, C. (Institute for Nuclear Research, Pitesti (Romania))

    1999-12-15

    The Pitesti TRIGA reactor is a research and material testing reactor situated on the bottom of an open pool of 300m3, whose steady state nominal power is 14 MW. It is cooled by a primary cooling system which comprises: 4 pumps (2 in operation, 2 in standby) and 3 heat exchangers. The generated heat in the reactor core is removed by a secondary circuit with forced convection towers (provided with 6 ventilators). The reactor was used for complete CANDU fuel testing, structural material (steel, zircaloy) testing and isotope production. The TRIGA Material Test and Research reactor was commissioned at the beginning of 1980. Since that there were made extensive tests on CANDU type fuel and structural materials. It is needed the increase the reliability of equipment's and demanded an improved performance of our facilities. Good maintenance is seen as one of the main keys to improve the performance of TRIGA reactors. For a better operation we are obliged to find and use each up to date methods and strategies. Among these new techniques we could quote the probabilistic assessments, and some of predictive maintenance's techniques. Probabilistic safety and statistical analysis provided useful insights for our reactor operation. During the reactor operation there were unexpected shutdowns, reactor components failures. The data collected were statistically processed in order to obtain a reliability data base. This paper does, indifferently the cause, analysis the failures. The study emphasizes that the most reactor's scrams took place on the first year of work. The scrams number began to lower thereafter and at the end of eighties began to increase again. The greatest number of scrams were caused by the reactor electrical control and instrumentation. An important number of scrams were caused by the irradiation devices. The main conclusion of this study is that the insights are very useful to our operational procedures, to improve the maintenance strategy and the

  7. TRIGA Research Reactor Conversion to LEU and Modernization of Safety Related Systems

    The USA and IAEA proposed an international programme to reduce the enrichment of uranium in research reactors by converting nuclear fuel containing HEU into fuel containing 20% enriched uranium. The Government of Romania joined the programme and actively supported political, scientific, technical and economic actions that led to the conversion of the active area of the 14 MW TRIGA reactor at the Institute for Nuclear Research in Piteşti in May 2006. This confirmed the continuity of the Romanian Government’s non-proliferation policy and their active support of international cooperation. Conversion of the Piteşti research reactor was made possible by completion of milestones in the Research Agreement for Reactor Conversion, a contract signed with the US Department of Energy and Argonne National Laboratory. This agreement provided scientific and technical support and the possibility of delivery of all HEU TRIGA fuel to the United States. Additionally, about 65% of the fresh LEU fuel needed to start the conversion was delivered in the period 1992–1994. Furthermore, conversion was promoted through IAEA Technical Cooperation project ROM/4/024 project funded primarily by the United States that supported technical and scientific efforts and the delivery of the remaining required LEU nuclear fuel to complete the conversion. Nuclear fuel to complete the conversion was made by the French company CERCA with a tripartite contract among the IAEA, CERCA and Romania. The contract was funded by the US Department of Energy with a voluntary contribution by the Romanian Government. The contract stipulated manufacturing and delivery of LEU fuel by CERCA with compliance measures for quality, delivery schedule and safety requirements set by IAEA standards and Romanian legislation. The project was supported by the ongoing technical cooperation, safeguards, legal and procurement assistance of the IAEA, in particular its Department of Nuclear Safety. For Romanian research, the

  8. Criticality calculation in TRIGA MARK II PUSPATI Reactor using Monte Carlo code

    A Monte Carlo simulation of the Malaysian nuclear reactor has been performed using MCNP Version 5 code. The purpose of the work is the determination of the multiplication factor (keff) for the TRIGA Mark II research reactor in Malaysia based on Monte Carlo method. This work has been performed to calculate the value of keff for two cases, which are the control rod either fully withdrawn or fully inserted to construct a complete model of the TRIGA Mark II PUSPATI Reactor (RTP). The RTP core was modeled as close as possible to the real core and the results of keff from MCNP5 were obtained when the control fuel rods were fully inserted, the keff value indicates the RTP reactor was in the subcritical condition with a value of 0.98370±0.00054. When the control fuel rods were fully withdrawn the value of keff value indicates the RTP reactor is in the supercritical condition, that is 1.10773±0.00083. (Author)

  9. Initial core calculation of 1 MW reactor TRIGA PUSPATI (RTP) using SRAC code system

    The 1 MWatt TRIGA PUSPATI Reactor (RTP) was located in Malaysian Institute for Nuclear Technology Research (MINT). This research reactor was from TRIGA MARK II type and was put into operation on 1983 and has reached its first criticality on 28 June 1982. Since then, this reactor has been used for various beam experiments, irradiation facilities, radioisotope production and education and training. The RTP uses three types of fuel elements, namely, 8.5wt%, 12wt% and 20wt% which enriched to about 20% of U-235 for all types. The RTP has four control rods which made up of boron carbide. It has cylindrical core but not in periodically in its lattice structure, which possibly locates 127 of fuel elements. Both of the coolant and moderator uses light water system and the reflector was made from high purity graphite. Because of this research reactor's power is relatively small compared to the power reactor; it uses natural convection for its cooling system. To ensure the integrity of the core, fuel shuffling have been made for several times. Until now, there are 11 configurations of the core and recently has achieved the 12th configuration. This paper will described the first core configuration calculation using SRAC code system which was first introduced in 2005 during the FNCA workshop. (author)

  10. Neutronics analysis of the proposed 25-MW LEU TRIGA Multipurpose Research Reactor

    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-NW LEU TRIGA Multipurpose Research Reactor (MPRR) proposed by the General Atomic Company of the United States of America. In all aspects except for the shutdown margin, the 25-MW LEU TRIGA Multipurpose Research Reactor performs very well. The high uranium density of the U-ZrH-Er fuel with its burnable poison makes possible a long equilibrium cycle length with a relatively small reactivity swing. Therefore, control rod movement is minimized during the cycle, leading to a stable flux. The lack of adequate shutdown margin can probably be remedied by the use of a higher-worth design of the control rods

  11. Development of neutron beam projects at the University of Texas TRIGA Mark II Reactor

    Recently, the UT-TRIGA research reactor was licensed and has become fully operational. This reactor, the first new US university reactor in 17 years, is the focus of a new reactor laboratory facility which is located on the Balcones Research Center at The University of Texas at Austin. The TRIGA Mark II reactor is licensed for 1.1 MW steady power operation, 3 dollar pulsing, and includes five beam ports. Various neutron beam-line projects have been assigned to each beam port. Neutron Depth Profiling (NDP) and the Texas Cold Neutron Source (TCNS) are close to completion and will be operational in the near future. The design of the NDP instrument has been completed, a target chamber has been built, and the thermal neutron collimator, detectors, data acquisition electronics, and data processing computers have been acquired. The target chamber accommodates wafers up to 12'' in diameter and provides remote positioning of these wafers. The design and construction of the TCNS has been completed. The TCNS consists of a moderator (mesitylene), a neon heat pipe, a cryogenic refrigerator, and neutron guide tubes. In addition, fission-fragment research (HIAWATHA), Neutron Capture Therapy, and Neutron Radiography are being pursued as projects for the other three beam ports. (author)

  12. Evaluation of thermal-hydraulic parameter uncertainties in a TRIGA research reactor

    Mesquita, Amir Z.; Costa, Antonio C.L.; Ladeira, Luiz C.D.; Rezende, Hugo C., E-mail: amir@cdtn.br, E-mail: aclc@cdtn.br, E-mail: lcdl@cdtn.br, E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    Experimental studies had been performed in the TRIGA Research Nuclear Reactor of CDTN/CNEN to find out the its thermal hydraulic parameters. Fuel to coolant heat transfer patterns must be evaluated as function of the reactor power in order to assess the thermal hydraulic performance of the core. The heat generated by nuclear fission in the reactor core is transferred from fuel elements to the cooling system through the fuel-cladding (gap) and the cladding to coolant interfaces. As the reactor core power increases the heat transfer regime from the fuel cladding to the coolant changes from single-phase natural convection to subcooled nucleate boiling. This paper presents the uncertainty analysis in the results of the thermal hydraulics experiments performed. The methodology used to evaluate the propagation of uncertainty in the results was done based on the pioneering article of Kline and McClintock, with the propagation of uncertainties based on the specification of uncertainties in various primary measurements. The uncertainty analysis on thermal hydraulics parameters of the CDTN TRIGA fuel element is determined, basically, by the uncertainty of the reactor's thermal power. (author)

  13. The evaluation of research reactor TRIGA MARK II safety

    In the paper the Probabilistic Safety Analysis (PSA) of a research reactor is described. Five different initiating events were selected and analyzed with the use of event trees. Seven reactor systems were modeled with fault trees. Three groups of radiation releases were introduced - Success, Reactor-Hall, Environment - and their frequencies were estimated. The importance factors of initiating events, human errors and basic events were calculated regarding the consequence groups. (author)

  14. Uses of TRIGA reactors in university and industrial research and development including prooftesting of thermionic generators and inspection of aircraft using neutron radiography

    TRIGA reactors owe their great flexibility to the special features of the uranium-zirconium hydride fuel matrix. The resulting inherent safety of this reactor system has encouraged the application to a very broad spectrum of experiments, tests, and development activities. Examples from fuel testing programs, new neutron radiography applications, and power TRIGA projects have been selected to demonstrate the diversity of applications. (author)

  15. Ten years of TRIGA reactor research at the University of Texas

    The 1 MW TRIGA Research Reactor at the Nuclear Engineering Teaching Laboratory is the second TRIGA at the University of Texas at Austin (UT). A small (10 kW-1963, 250 kW-1968) TRIGA Mark I was housed in the basement of the Engineering Building until is was shutdown and decommissioned in 1989. The new TRIGA Mark II with a licensed power of 1.1 MW reached initial criticality in 1992. Prior to 1990, reactor research at UT usually consisted of projects requiring neutron activation analysis (NAA) but the step up to a much larger reactor with neutron beam capability required additional personnel to build the neutron research program. The TCNS is currently used to perform Prompt Gamma Activation Analysis to determine hydrogen and boron concentrations of various composite materials. The early 1990s was a very active period for neutron beam projects at the NETL. In addition to the TCNS, a real-time neutron radiography facility (NIF) and a high-resolution neutron depth profiling facility (NDP) were installed in two separate beam ports. The NDP facility was most recently used to investigate alpha damage on stainless steel in support of the U.S. Nuclear Weapons Stewardship programs. In 1999, a sapphire beam filter was installed in the NDP system to reduce the fast neutron flux at the sample location. A collaborative effort was started in 1997 between UT-Austin and the University of Texas at Arlington to build a reactor-based, low-energy positron beam (TIPS). The limited success in obtaining funding has placed the project on hold. The Nuclear and Radiation Engineering Program has grown rapidly and effectively doubled in size over the past 5 years but years of low nuclear research funding, an overall stagnation in the U.S. nuclear power industry and a persuasive public distrust of nuclear energy has caused a precipitous decline in many programs. Recently, the U.S. DOE has encouraged University Research Reactors (URR) in the U.S. to collaborate closely together by forming URR

  16. American Nuclear Society standards for TRIGA reactors and their use

    The American Nuclear Society established a committee (ANS-15) with the expressed charter to develop standards for research reactors. These standards were to cover all aspects of research reactor operations, maintenance and administration. Standards have been written in every area of research reactor operations that the research reactor community has deemed important. One of the uppermost goals of the Standards Committee work is to produce standards that provide guidance and help to the research reactor community in a timely manner. To make the standards meaningful requires a great deal of cooperation between all segments of the reactor community. The research reactors - whether they are private, university or government owned - have a mission to perform. At the same time, the regulatory agencies also have a mission to perform, and with a spirit of mutual respect and cooperation, both can accomplish their goals. In the last five years this spirit has been present, and a number of very good standards have resulted. These standards should be a part of every research reactor library. In particular ANS-15.16 and ANS-15.1 have been endorsed by the regulatory agencies and are being used to evaluate submittals

  17. Results from Accelerator Driven TRIGA Reactor Experiments at The University of Texas at Austin

    Accelerator Driven Transmutation of High-Level Waste (ATW) is one possible solution to the fuel reprocessing back-end problem for the disposal of high level waste such as minor actinides (Am, Np or Cm) and long-lived fission products. International programs continue to support research towards the eventual construction and operation of a proton accelerator driven spallation neutron source coupled to a subcritical 'neutron amplifier' for more efficient HLW transmutation. This project was performed under DOE AFCI Reactor-Accelerator Coupling Experiments (RACE). A 20 MeV Electron Linac was installed in the BP no 5 cave placing neutron source adjacent to an offset reactor core to maximize neutron coupling with available systems. Asymmetric neutron injection 'wasted' neutrons due to high leakage but sufficient neutrons were available to raise reactor power to ∼100 watts. The Linac provided approximately 100 mA but only 50% reached target. The Linac cooling system could not prevent overheating at frequencies over 200 Hz. The Linac electron beam had harmonics of primary frequency and periodic low frequency pulse intensity changes. Neutron detection using fission chambers in current mode eliminated saturation dead time and produced better sensitivity. The Operation of 'dual shielded' fission chambers reduced electron noise from linac. Benchmark criticality calculation using start-up data showed that the MCNPX model overestimates reactivity. TRIGA core was loaded to just slightly supercritical by adding graphite elements and measuring reactivity of $0.037. MCNPX modeled TRIGA core with and without graphite to arrive at 'true' measured subcritical multiplication of 0.998733± 0.00069. Thus, Alpha for the UT-RACE TRIGA core was approximately 155.99 s-1. The Stochastic Feynman-Alpha Method (SFM) accuracy was evaluated during transients and reactivity changes. SFM was shown to be a potential real-time method of reactivity determination in future ADSS but requires stable

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

    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.

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

    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 (κeff) 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.

  20. Analysis Of Criticality Experiments Of Bandung Triga 2000 Reactor By Using MCNP-4B Code

    During the first core loading of Bandung TRIGA 2000 reactor, two kinds of criticality experiment have been conducted, i.e, sub critical core loading and critical core loading experiments. The purpose of the experiments is to maximize the utilization of the reactor as well as to provide benchmark data for neutronic computer codes. In the sub critical core loading experiment, the core is loaded up to 42 fuel elements ring D, 13 fuel elements in ring, D, 6 fuel elements and 3 graphite dummies in ring E, 2 fuel elements in ring B, 2 fuel elements in ring B, 1 fuel element in ring B. In the other case, during the critical loading experiment, the core is loaded following the loading pattern planned by General Atomics, i.e: 20 fuel elements in ring B, C and D plus 5 control rods in ring D, 11 fuel elements in ring D, 6 fuel elements and 3 graphite dummies in ring E, and then the core is loaded with additional fuel elements, step by step, until the core reached its first criticality, i.e., 55 fuel elements. Prior to conduct of criticality experiments MCNP-4B code is used to plan the fuel loading pattern of the sub critical loading experiment, i.e. to assure that the core is still in sub critical state with 42 fuel elements in the core. In the calculation is assumed that the mass of U-235 in each fuel element depends on the documented burnup data, the mass of U-238 is assumed to be the same as the one in fresh fuels. Furthermore, all fission patricides as well as poisonous materials in each fuel element are ignored. The experiment results showed that the calculations of MCNP-4B also predicted that TRIGA 2000 reactor with the above assumptions, is appropriate for predicting for predicting the neutronic characteristics of Bandung TRIGA 2000 reactor

  1. Validation of the neutron and gamma fields in the JSI TRIGA reactor using in-core fission and ionization chambers.

    Žerovnik, Gašper; Kaiba, Tanja; Radulović, Vladimir; Jazbec, Anže; Rupnik, Sebastjan; Barbot, Loïc; Fourmentel, Damien; Snoj, Luka

    2015-02-01

    CEA developed fission chambers and ionization chambers were utilized at the JSI TRIGA reactor to measure neutron and gamma fields. The measured axial fission rate distributions in the reactor core are generally in good agreement with the calculated values using the Monte Carlo model of the reactor thus verifying both the computational model and the fission chambers. In future, multiple absolutely calibrated fission chambers could be used for more accurate online reactor thermal power monitoring. PMID:25479432

  2. Development process of the new control console of ININ's TRIGA mark III reactor

    A description of the development of the new ININ's TRIGA Mark III reactor control console is presented in this meeting. Most of the operation and safety monitoring of the reactor is carried out by means of a personal computer (PC), some interface cards, and an auxiliary computer that drives the control rod mechanisms. In this console, the safety actions are taken by the Protection System (SEC), which acquires the data directly from the safety related systems, specified in the reactor's console design technical specifications. The console, based on the concept of virtual instrumentation, is composed of a group of systems that make easier to the operator the activation of the sequential steps required to operate the reactor. (authors)

  3. Experience with modernization and refurbishment of the Vienna TRIGA Mark II reactor I and C system

    The refurbishment of the instrumentation and control (I and C) system of a research reactor is a major task which needs careful planning and taking many aspects into account. At any early planning stage, the future of the facility has to be demonstrated to the national authorities by providing a detailed business plan and the cost of I and C replacement will be compared by financial authorities against the cost of decommissioning the facility. The TRIGA reactor Vienna was modernized in 1992 with a new digital instrumentation and control (I and C) system. The replacement procedure and the reactor-specific modifications to the standard reactor instrumentation offered by the supplier, the operation experience during the past 15 years and a compilation of benefits and other issues to be considered in these procedures (changing from analog to digital I and C system) are summarized in this report. (nevyjel)

  4. Uncertainty analysis on thermal hydraulics parameter of the IPR-R1 TRIGA research nuclear reactor

    Experimental studies have been performed in the IPR-R1 TRIGA Mark 1 Research Nuclear Reactor of CDTN/CNEN at Belo Horizonte (Brazil) to find out the temperature distribution as a function of reactor power, under steady-state conditions. During these experiments the reactor was set in many different power levels. These experiments are part of the research program, that have the main objective of commissioning the IPR-R1 reactor for routine operation at 250 k W. This paper presents the uncertainty analysis of the thermal-hydraulic experiments performed. The methodology used to evaluate the uncertainty propagation on the results was done based on the pioneering article of Kline and McClintock (1953), with the propagation of uncertainties based on the specification of uncertainties in various primary measurements. (author)

  5. Thermal hydraulic parameter studies of heat exchanger for the TRIGA MARK II research reactor

    Thermal Hydraulic studies have being conducted at PUSPATI TRIGA Mark II (RTP) Nuclear Research Reactor. The purpose of this study is to determine the heat transfer characteristic and heat exchanger performance at difference reactor power. Fundamental concept and a plate type application of heat exchanger in RTP are presented in this study. A plate type heat exchanger is a device for RTP reactor cooling system built for efficient heat transfer from one fluid to another. The study involves the observation of inlet and outlet temperature profile, flow rate and pressure at the reactor pool and heat exchanger. The observed parameters are compared to basic engineering calculation and the output of the study has been beneficial to evaluate the performance of newly-installed plate type heat exchanger. (author)

  6. Development process of the new control console of ININ's TRIGA mark III reactor

    Rivero-Gutierrez, T. [Inst. Nacional de Investigationes Nucleares ININ, Gerencia de Ciencias Aplicadas, Carretera Mexico-Toluca S/N, Estado de Mexico, C.P. 52750 (Mexico); Inst. Tecnologico de Toluca, Div. de Estudios de Postgrado e Investigacion, Av. Tecnologico S/N, Estado de Mexico, C.P. 52140 (Mexico); Sainz-Mejia, E. [Inst. Nacional de Investigationes Nucleares ININ, Gerencia de Ciencias Aplicadas, Carretera Mexico-Toluca S/N, Estado de Mexico, C.P. 52750 (Mexico); Benitez-Read, J. S. [Inst. Nacional de Investigationes Nucleares ININ, Gerencia de Ciencias Aplicadas, Carretera Mexico-Toluca S/N, Estado de Mexico, C.P. 52750 (Mexico); Inst. Tecnologico de Toluca, Div. de Estudios de Postgrado e Investigacion, Av. Tecnologico S/N, Estado de Mexico, C.P. 52140 (Mexico); Marroquin, J. L. G. [Inst. Nacional de Investigationes Nucleares ININ, Gerencia de Ciencias Aplicadas, Carretera Mexico-Toluca S/N, Estado de Mexico, C.P. 52750 (Mexico)

    2006-07-01

    A description of the development of the new ININ's TRIGA Mark III reactor control console is presented in this meeting. Most of the operation and safety monitoring of the reactor is carried out by means of a personal computer (PC), some interface cards, and an auxiliary computer that drives the control rod mechanisms. In this console, the safety actions are taken by the Protection System (SEC), which acquires the data directly from the safety related systems, specified in the reactor's console design technical specifications. The console, based on the concept of virtual instrumentation, is composed of a group of systems that make easier to the operator the activation of the sequential steps required to operate the reactor. (authors)

  7. Initial operation and utilization of the Bangladesh 3 Mw TRIGA reactor

    A 3 Mw TRIGA MK-II pulsing type research reactor fuelled with low enrichment uranium having 19.7% U-235 and 20 wt % Uranium, 0.47% Erbium and Zirconium Hydride, has been installed at the Atomic Energy Research Establishment, savar in the last week of October, 1986. This multi-purpose reactor, capable of both steady-state and pulsing operation, has been put into service in several disciplines since its commissioning and presently in operation without any major problem. The paper describes the initial operating experience and the reactor utilization made in several areas including the operator training conducted for the formation of the initial crew for the reactor. (author)

  8. Simulation of a TRIGA Reactor Core Blockage Using RELAP5 Code

    Patrícia A. L. Reis

    2015-01-01

    Full Text Available Cases of core coolant flow blockage transient have been simulated and analysed for the TRIGA IPR-R1 research reactor using the RELAP5-MOD3.3 code. The transients are related to partial and to total obstruction of the core coolant channels. The reactor behaviour after the loss of flow was analysed as well as the changes in the coolant and fuel temperatures. The behaviour of the thermal hydraulic parameters from the transient simulations was analysed. For a partial blockage, it was observed that the reactor reaches a new steady state operation with new values for the thermal hydraulic parameters. The total core blockage brings the reactor to an abnormal operation causing increase in core temperature.

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

    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.

  10. Fuel element reshuffling and fuel follower control rods (FFCR) replacement for PUSPATI TRIGA reactor

    The PUSPATI TRIGA Reactor has been utilized for more than 25 years using the same fuel elements and control rods. Generally, there are four control rods being used to control the neutron production inside the reactor core. A maintenance program has been developed to ensure its integrity, capability and safety of the reactor and it has been maintained twice a year since the first operation in 1982. The activities involve during the maintenance period including fuel elements and control rods inspections, electronics and mechanical systems, and others related works. During the maintenance in August 2008, there are some irregularities found on the fuel follower control rods and needed to be replaced. Even though the irregularities was not contributed into any unwanted incident, it were decided to replace with new control rods to avoid any potential hazards and unsafe condition occurred during operation later. Replacing any of the control rods would involved in imbalance of neutron flux and power distribution inside the core. Therefore, a number of fuel elements need to be reshuffled in order to compensate the neutron flux and power distribution as well as to balance the fuel elements burn-up in the core. This paper will described the fuel elements reshuffling and fuel follower control rods (FFCR) replacement for PUSPATI TRIGA Reactor. (Author)

  11. Destructive Examination of Experimental Candu Fuel Elements Irradiated in TRIGA-SSR Reactor

    The object of this work is the behaviour of CANDU fuel elements under power cycling conditions. The tests were run in the 14 MW(th) TRIGA-SSR (Steady State Reactor) reactor from Institute for Nuclear Research (INR) Pitesti. zircaloy-4 is the material used for CANDU fuel sheath. The importance of studying its behaviour results from the fact that the mechanical properties of the CANDU fuel sheath suffer modifications during normal and abnormal operation. In the nuclear reactor the fuel elements endure dimensional and structural changes as well as cladding oxidation, hydriding and corrosion. These changes can lead to defects and even to the loss of integrity of the cladding. This paper presents the results of examinations performed in the Post- irradiation Examination Laboratory (PIEL) from INR Pitesti, on samples from a fuel element irradiated in TRIGA-SSR reactor: (i) Dimensional and macrostructural characterization; (ii) Microstructural characterization by metallographic analyses; (iii) Determination of mechanical properties; (iv) Fracture surface analysis by scanning electron microscopy (SEM). The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for CANDU fuel improvement. (author)

  12. Post Irradiation Examination of Experomental CANDU Fuel Elements Irradiated in TRIGA-SSR Reactor

    The object of this work is the behaviour of CANDU fuel elements under power cycling conditions. The tests were run in the 14 MW (th) TRIGA-SSR (Steady State Reactor) reactor from Institute for Nuclear Research (INR) Pitesti. Zircaloy-4 is the material used for CANDU fuel sheath. The importance of studying its behaviour results from the fact that the mechanical properties of the CANDU fuel sheath suffer modifications during normal and abnormal operation. In the nuclear reactor the fuel elements endure dimensional and structural changes as well as cladding oxidation, hydriding and corrosion. These changes can lead to defects and even to the loss of integrity of the cladding. This paper presents the results of examinations performed in the Post Irradiation Examination Laboratory (PIEL) from INR Pitesti, on samples from a fuel element irradiated in TRIGA-SSR reactor: (i) Dimensional and macrostructural characterization; (ii) Gamma scanning and tomography; (iii) Measurement of pressure, volume and isotopic composition of fission gas; (iv) Microstructural characterization by metallographic analyses; (v) Determination of mechanical properties; amd (vi) Fracture surface analysis by scanning electron microscopy (SEM). The obtained data could be used to evaluate the security, reliability and nuclear fuel performance, and for CANDU fuel improvement. (author)

  13. Load following tests on CANDU-type fuel elements in TRIGA research reactor of INR Pitesti

    Two load following (LF) tests on CANDU-type fuel elements were performed in the TRIGA Research Reactor of INR Pitesti, where the tests were designed to represent fuel in a CANDU reactor operating in a load following regime. In the first LF test the designated '78R' fuel element successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, developed under INR-AECL co-operation, the fuel element designated as 'ME01' withstood 200 power cycles from 27 to 54 kW/m average linear power, as well as additional ramps due to reactor trips and restarts during the test period. This experimental program is ongoing at INR Pitesti. Both LF tests were simulated with finite element computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the cladding arising from expansion and contraction of the pellets. New LF tests are planned to be performed in order to establish the limits and capabilities for CANDU fuel in LF conditions. This paper presents the results of the LF tests performed in the INR TRIGA Research Reactor compared with the analytical assessment for SCF conditions and their relation to CANDU fuel performance in LF conditions. (author)

  14. An Experimental Study of Natural Convection in The Hottest Channel of TRIGA 2000 k W Reactor

    With the increase of radioisotope demand, in 1995, National Nuclear Energy Agency of Indonesia made a decision to upgrade the power of the TRIGA Mark II reactor from 1 MW to 2 MW maximum power. The reactor reached its first criticality on May 13, 2000. To accomplish the safety evaluation of the reactor, a thermal hydraulic analysis was carried out by using thermal hydraulic computer code. This code calculates the natural convection flow through water coolant bounded by vertical cylindrical heat sources. In this paper, it will be reported the experimental study of natural convection in the hottest channel of TRIGA 2000 k W reactor. The purpose of the experimental study is to verify the theoretical analysis, especially the temperature distribution in the hottest coolant channel. In this experiment, a special probe for temperature detection has been designed and inserted to central thimble (CT). In the experiment, eight thermocouples were used to measure the bulk temperature of the water at different position in the cooling channel and simultaneous quantitative measurement of the temperature distribution were done by using a data acquisition cards system. The result obtained theoretically using the STAT code has been verified by this experimental study. (author)

  15. Load following tests on CANDU-type fuel elements in TRIGA research reactor of INR Pitesti

    Horhoianu, G. [Inst. for Nuclear Research (INR), Pitesti (Romania); Palleck, S. [Atomic Energy of Canada Limited., Mississauga, Ontario (Canada); Ionescu, D. [Inst. for Nuclear Research (INR), Pitesti (Romania)

    2010-07-01

    Two load following (LF) tests on CANDU-type fuel elements were performed in the TRIGA Research Reactor of INR Pitesti, where the tests were designed to represent fuel in a CANDU reactor operating in a load following regime. In the first LF test the designated '78R' fuel element successfully experienced 367 power cycles, mostly between 23 and 56 kW/m average linear power. In the second LF test, developed under INR-AECL co-operation, the fuel element designated as 'ME01' withstood 200 power cycles from 27 to 54 kW/m average linear power, as well as additional ramps due to reactor trips and restarts during the test period. This experimental program is ongoing at INR Pitesti. Both LF tests were simulated with finite element computer codes in order to evaluate Stress Corrosion Fatigue (SCF) of the cladding arising from expansion and contraction of the pellets. New LF tests are planned to be performed in order to establish the limits and capabilities for CANDU fuel in LF conditions. This paper presents the results of the LF tests performed in the INR TRIGA Research Reactor compared with the analytical assessment for SCF conditions and their relation to CANDU fuel performance in LF conditions. (author)

  16. The TRIGA reactor Frankfurt construction and experimental facilities

    The new reactor FRF 2 was designed by Gutehoffnungshutte Sterkrade AG in cooperation with the reactor group of the Institut fur Kernphysik. The maximum power level is 1 MW; later installation of facilities for pulsed operation is possible. Performance and design data of the FRF 2 are given. The reactor is expected to start operation in 1973. Since the FRF 2 will be installed inside the biological shield and reflector of the FRF 1, the FRF 2 core has to correspond to the FRF 1 core structure

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

    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

  18. McClellan Nuclear Radiation Center (MNRC) TRIGA reactor: Four years of operations

    McClellan Air Force Base, at Sacramento, California, is headquarters for the Sacramento Air Force Logistics Center (SM-ALC). McClellan Air Force Base provides extensive inspection and maintenance capabilities for the F-111, F-1 5, and other military aircraft. Criticality of the MNRC TRIGA reactor was obtained on January 20, 1990 with 63 standard TRIGA fuel elements, three fuel-followed control rods and one air-followed control rod. Presently there are 93 fuel elements in the reactor core. The reactor can be operated at 1 MW steady state power, producing pulses up to three dollars worth of reactivity addition, and can be square waved up to 1 MW. The reactor core contains a circular grid plate and a graphite reflector assembly surrounding the core. Four tangential beam ports installed in the reflector assembly provide a thermal neutron flux to four radiography bays. The reactor tank is twenty-four (24) feet deep, seven and one-half (7.5) feet in diameter, and has a protrusion in the upper portion of the reactor tank. This protrusion is scheduled for use as a neutron thermal collimator in the future. Besides the neutron radiography capabilities, the reactor contains a pneumatic rabbit system, a central thimble, an in-core irradiation facility, and three additional cutouts that provide locations for additional irradiation facilities. The central thimble can be removed along with the B-ring locations of the upper portion of the grid plate to provide an additional and larger in-core irradiation facility. A new upper grid plate has been manufactured to expand one triangular cutout so that larger experiments can be inserted directly into the reactor core. Some operational problems experienced during the first four years of operations are the timeout of the CSC and DAC watchdogs, deterioration of the heat exchanger gaskets, and loss of thermocouples in the instrumented fuel elements. (author)

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

    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)

  20. Spent Fuel Management Program in the 3MW TRIGA MARK-II Research Reactor of Bangladesh

    Bangladesh Atomic Energy Commission (BAEC) has been operating a 3 MW TRIGA MARK II research reactor since 1986. The reactor was installed in the campus of the Atomic Energy Research Establishment (AERE) at Savar, Dhaka. It is one of the main nuclear research facilities in the country. The reactor uses TRIGA LEU fuel with uranium content of 20% by weight. The enrichment level of the fuel is 19.7%. The reactor has so far been operated for 7834 hours with a total cumulative burn up of 15898 MWh (662.5 MWd). The total burn up life of the present core is 1200 MWd. The main areas of use are: training of man-power for nuclear power plant applications, radioisotope (RI) production, neutron activation analysis (NAA), neutron radiography (NR) and neutron scattering. The government of Bangladesh has taken decision to establish nuclear power programme in the country. There is an ADP (Annual Development Project) to accomplish necessary activities for construction of medium size nuclear power plant (NPP) in the western zone of the country. Now, with regard to the safe management, storage of spent fuel and disposal of radioactive waste arising from operation of the research reactor and also from the proposed NPP expected to be constructed in future, BAEC is drawing up short and long-term plans and programs. At present, there does not exist any spent fuel element in the reactor facility. It is to be mentioned that Bangladesh is aware of the US DOE’s ‘Take Back Program’ in connection with the research reactor spent fuel of US origin, and is very much interested to take part in this program. The paper presents the current status of handling and storage facilities available for spent fuel and strategy for the safe management of spent fuel to be generated from the research reactor in near future. (author)

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

    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)

  2. Present and future activities of TRIGA RC-1 Reactor

    A summary of reactor activities is presented and discussed. The RC-1 reactor is used by ENEA's laboratories, research institutes and national industries for different aims: research, analysis materials behaviour under neutron flux, etc. To satisfy the requests increase it is important to signalize: - the realization of a new radiochemical laboratory for radioisotopes production, to be used in a medical and/or diagnostic field in general; - the realization of a tritium handling laboratory, to study tritium solubility, release and diffusion in different material (particularly in ceramic breeder as lithium aluminate) to support Italian programs on fusion technology; - a research activity on the reactors computerized control by a console of advanced conception. The aim of this activity is the development of an ergonomic control room that could be a reference point for the planning of the power reactor control rooms

  3. Study of a new automatic reactor power control for the TRIGA Mark II reactor at University of Pavia

    The installation of a new Instrumentation and Control (IC) system for the TRIGA Mark-II reactor at University of Pavia has recently been completed in order to assure a safe and continuous reactor operation for the future. The intervention involved nearly the whole IC system and required a channel-by-channel component substitution. One of the most sensitive part of the intervention concerned the Automatic Reactor Power Controller (ARPC) which permits to keep the reactor at an operator-selected power level acting on the control rod devoted to the fine regulation of system reactivity. This controller installed can be set up using different control logics: currently the system is working in relay mode. The main goal of the work presented in this paper is to set up a Proportional-Integral-Derivative (PID) configuration of the new controller installed on the TRIGA reactor of Pavia so as to optimize the response to system perturbations. The analysis have shown that a continuous PID offers generally better results than the relay mode which causes power oscillations with an amplitude of 3% of the nominal power

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

    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.)

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

    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

  6. Failure of triga fuel cladding at the Berkeley Research Reactor

    On September 16, 1985, following a long maintenance shutdown, unusually high concentrations of radioisotopes were detected in the reactor-room air on a Constant Air Monitor (CAM) after two and a half hours of full power operation. It was thought that the activity could be coming from some contamination in the pool water. Thus the water was cleaned and the water conductivity was reduced fourfold. However, a full-power operation again showed high count rates on the CAM. A third test was conducted with a germanium detector. Following two hours of operation, three fission-product gasses were identified in the reactor-room air; Kr85, Kr37 and Kr88. Once again no unusual activities could be detected on the CAM filter, in the pool water, or in the demineralizer resins. It was concluded that the gasses must be coming from a leaking fuel element. Three old, instrumented elements with defective thermocouples were selected to be the first ones isolated from the core. After removing the elements, the reactor was operated at full-power for two hours with no abnormal activities detected. New standard elements were loaded and the reactor was again operated at full-power to confirm that no leaking element remained in the core. Since then, the reactor has been operated, with no abnormal activities detected. (Nogami, K.)

  7. In-situ gamma spectrometry measurements of time-dependent Xenon-135 inventory in the TRIGA Mark II reactor Vienna

    Riede, Julia; Boeck, Helmuth

    2013-01-01

    In this work, it has been shown that the time dependent Xe-135 inventory in the TRIGA Mark II reactor in Vienna, Austria can be measured via gamma spectrometry even in the presence of strong background radiation. It is focussing on the measurement of (but not limited to) the nuclide Xe-135. The time dependent Xe-135 inventory of the TRIGA Mark II reactor Vienna has been measured using a temporary beam line between one fuel element of the core placed onto the thermal column after shutdown and ...

  8. Experience of Ageing Management at 14 MW TRIGA Research Reactor from INR Pitesti, Romania

    The 14 MW TRIGA Research Reactor designed in the early 70s is a relative new research reactor with an operational experience of 30 years.The specific design of reactor core objectives, were to manufacture, build and operate a flexible structure which incorporate previous experience of pool type research reactors. Aluminum alloy 6061 and stainless steel are only materials used for core structural components, which are all easily remotely removable and replaceable by simple hand tools. Properties of those categories of materials were well characterized / known for many other reactors predecessors, and no special criteria or preliminary tests were performed. In spite of well known materials properties, the behavior uncertainties of those materials in each reactor case may have special aspects related to design of components, manufacturing technologies, surface finishing and processing, quality control methods, price of specific components, complex conditions in core and vicinity, history of operation, inspection and verification of components, radioactive waste characterization at the end of life of components. Limited assessment of materials properties and suitability for certain application without considering the each individual component load, exposure and life time, may produce limited information on material itself in fact the issue is the selection criteria for a standard material suitable for a certain application and consequent failure of components. The degradation and ageing are specific to components starting from design, manufacturing technology and expected life when the component should be replaced. The paper presents the practical experience on maintenance requirements specific to TRIGA core components and some techniques of material investigations available at Institute for Nuclear Research Pitesti Post Irradiation Laboratory as well as in the Materials Development and Research Department. Some consideration concerning correlation between the reactor

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

    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)

  10. Biological Tests for Boron Neutron Capture Therapy Research at the TRIGA Mark II Reactor in Pavia

    The thermal column of the TRIGA Mark II reactor of the Pavia University is used as an irradiation facility to perform biological tests and irradiations of living systems for Boron Neutron Capture Therapy (BNCT) research. The suitability of the facility has been ensured by studying the neutron flux and the photon background in the irradiation chamber inside the thermal column. This characterization has been realized both by flux and dose measurements as well as by Monte Carlo simulations. The routine irradiations concern in vitro cells cultures and different tumor animal models to test the efficacy of the BNCT treatment. Some results about these experiments will be described. (author)

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

    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.)

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

    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

  13. Study of U-Zr alloy fabrication for TRIGA reactor fuels

    Triga-IPR/R1 research reactor has a U-Zr H alloy as a fuel that contains 8% wt of uranium, 91% wt of zirconium and 1% wt of hydrogen. It was developed a project with the objective to acquire capacitation in manufacturing fuel with improved neutronics characteristics. The first phase of this project is described in this paper. It includes the melting in an arc furnace under vacuum (VAR), consumable electrode pressing and welding, and mechanical forming. It is also presented some characterization results. (author)

  14. Biological Tests for Boron Neutron Capture Therapy Research at the TRIGA Mark II Reactor in Pavia

    Protti, N.; Ballarini, F.; Bortolussi, S.; De Bari, A.; Stella, S.; Altieri, S. [Department of Nuclear and Theoretical Physics, University of Pavia, Pavia (Italy); Nuclear Physics National Institute (INFN), Pavia (Italy); Bruschi, P. [Department of Nuclear and Theoretical Physics, University of Pavia, Pavia (Italy); Bakeine, J.G.; Cansolino, L.; Clerici, A.M. [Laboratory of Experimental Surgery, Department of Surgery, University of Pavia, Pavia (Italy)

    2011-07-01

    The thermal column of the TRIGA Mark II reactor of the Pavia University is used as an irradiation facility to perform biological tests and irradiations of living systems for Boron Neutron Capture Therapy (BNCT) research. The suitability of the facility has been ensured by studying the neutron flux and the photon background in the irradiation chamber inside the thermal column. This characterization has been realized both by flux and dose measurements as well as by Monte Carlo simulations. The routine irradiations concern in vitro cells cultures and different tumor animal models to test the efficacy of the BNCT treatment. Some results about these experiments will be described. (author)

  15. Modification of the radial beam port of ITU TRIGA Mark II research reactor for BNCT applications.

    Akan, Zafer; Türkmen, Mehmet; Çakir, Tahir; Reyhancan, İskender A; Çolak, Üner; Okka, Muhittin; Kiziltaş, Sahip

    2015-05-01

    This paper aims to describe the modification of the radial beam port of ITU (İstanbul Technical University) TRIGA Mark II research reactor for BNCT applications. Radial beam port is modified with Polyethylene and Cerrobend collimators. Neutron flux values are measured by neutron activation analysis (Au-Cd foils). Experimental results are verified with Monte Carlo results. The results of neutron/photon spectrum, thermal/epithermal neutron flux, fast group photon fluence and change of the neutron fluxes with the beam port length are presented. PMID:25746919

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

    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)

  17. Recent, ongoing, and projected academic uses of the Cornell TRIGA reactor

    This paper continues a series of reports on the academic uses of the Cornell TRIGA reactor presented at American Nuclear Society meetings in Boston in 1992, San Francisco in 1993, Washington in 1994, and Philadelphia in 1995. Many of the uses previously described have continued at similar or altered levels, but there are important differences. We begin with a description of a major change in organizational structure that will strongly influence the overall pattern of future use. The next section updates facility developments, followed by examples of use. We conclude with a forecast of emerging patterns for future academic use

  18. A wide range in-core neutron monitoring system for high powered TRIGA reactors

    High power movable core TRIGA reactors present unique problems of determining power levels from a neutron flux measurement because of (1) difficulty of locating detectors; (2) water thermal effects and (3) effect of experimental facilities. A solution, along with experimental results, will be described that uses a beam tube to effectively make in-core flux measurements with an out-of-core detector. The application of this new type of detector assembly to wide range linear and log power measurement will also be discussed. (author)

  19. Operational Experience with the TRIGA Mark II Reactor of the University of Pavia

    The Laboratory of Applied Nuclear Energy (LENA) is an Interdepartmental Research Centre of the University of Pavia which operates a 250 kW TRIGA Mark II Research Nuclear Reactor, a Cyclotron for the production of radioisotopes and other irradiation facilities. The reactor is in operation since 1965 and many home-made upgrading were realized in the past years in order to assure a continuous operation of the reactor for the future. The annual reactor operational time at nominal power 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, BNCT research, samples irradiation and training. In specific, few tens of hours of reactor operation per year are dedicated to training courses for University students and for professionals. Besides, the LENA Centre hosts every year more than one thousand high school students in visit. Lately, LENA was certified ISO 9001:2008 for the ''operation and maintenance of the reactor'' and for the ''design and delivery of the irradiation service''. Nowadays the reactor shows a good technical state and, at the moment, there are no political or economical reason to consider the reactor shut-down. (author)

  20. Operational Experience with the TRIGA Mark II Reactor of the University of Pavia

    Tigliole, A. Borio Di; Alloni, D.; Cagnazzo, M.; Coniglio, M.; Lana, F.; Losi, A.; Magrotti, G.; Manera, S.; Marchetti, F.; Pappalardo, P.; Prata, M.; Provasi, M.C.; Salvini, A.; Scian, G.; Vinciguerra, G. [University of Pavia, Laboratory of Applied Nuclear Energy (L.E.N.A), Via Aselli 41, 27100 Pavia (Italy)

    2011-07-01

    The Laboratory of Applied Nuclear Energy (LENA) is an Interdepartmental Research Centre of the University of Pavia which operates a 250 kW TRIGA Mark II Research Nuclear Reactor, a Cyclotron for the production of radioisotopes and other irradiation facilities. The reactor is in operation since 1965 and many home-made upgrading were realized in the past years in order to assure a continuous operation of the reactor for the future. The annual reactor operational time at nominal power 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, BNCT research, samples irradiation and training. In specific, few tens of hours of reactor operation per year are dedicated to training courses for University students and for professionals. Besides, the LENA Centre hosts every year more than one thousand high school students in visit. Lately, LENA was certified ISO 9001:2008 for the ''operation and maintenance of the reactor'' and for the ''design and delivery of the irradiation service''. Nowadays the reactor shows a good technical state and, at the moment, there are no political or economical reason to consider the reactor shut-down. (author)

  1. Utilization and operating experience of the 250 kw TRIGA Mark II research reactor in Ljubljana

    In its 35th year, the TRIGA Mark II 250 kW pulsing research reactor in Ljubljana is continuing its busy operation. With the maximum neutron flux in the central thimble of 10 13 n/cm 2 sec and many sample radiation positions the reactor has been used to perform many experiments in the following fields: solid state physics (elastic and inelastic neutron scattering), neutron dosimetry, neutron radiography, reactor physics including burn up measurements and calculations, boron neutron capture therapy and neutron activation analysis which represents one of the major usage of our reactor. Besides these, applied research around the reactor has been conducted, such as doping of silicon monocrystals, a routine production of various radioactive isotopes for industry ( 60Co, 64Zn, 24Na, 82Br) and medical use ( 18F, 99m Tc, etc.) and other activities. During the past decade the reactor was almost completely reconstructed (new grid plates, the control mechanisms and the control unit, modification of the spent fuel storage pool, etc). The main novelty in the reactor physics and operation features of the reactor was the installation of a pulse rod, therefore the reactor can be operated in a pulse mode. After reconstruction, the core was loaded with fresh 20% enriched fuel elements. In 1999 all spent fuel elements were shipped to the USA. (author)

  2. Current development at the Finnish TRIGA reactor towards the operation of the new BNCT irradiation facility

    The FiR 1-reactor, a 250 kW Triga reactor, with its subsystems has experienced a large renovation work. The main purpose of the upgrading has been to install the new Boron Neutron Capture Therapy (BNCT) irradiation facility. During the renovation the ventilation, electricity lines, water and waste water pipe lines and the reactor cooling system were renewed. The epithermal beam facility of the BNCT-irradiation station got its final form. In order to increase the availability of the reactor and the necessary systems in the reactor building all the pumps of the reactor cooling system and the main fans of the ventilation have been doubled. The reactor instrumentation is fed by an Uninterruptible Power System and a diesel aggregate feeds power to the reactor cooling and the ventilation systems, if the normal power supply fails. The epithermal neutrons are produced from the fast fission neutrons by a moderator block consisting of Al+AIF3 (FLUENTALTM), Which showed to be the optimum material for this purpose. Independently of the large renovation work the application for a new operating license for the reactor had to be submitted at the end of the year 1998 after nine years' operating time With the old license. (author)

  3. Design and safety considerations for the 10 MW(t) multipurpose TRIGA reactor in Thailand

    General Atomics (GA) is constructing the Ongkharak Nuclear Research Center (ONRC) near Bangkok, Thailand for the Office of Atomic Energy for Peace. The ONRC complex includes the following: A multipurpose 10 MW(t) research reactor; An Isotope Production Facility; Centralized Radioactive Waste Processing and Storage Facilities. The Center is being built 60-km northeast of Bangkok, with a 10 MW(t) TRIGA type research reactor as the centerpiece. Facilities are included for neutron transmutation doping of silicon, neutron capture therapy neutron beam research and for production of a variety of radioisotopes. The facility will also be utilized for applied research and technology development as well as training in reactor operations, conduct of experiments and in reactor physics. The multipurpose, pool-type reactor will be fueled with high-density (45 wt%), low-enriched (19.7 wt%) uranium-erbium-zirconium-hydride (UErZrH) fuel rods, cooled and moderated by light water, and reflected by beryllium and heavy water. The general arrangement of the reactor and auxiliary pool structure allows irradiated targets to be transferred entirely under water from their irradiation locations to the hot cell, then pneumatically transferred to the adjacent Isotope Production Facility for processing. The core configuration includes 4 x 4 array standard TRIGA fuel clusters, modified clusters to serve as fast-neutron irradiation facilities, control rods and an in-core Ir-192 production facility. The active core is reflected on two sides by beryllium and on the other two sides by D2O. Additional irradiation facilities are also located in the beryllium reflector blocks and the D2O reflector blanket. The fuel provides the fundamental safety feature of the ONRC reactor, and as a result of all the well established accident-mitigating characteristics of the UErZrH fuel itself (large prompt negative temperature coefficient of reactivity, fission product retention and chemical stability), a

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

    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.

  5. Consideration factors on the spent fuel shipment for PUSPATI TRIGA Reactor (RTP)

    Malaysian Institute for Nuclear Technology Research (MINT) operates a 1MW TRIGA MARK II type research reactor since 1982. The PUSPATI TRIGA Reactor (RTP) reached its first criticality on 23 June 1982 and since then, it has been used for beam experiments, neutron activation analysis, radioisotopes production, education and training. RTP uses three types of fuel elements, namely, 8.5 wt%, 12wt% and 20 wt%. For all the three type the enrichment level of U-235 is 20%. Until the end of 2005, RTP has accumulated 21 906 hrs of operation time, and 13 166 MWhrs of burnup. Based on the neutronics calculation, all the fuel elements are expected to be fully utilized by the year 2015. At present, there is no decision for the government to take part in return of the spent nuclear fuel back to the country of origin, where it was enriched. This paper describes the current status of the fuel elements and the availability of local infrastructure, considering the eventual agreement of the government to join the US Foreign Research Reactor Spent Nuclear Fuel Acceptance Programme for the shipment of the spent nuclear fuels. The involvement of national regulatory body is also briefly described. (author)

  6. Thermal hydraulic analysis of 3 MW TRIGA research reactor of bangladesh considering different cycles of burnup

    Burnup dependent steady state thermal hydraulic analysis of TRIGA Mark-II research reactor has been carried out utilizing coupled point kinetics, neutronics and thermal hydraulics code EUREKA-2/RR. From the previous calculations of neutronics parameters including percentage burnup of individual fuel elements performed so far for 700 MWD burnt core of TRIGA reactor showed that the fuel rod predicted as hottest at the beginning of cycle (fresh core) was found to remain as the hottest until 200 MWD of burn, but, with the progress of core burn, the hottest rod was found to be shifted and another rod in the core became the hottest. The present study intends to evaluate the thermal hydraulic parameters of these hottest fuel rods at different cycles of burnup, from beginning to 700 MWD core burnt considering reactor operates under steady state condition. Peak fuel centerline temperature, maximum cladding and coolant temperatures of the hottest channels were calculated. It revealed that maximum temperature reported for fuel clad and fuel centerline found to lie below their melting points which indicate that there is no chance of burnout on the fuel cladding surface and no blister in the fuel meat throughout the considered cycles of core burnt. (author)

  7. Thermal analysis of LEU modified Cintichem target irradiated in TRIGA reactor

    Actions conceived during last years at international level for conversion of Molybdenum fabrication process from HEU to LEU targets utilization created opportunities for INR to get access to information and participating to international discussions under IAEA auspices. Concrete steps for developing fission Molybdenum technology were facilitated. Institute of Nuclear Research bringing together a number of conditions like suitable irradiation possibilities, direct communication between reactor and hot cell facility, handling capacity of high radioactive sources, and simultaneously the existence of an expanding internal market, decided to undertake the necessary steps in order to produce fission molybdenum. Over the course of last years of efforts in this direction we developed the steps for fission Molybdenum technology development based on modified Cintichem process in accordance with the Argonne National Laboratory proved methodology. Progress made by INR to heat transfer computations of annular target using is presented. An advanced thermal-hydraulic analysis was performed to estimate the heat removal capability for an enriched uranium (LEU) foil annular target irradiated in TRIGA reactor core. As a result, the present analysis provides an upper limit estimate of the LEU-foil and external target surface temperatures during irradiation in TRIGA 14 MW reactor. (authors)

  8. Thermal Hydraulic Analysis of 3 MW TRIGA Research Reactor of Bangladesh Considering Different Cycles of Burnup

    M.H. Altaf

    2014-12-01

    Full Text Available Burnup dependent steady state thermal hydraulic analysis of TRIGA Mark-II research reactor has been carried out utilizing coupled point kinetics, neutronics and thermal hydraulics code EUREKA-2/RR. From the previous calculations of neutronics parameters including percentage burnup of individual fuel elements performed so far for 700 MWD burnt core of TRIGA reactor showed that the fuel rod predicted as hottest at the beginning of cycle (fresh core was found to remain as the hottest until 200 MWD of burn, but, with the progress of core burn, the hottest rod was found to be shifted and another rod in the core became the hottest. The present study intends to evaluate the thermal hydraulic parameters of these hottest fuel rods at different cycles of burnup, from beginning to 700 MWD core burnt considering reactor operates under steady state condition. Peak fuel centerline temperature, maximum cladding and coolant temperatures of the hottest channels were calculated. It revealed that maximum temperature reported for fuel clad and fuel centerline found to lie below their melting points which indicate that there is no chance of burnout on the fuel cladding surface and no blister in the fuel meat throughout the considered cycles of core burnt.

  9. A preliminary report on methods of measuring and reducing Argon-41 production by a TRIGA reactor

    Methods to accurately determine and techniques to reduce the Argon-41 released from the one-megawatt Geological Survey TRIGA Reactor facility have been developed. Knowledge of the composition of the exhaust-gas effluent is of prime importance to the U.S. Geological Survey in minimizing all radioactive releases to the environment. The counting systems and control measures have enabled the Geological Survey TRIGA Reactor staff to reduce the amount of Argon-41 released from the facility by a factor of two, with no reduction in operation level of the reactor. The counting system has also enabled the staff to categorize the principal sources of Argon-41. Under normal conditions, a fully-loaded rotating-specimen rack is by far the largest contributor. With the current counting system, 10-7 microcuries per cubic centimeter can be detected in the exhaust stack. It is intended to further improve this system to increase both the sensitivity and the reliability. The sensitivity is expected to be increased by utilizing a larger counting volume. To improve the reliability, it is planned to fabricate a loop parallel to the exhaust system, eliminating the need for a separate pump. (author)

  10. Production and use of {sup 18}F by TRIGA nuclear reactor: a first report

    Burgio, N.; Ciavola, C.; Festinesi, A.; Capannesi, G. [ENEA, Centro Ricerche Casaccia, Rome (Italy). Dipt. Innovazione

    1999-02-01

    The irradiation and radiochemical facilities at public research centre can contribute to the start up of the regional PET centre. In particular, the TRIGA reactor of Casaccia Research Centre could produce a sufficient amount of {sup 18}F to start up a PET centre and successively integrated the cyclotron production. This report establishes, in the light of the preliminary experimental works, a guideline to the reactor`s production and extraction of {sup 18}F in a convenient form for the synthesis of the most representative PET radiopharmaceutical: {sup 18}F-FDG. [Italiano] Le facilities di irraggiamento e i laboratori Radiochimici dei Centri Statali di Ricerca possono contribuire allo sviluppo di centri regionali PET (Tomografia ed Emissione Positronica). In particolare, il reattore TRIGA del Centro Ricerca Casaccia potrebbe produrre un quantitativo di {sup 18}F sufficiente alle attivita` formative propedeutiche al centro PET che, successivamente sarebbe in grado di avviare una propria produzione da ciclotrone. Questo rapporto stabilisce le linee guida sperimentali per la produzione del {sup 18}F da reattore nucleare e la sua successiva estrazione in una forma conveniente per la sintesi del piu` rappresentativo dei radiofarmaci PET: il {sup 18}F-FDG.

  11. First results of the new UCN source D at the TRIGA Mainz reactor

    Research in fundamental physics with the free neutron is one of the key tools for testing the Standard Model at low energies. Significant improvements of the experimental performance using ultracold neutrons (UCN) require reduction of both systematic and statistical errors. The development of new UCN sources based on the superthermal concept is therefore an important step. Besides the construction of new huge UCN sources at several big research centers around the world, there exists also the idea of competitive UCN sources using pulsed reactors of the TRIGA type. To demonstrate the feasibility of a UCN source at these reactors, a superthermal UCN source based on solid deuterium was built at the tangential beamport C of the reactor TRIGA Mainz. Based on the experience obtained during three years of successfull operation, a second upgraded source was built for the radial beamport D which should increase the UCN output by at least one order of magnitude. This talk gives an overview on the first UCN storage results at ''UCN D'', obtained during source commissioning in 2011.

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

    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)

  13. Pre-Analysis of Triga Mark II Reactor Cooling System

    AKAY, Orhan Erdal

    2012-01-01

    In this study, work of the reactor cooling system is divided into two time zone. The second cooling circuit has been that the conditions required operating. Cooling system which is the center of the heat exchanger total heat transfer coefficient correlations were calculated using the theoretical. The design values were compared with results obtained by calculation.

  14. Current utilization and long term strategy of the Finnish TRIGA research reactor FIR 1

    The Finnish TRIGA reactor, FiR 1, started operation in 1962. From early on the reactor created versatile research to support the national nuclear program as well as generally the industry and health care sector. Production of short-lived radioisotopes is still a basic service. Education and training play a role in the form of university courses and training of nuclear industry personnel in the Baltic region. In the 1990's a BNCT cancer treatment facility was build. Over 200 patient irradiations have been performed since May 1999. FiR 1 is one of the few facilities in the world providing these treatments. A long term strategy is being worked out for FiR 1 by VTT supported by an independent survey. The survey recommends operation of the reactor at least till 2016 to enable continuation of the promising development of BNCT in parallel of developing accelerator based neutron sources for this treatment. (author)

  15. Application of a triga research reactor as the neutron source for a production neutron radiography facility

    GA Technologies Inc. (GA) has developed a Stationary Neutron Radiography System (SNRS) using a 250-1000 KW TRIGA reactor as the neutron source. The partially below ground reactor will be equipped with four vertical beam tubes originating in the reactor graphite reflector and installed tangential to the core to provide a strong current of thermal neutrons with minimum gamma-ray contamination. The vertical beam tubes interface with rugged component positioning systems designed to handle intact F-111 aircraft wings, partial A-10 aircraft wings, pyrotechnics, and other honeycomb aircraft structures. The SNRS will be equipped with real-time, near-real-time, and film-radiographic imaging systems to provide a broad spectrum of capability for detection or corrosion of entrained moisture in large aircraft panels. (author)

  16. Application of a triga research reactor as the neutron source for a production neutron radiography facility

    GA Technologies Inc. (GA) has developed a Stationary Neutron Radiography System (SNRS) using a 250-1000 kW TRIGA reactor as the neutron source. The partially below ground reactor will be equipped with four vertical beam tubes originating in the reactor graphite reflector and installed tangential to the core to provide a strong current of thermal neutrons with minimum gamma-ray contamination. The vertical beam tubes interface with rugged component positioning systems designed to handle intact F-11 aircraft wings, partial A-10 aircraft wings, pyrotechnics, and other honeycomb aircraft structures. The SNRS will be equipped with real-time, near-real-time, and film-radiographic imaging systems to provide a broad spectrum of capability for detection of corrosion or entrained moisture in large aircraft panels

  17. Operating experience and maintenance at the TRIGA Mark II LENA reactor

    Reactor operation at the Triga Mark II LENA plant, at the University of Pavia, in the past two years has been greatly affected by fulfilment of the new Italian fire prevention act's requirements, by the final red-tape work to get the renewal of the operation licence and by answering to the observations of Inspectors of the Italian Ministry of Labour and Social Security. All personnel was involved in the revision of manuals and prescriptions according to government rules and new ideas on modern nuclear safety. Consequently reactor operation was largely reduced due to works going on in the plant and to the lack of practicability of the Radiochemistry Laboratory. Finally, at the end of May 1990, the Reactor Operation Licence was renewed for the time period 1990-1995 by the Italian Ministry of Industry. (orig.)

  18. Review of non-proliferable fuel options in research reactors (TRIGA)

    This study is to examine the following aspects resulted from uranium enrichment reduction in TRIGA Mark-III reactor: reactor performance, fuel cycle costs, safety and reliability, and non-proliferation aspects. Fuel models adopted are: existing fuels (20% and 70% enriched with 8.5 wt% U-loaded); proposed fuels (20% and 30% enriched with 12 wt% U-loaded, 40% enriched with 8.5 wt% U-loaded, and 20% enriched with 20 wt% U-loaded). As results, the proposed fuels are disadvantageous over current FLIP-fuelled core in safety margins, reliability and reactor performance. Besides, operating costs will double with 12 wt% U-loaded fuels than with FLIP fuels

  19. Data acquisition and display system for the 14 MW SSR TRIGA Reactor from INR Pitesti

    This paper briefly describes the data acquisition system used for almost ten years at the 14 MW SSR TRIGA Reactor in INR Pitesti. Starting with a small system for the reactor thermal power evaluation, the system was developed to cover multiple experiment monitoring requirements. With a relative large number of channels (280) it satisfies now the necessity of monitoring both the 14 MW Reactor and different irradiation experiments for the steady-state mode at high resolution (10 μV, equivalent to 0.25 deg C for type K thermocouples) in a noisy environment. The user interface allows multiple user-configuration displays and both predefined and/or user-defined printouts. A real time system and multi-user capabilities offers both fast response and a very flexible data processing system. (authors)

  20. Main configurations of the reactor core TRIGA Mark III of the ININ, during their operation

    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)

  1. Characterization of the TRIGA Mark III reactor for k0 neutron activation analysis

    The k0 standardization for instrumental neutron activation analysis is a relatively new nuclear analytical technique. It is extended i n more than 20 countries of the world with reactor facilities, including some from Latin America. The great advantages of this technique (low uncertainties, fast and massive analysis, no standard necessity) with respect to relative, absolute and radiochemical activation analysis, are the reason of it fast introduction in Geology, Medicine, Agriculture and other fields of applications. But for the k0 instrumental neutron activation analysis implementation, the good knowledge of some reactor neutron flux and isotopes characteristics is necessary. The non ideality of the epithermal neutron flux temperature (Tn) and the k0 factors for more than 20 isotopes were determinate in the 3 typical irradiation positions of the TRIGA Mark III reactor of the National Nuclear Research Institute, Salazar, Mexico, using different experimental methods with conventional and non-conventional monitors

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

    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.

  3. Neutron spectra at two beam ports of a TRIGA Mark III reactor loaded with HEU fuel.

    Vega-Carrillo, H R; Hernández-Dávila, V M; Aguilar, F; Paredes, L; Rivera, T

    2014-01-01

    The neutron spectra have been measured in two beam ports, one radial and another tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research in Mexico. Measurements were carried out with the reactor core loaded with high enriched uranium fuel. Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a (6)LiI(Eu) scintillator and 2, 3, 5, 8, 10 and 12 in.-diameter high-density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code. For each spectrum total flux, mean energy and ambient dose equivalent were determined. Measured spectra show fission, epithermal and thermal neutrons, being harder in the radial beam port. PMID:23746708

  4. Design of a multipurpose 10 MW TRIGA reactor for experiments

    The optimum design for a 10 MW multipurpose reactor must take into account the broad range of applications planned for the facility. These applications depend on one or more of the following features of the reactor: - high neutron fluxes; - location of thermal and/or fast neutron irradiations; - beamports with sources of thermal or fast neutrons; - beamports for cold (cryogenic) or hot (∼ 2000 deg K) sources; - provision for pneumatic transfer of activation analysis targets; - convenient, user-friendly access to a hot cell for handling irradiated targets; - ample, convenient shielded storage for irradiated targets and experiments, handling tools, and spent reactor fuel. After selecting the reactor fuel guided by the requirements for fuel safety, operational convenience, and implications of the in-core irradiation sites, the selection of reflector is extremely important. This is because it is usually the reflector which provides the locations for many irradiation locations and sources for beamport with cold, thermal, epithermal, and fast neutrons. It is clearly the reflector which must accommodate the cold (cryogenic) source or the hot ( ∼ 2000 deg K) source for the associated beamports. Several aspects of reflectors are discussed having in view that these provide optimum features for the multipurpose reactor. For in-core irradiations, different reflectors affect mainly the reactivity and, hence, lifetime performance of the core. However, many types of applications require locations in the reflector. The choice of reflector(s) will have significant effect on the available neutron flux levels. Beryllium and D2O have special advantages. The advantages of D2O lie in its superior thermal neutron flux levels and broad spatial distribution. For beryllium, one of its advantages is that it is a solid. The reflector can therefore be assembled from components that can be interchanged and/or replaced as needed. Unfortunately, its thermal neutron flux is less than that for

  5. Utilization of the 250 kW TRIGA Mark II reactor in Ljubljana. Thirty years of experiences

    In its 30th year, the TRIGA Mark II 250 kW pulsing reactor is continuing its busy operation. With the maximum neutron flux in the central thimble of 1.1013 n/cm2 sec and many sample radiation positions the reactor has been used for a number of sophisticated experiments in the following fields: solid state physics (elastic and inelastic scattering of neutrons), neutron dosimetry, neutron radiography, reactor physics including nuclear burn up measurements and calculations and neutron activation analysis which represents one of the major usage of our reactor. Besides these, applied research around the reactor has been conducted, such as dopping of silicon monocrystals, a routine production of various radioactive isotopes for industry and medical use (18F,99mTc). At the Nuclear Training Centre the TRIGA reactor is the main teaching equipment. This training centre can fulfil the training requirements of the first Slovenian Nuclear Power Plant Krsko. (orig.)

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

    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. A Multi-Phased Sampling Effort to Characterize a University TRIGA Research Reactor

    A radiological characterization project was conducted at the University of Illinois (University) TRIGA research nuclear reactor in July 2005 as part of the long-term facility decommissioning project. The characterization effort included multiple survey and sampling techniques designed to assess both contamination of the reactor building and equipment and activation of reactor components and the reactor bio-shield. Radiation measurements included alpha and beta surface contamination measurements, gamma dose rate measurements, and gross gamma radiation measurements. Modeling was conducted based on the field measurements to predict concentrations of activation products in reactor components that were not directly sampled. The sampling effort included collecting removable contamination swipes, concrete samples from the reactor room floor and bio-shield, soil samples from below and around the perimeter of the reactor building, graphite samples from graphite moderator, and metal samples from reactor components. Concrete samples were obtained using an innovative technology that allowed for quick sample collection and analysis. Concrete, soil, graphite, and metal samples were analyzed on-site using liquid scintillation counters and gamma spectroscopy. Additional samples were sent off-site for analysis. (authors)

  8. Spent Fuel Management and Storage at the Finnish FIi TRIGA Reactor

    The FiR 1 reactor, a 250 kW TRIGA reactor, has been in operation since 1962. Presently spent fuel elements are stored in the racks on the walls of the reactor tank and in dry storage pits. After a sufficient cooling, the spent fuel elements are transferred from the reactor tank to the storage pits. The handling of fuel elements is made easier if the elements to be transferred have been stored for several years in the reactor tank. The capacity of the dry storage pits is sufficient for all the spent fuel elements of the reactor. After an eventual shut down of the reactor, the spent fuel temporary storage at site will only be required for quite a limited time if a decision is made to return the spent fuel to the USA. If we decide to continue operation of the reactor after the expiry date of the USDOE spent fuel acceptance policy (May 2016), the spent fuel can be incorporated into the domestic final repository for spent fuel of the Finnish Nuclear Power Plants. The final disposal facility is due to start operations in 2020. (author)

  9. Some aspects related to the management of maintenance for a TRIGA research reactor In Romania

    Safety management for a nuclear research reactor involves 'good dependability management' of operations activities, such as: reliability, availability, maintainability and maintenance support. In order to evaluate the safety management aspects intended to be applied by research reactor management, the performance dependability indicators and their impact on reactor availability and reactor safety have to be established. The document ISO 9000-4/IEC 300-1 'Dependability Management' (1995), describes five internationally agreed indicators of reactor equipment dependability. Each of them can be used for corrective maintenance or for preventive maintenance, such as: I1 - equipment Maintenance Frequency; I2 - equipment Maintenance Effort; I3 - equipment Maintenance Downtime Factor; I4 - equipment Maintenance Contribution to the System Function Downtime Factor; I5 - equipment Maintenance Contribution to the reactor Capability Loss Factor. This paper presents an evaluation of those 5 mentioned indicators with reference to the primary circuit of the INR's TRIGA research reactor and conclusion. The analyzed period was between 1994- 1999. It is to be noted that this type of analysis is performed for the first time for a research reactor. (author)

  10. Plan for the safe decommissioning of the BAEC 3MW TRIGA MARK-II research reactor

    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 (131I, 99mTc, 46Sc), 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 remained suspended for about 4 years. However, the reactor operation was continued during this period at a power level of 250 kW to cater the needs of various R and D groups, which required lower neutron flux for their experiments. This was made possible by establishing a temporary by pass connection across the decay tank using local technology. The reactor was made operational again at full power after successful replacement of the damaged decay tank in August 2001. At present the reactor is operated 5 days a week at a full power level of 3 MW for production of I-131 and R and D purposes. Up to December 2005 total burn-up of the core stands at about 358 Megawatt Days (MWDs). BAEC has planned to increase the production of 131I and as such, the core burn-up is expected to be increased very significantly in the years to come. There is a declaration from the US DOE that all US origin research reactor spent fuel generated within 2006 will be taken away to the USA at their own cost within 2009. But the fuel burn up of the BAEC research reactor is about 6%. So the reactor can operate for about 10-20 years more. An initial decommissioning plan for the BAEC TRIGA reactor and relevant facilities should be established as early as possible as recommended in the IAEA Safety Standards Series No.WS-G-2.1 (Decommissioning of Nuclear Power Plants and Research Reactors - Safety Standards Series No.WS-G-2.1, IAEA, Vienna, 1999). During the design and construction

  11. Power upgrading tests of the TRIGA IPR-R1 nuclear reactor to 250 kW

    The thermal power of the TRIGA IPR-R1 reactor, at the Center of Development of Nuclear Technology - CDTN, was increased from 100 kW to 250 kW. To obtain this goal, four fresh fuel elements were inserted in the reactor core. The paper gives a brief overview of experiments that have been performed to test the safety conditions of the new core configuration in order to obtain the license to operate the reactor at the new power. (author)

  12. Calculation of neutron flux in PUSPATI TRIGA MARK II reactor using Monte-Carlo n-particle approach

    A Monte Carlo simulation of neutron flux at the TRIGA MARK II PUSPATI (RTP) nuclear research reactor at Agensi Nuklear Malaysia was carried out using the MCNP5 program. The objective of the work is to simulate the neutron flux inside the reactor core. Calculations of neutron flux for fast and thermal neutron were carried out under the conditions in which the control rod was either fully withdrawn from or fully inserted into the reactor. (Author)

  13. Power density mapping in a 14-MW TRIGA MTR reactor

    Ciocanescu, M.; Lupu, M.; Costescu, C.; Racataianu, I.; Simionovici, D. (Institute for Nuclear Power Reactors, Pitesti (Romania))

    1984-07-01

    For small, high-flux testing reactors, tri dimensional analysis of power density is required to control hot points and core configuration to design and conduct the irradiation experiments. In this respect, an experimental program and computational system are reported. The experimental program, based on fuel gamma scanning, provides reference data on reactor power peaking. Particular attention is paid to transversal and axial perturbations introduced by the typical control rods insertion and to gradients near irradiation locations. The computational system, based on three-dimensional neutronic analyses, two groups with control rod regions tested on the above data, shows the general, influence of the control system on the power peaking (pin and axial) and on the experimental location fluxes, which fill a gap in the GA Safety Report Analysis data. (orig.)

  14. Effective physics-based uncertainty quantification for ZrHx thermal neutron scattering in TRIGA reactors

    The thermal neutron scattering cross sections of ZrHx are heavily affected by the solid frequency distributions, also called “phonon spectra”, of Zr and H in ZrHx. Although the phonon spectra vary for different x in ZrHx, current reference data, e.g. ENDF, are based on ZrH2. This may introduce non-negligible errors in the simulations for TRIGA reactors. In the previous work, we have proposed parameterized phonon spectrum (PPS) models to explore the effects of changing the spectra by varying the parameters and investigated the effects on reactivity and fuel fission rate density on TRIGA lattice model. In this work, we extend the analyses to quantities of interest (QOIs) on the realistic full-core geometry of the TRIGA reactor at Texas A and M University. In this work, we sampled the parameters with Latin Hypercube sampling designs (LHS) in a novel way and generated corresponding phonon spectra. NJOY and MCNP were used to carry out the calculations. We investigated reactivity (ρ), neutron mean generation time (Λ), fuel temperature feedback coefficient αTFuel(293.6 - 600 K) and ex-core/in-core detector reaction rates. Statistical analyses indicate ρ, Λ and αTFuel are sensitive to the parameters while other QOIs are not. We calibrated the parameters for ENDF-VII as a surrogate for experimental data. Results show the feasibility of the parameter calibrations. Future work will perform experiments to archive QOIs and to calibrate the parameters in the PPS model to generate thermal scattering cross sections used in future simulations. (author)

  15. Monte Carlo simulation for calculation of kinetic parameters in an Accelerator Driven Subcritical TRIGA reactor

    Highlights: • Among the kinetic parameters, the most important ones are βeff and Λ. • Several methods including the Rossi-α and Feynman-α techniques, slope fit and MCNPX code have been investigated. • The Monte Carlo MCNPX code was used to simulate a geometrical model of the TRIGA core. • The results of the methods have been validated. - Abstract: In this study, noise analysis techniques including Feynman-α (variance-to-mean) and Rossi-α (correlation) and dynamic method such as slope fit method have been used to calculate effective delayed neutron fraction (βeff) and neutron reproduction time (Λ) in Accelerator Driven Subcritical TRIGA reactor. The obtained results have been compared with MCNPX code results. The relative difference between MCNPX code with Feynman-α and Rossi-α techniques and slope fit method for βeff are approximately −5.4%, 1.2%, and −10.6%, −14.8%, respectively, and also for Λ is approximately 2.1%. According to results, the noise methods can been considered ideal for detection with high efficiency and zero dead time and in the slope fit method, the decay of the delayed neutrons has been neglected and only the prompt neutrons have been taken into account. In addition, quantities simulated in the current study are validated against both the reference data and the results of MCNPX code. Therefore, the purpose of this study is to simulate the commonly used experimental methods by MCNPX code and investigate the convergence as well as accuracy of the computational results for different analysis methods in calculation of the kinetic parameters in an Accelerator Driven Subcritical TRIGA reactor

  16. Design of a 16-pin fuel cluster for conversion of MTR-plate type reactors to TRIGA

    General Atomic is currently designing a 16-rod TRIGA fuel cluster to be used in converting and upgrading MTR-plate type reactor cores. The TRIGA conversion cluster is designed to operate at power levels up to 10 MW, however, the achievable power level will be dependent upon the cooling system available in the converted reactor. A coolant flow rate of about 18,900 L/min (5000 GPM) is needed for 10 MW operation. A cooling system with a flow rate of 8300 L/min (2200 GPM) will allow 5 MW operation. The fuel rods used in the conversion bundle are identical in size and uranium content to the fuel rods to be used in the 14 MW TRIGA core being built for the Romanian Institute for Nuclear Technologies. The 12.95 mm (0.51 in.) OD TRIGA fuel-moderator material is U-Er-ZrH1.6 with 10 wt-% uranium (93% enriched). Erbium is included as a burnable poison and is a major contributor to the prompt negative temperature coefficient - the dominant safety feature of TRIGA fuel. Fuel rods are clad with 13.77 mm (0.542 in.) OD incoloy and have an active fuel height of 558.8 mm (22.0 in.). Fuel rod spacing within the cluster is identical to the 14 MW TRIGA design, with 2.54 mm (0.10 in.) between fuel rods and between rods and the cluster shroud. Two intermediate spacers are used within the cluster to maintain clearances along the length of the fuel rods. Maintaining the rod dimensions and spacing equivalent to the 14 MW TRIGA design allows utilization of existing nuclear, thermal and mechanical design information in developing the 16-pin cluster design. (author)

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

    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

  18. Complex degradation and ageing phenomena of research reactor core structural materials - experience at 14 MW TRIGA reactor from INR Pitesti

    The 14 MW TRIGA Research Reactor designed in the early '70s is a relative new research reactor with an operational experience of 30 years. The specific design of reactor core objectives, were to manufacture, build and operate a flexible structure which incorporate previous experience of pool type research reactors. Aluminum alloy 6061 and stainless steel are only materials used for core structural components, which are all easily remotely removable and replaceable by simple hand tools. Properties of those categories of materials were well characterized / known for many other reactors predecessors, and no special criteria or preliminary tests were performed. The mechanical core structure is presented in the paper and designed procedure for periodic testing and inspection is also described. In spite of well known materials properties, the behavior uncertainties of those materials in each reactor case may have special aspects related to design of components, manufacturing technologies, surface finishing and processing, quality control methods, price of specific components, complex conditions in core and vicinity, history of operation, inspection and verification of components, radioactive waste characterization at the end of life of components. Limited assessment of materials properties and suitability for certain application without considering the each individual component load, exposure and life time, may produce limited information on material itself, in fact the issue is the selection criteria for a standard material suitable for a certain application and consequent failure of components. The degradation and ageing are specific to components starting from design, manufacturing technology and expected life when the component should be replaced. The paper presents the practical experience on maintenance requirements specific to TRIGA core components and some techniques of material investigations available at Institute for Nuclear Research Pitesti Post Irradiation

  19. Taking samples from the reactor components in preparation for dismantling the TRIGA reactor at the Medical University of Hannover

    After shutting down the facility at the end of 1996 the spent TRIGA fuel elements from the research reactor at the Medical University of Hanover (MHH) were returned to the United States in the summer of 1999 and thus disposed of for the MHH. Consequently one of the main prerequisites for dismantling the TRIGA reactor as planned has been fulfilled. In preparation for dismantling the facility a number of samples were taken from the various reactor components in 2000. The aim of the samples being taken was to establish the radiological condition of the facility in more detail, in particular the condition of the activated components in the reactor tank and the biological shield in the core area. Up to now the calculated estimates for these components had been based mainly on the details provided in the facility documentation when operation started at the beginning of the 1970s, showing that the evaluation of the activity and dose rates was too high. This was confirmed in 1998 in the course of measuring contamination and dose rates when samples were taken from some reactor components before the fuel elements were removed. For example, drill samples were taken from the bottom part of a graphite blind element and from the central radiation beam tube in the core area and then analyzed by the U.R.A. Laboratory of the University of Regensburg. As it is planned to dismantle the reactor facility completely by hand, it is necessary to have realistic radiological data in order to prepare for the dismantling procedure. Furthermore, both the release of radioactive materials into the environment and the costs for external disposal of the radioactive waste from the dismantling of the reactor are to be kept to a minimum. (orig.)

  20. Applied research performed and in progress by using a Triga nuclear reactor

    The TRIGA reactor of ENEA Casaccia has been used in different applied research fields among which particularly the following applications of neutron activation analysis (NAA) are reported: a) Environmental. 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. b) Forensic. Many applications of NAA have been performed in this field 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. c) Plant nutrition. 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. d) Geological. 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. (author)

  1. Neutronics analysis of the current core of the TRIGA Mark II reactor Vienna

    This paper presents the part of PhD work performed at the TRIGA Mark II Vienna. A detailed three dimensional MCNP model of the reactor was developed. The neutronics library JEFF3.1 was applied to this model. The model was completed by employing the fresh fuel composition experiments and was confirmed by the initial criticality, reactivity distribution and thermal flux distribution performed in 1962. To analyse the current burned core, burn up and its relevant material composition was calculated by ORIGEN2 and confirmed by gamma spectroscopy of six spent Fuel Elements FE(s). This new material composition of the current core was incorporated into the already developed MCNP model. This paper presents the current core calculations employing MCNP5 and its experimental validation through criticality and reactivity distribution experiments, performed at the TRIGA Mark II research reactor Vienna. The MCNP predicts the criticality of the current core on loading of 78th FE in the core which is also confirmed experimentally. Five FE(s) were calculated and measured for their reactivity worths. The deviations between theoretical results and experimental observations were in range from 3% to 17%. (author)

  2. 3 MW TRIGA Research Reactor facility of BAEC and its Utilization

    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. [Bangladesh Atomic Energy Commission, Atomic Energy Research Establishment, Dhaka (Bangladesh)

    1999-08-01

    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)

  3. Cross sections for fuel depletion and radioisotope production calculations in TRIGA reactors

    For TRIGA Reactors, the fuel depletion and isotopic inventory calculations, depends on the computer code and in the cross sections of some important actinides used. Among these we have U-235, U-238, Pu-239, Pu-240 and Pu-241. We choose ORIGEN2, a code with a good reputation in this kind of calculations, we observed the cross sections for these actinides in the libraries that we have (PWR's and BWR), the fission cross section for U-235 was about 50 barns. We used a PWR library and our results were not satisfactory, specially for standard elements. We decided to calculate cross sections more suitable for our reactor, for that purpose we simulate the standard and FLIP TRIGA cells with the transport code WIMS. We used the fuel average flux and COLAPS (a home made program), to generate suitable cross sections for ORIGEN2, by collapsing the WIMS library cross sections of these nuclides. For the radioisotope production studies using the Central Thimble, we simulate the A and B rings and used the A average flux to collapse cross sections. For these studies, the required nuclides sometimes are not present in WIMS library, for them we are planning to process the ENDF/B data, with NJOY system, and include the cross sections to WIMS library or to collapse them using the appropriate average-flux and the program COLAPS. (author)

  4. Recent research programs at the TRIGA Mark II reactor in Ljubljana

    Recent developments and new research activities which make use of the TRIGA reactor in Ljubljana are reported. They are spread over a broad range of research fields from nuclear and solid state physics, reactor physics and engineering, neutron radiography, analytical chemistry, medicine and biology, and industrial applications. The following investigations are briefly described: Improvements in the thermal neutron beam facility for nuclear capture studies, a rotating crystal time-of-flight spectrometer and its use for measurements of dynamics of crystal lattices in liquid crystals and ferroelectrics, measurements by the fast neutron spectroscopy and dosimetry group of fission-spectrum averaged activation cross-sections for some threshold detectors; measurements of fast neutron spectra in standard TRIGA seed irradiation facilities and improvements of activation data unfolding program ITER II and its application to unfolding of single crystal fast neutron scintillation spectrometers, a simple nuclear power plant simulator to be used for education of plant personnel; neutron activation analysis falls into two parts: ecological studies of the uptake and distribution of mercury and some other micro-elements in particular in the Idrija area (mercury mining), and the development of methods for the analysis of trace elements in standard reference materials, biological samples, and high purity materials. (U.S.)

  5. Numerical simulation of non-steady state neutron kinetics of the TRIGA Mark II reactor Vienna

    Riede, J., E-mail: jriede@ati.ac.at; Boeck, H., E-mail: boeck@ati.ac.at

    2013-12-15

    Highlights: • Power changes after reactivity changes have been measured with high time resolution. • Time dependent power changes after reactivity changes have been calculated numerically including feedback mechanisms. • The model has been verified by comparing numerical results to experimental data. • The verified model has been used to predict time dependent power changes after several reactivity changes. - Abstract: This paper presents an algorithm for numerical simulations of non-steady states of the TRIGA Mark II reactor in Vienna, Austria. The primary focus of this work has been the development of an algorithm which provides time series of integral neutron flux after reactivity changes introduced by perturbations without the usage of thermal-hydraulic/neutronic numerical code systems for the TRIGA reactor in Vienna, Austria. The algorithm presented takes into account both external reactivity changes as well as internal reactivity changes caused by feedback mechanisms like effects caused by temperature changes of the fuel and poisoning effects. The resulting time series have been compared to experimental results.

  6. Numerical simulation of non-steady state neutron kinetics of the TRIGA Mark II reactor Vienna

    Highlights: • Power changes after reactivity changes have been measured with high time resolution. • Time dependent power changes after reactivity changes have been calculated numerically including feedback mechanisms. • The model has been verified by comparing numerical results to experimental data. • The verified model has been used to predict time dependent power changes after several reactivity changes. - Abstract: This paper presents an algorithm for numerical simulations of non-steady states of the TRIGA Mark II reactor in Vienna, Austria. The primary focus of this work has been the development of an algorithm which provides time series of integral neutron flux after reactivity changes introduced by perturbations without the usage of thermal-hydraulic/neutronic numerical code systems for the TRIGA reactor in Vienna, Austria. The algorithm presented takes into account both external reactivity changes as well as internal reactivity changes caused by feedback mechanisms like effects caused by temperature changes of the fuel and poisoning effects. The resulting time series have been compared to experimental results

  7. 3 MW TRIGA Research Reactor facility of BAEC and its Utilization

    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)

  8. Utilization and operating experience of the TRIGA Mark II research reactor in Ljubljana

    Dimic, V. (J. Stefan Institute, Ljubljana (Slovenia))

    1999-12-15

    The operating experience of the 250 kW TRIGA Mark-II reactor of the J. Stefan Institute in Ljubljana, Slovenia in the years 1996 and 1997 is reported. The reactor has been in operation without long undesired shut-down. In 1996 the production of energy was 401 MWh (around 1600 hours in operation) and there was 7 unplanned shut-downs because of electricity broke down. In 1997 the production of energy was 272 MWh (around 1090 hours in operation). In 1991 and 1997 the reactor was almost completely reconstructed and upgraded. The reconstruction consisted mainly of replacing the grid plates, the control rod mechanisms and the control unit. Recently, the new PC based system was adopted and developed to collect the operational radiation data of the reactor. The new wiring of the electric power system, part of the primary and secondary coolant system piping and the spent fuel storage pool have been modified and the new air-exchange system in the control room were installed. Because of this large reconstruction of the reactor, for the last years in the operation of the reactor no significant problems have been detected. The facility is expected to operate without major investment at least until 2006. The reactor has been utilized in the projects: Neutron activation analysis, Boron neutron capture therapy, Real time neutron radiography, Neutron tomography, and Dosimetry research. The activities of neutron activation analysis, neutron radiography and tomography as well as boron neutron capture therapy are shortly presented

  9. Utilization and operating experience of the TRIGA Mark II research reactor in Ljubljana

    The operating experience of the 250 kW TRIGA Mark-II reactor of the J. Stefan Institute in Ljubljana, Slovenia in the years 1996 and 1997 is reported. The reactor has been in operation without long undesired shut-down. In 1996 the production of energy was 401 MWh (around 1600 hours in operation) and there was 7 unplanned shut-downs because of electricity broke down. In 1997 the production of energy was 272 MWh (around 1090 hours in operation). In 1991 and 1997 the reactor was almost completely reconstructed and upgraded. The reconstruction consisted mainly of replacing the grid plates, the control rod mechanisms and the control unit. Recently, the new PC based system was adopted and developed to collect the operational radiation data of the reactor. The new wiring of the electric power system, part of the primary and secondary coolant system piping and the spent fuel storage pool have been modified and the new air-exchange system in the control room were installed. Because of this large reconstruction of the reactor, for the last years in the operation of the reactor no significant problems have been detected. The facility is expected to operate without major investment at least until 2006. The reactor has been utilized in the projects: Neutron activation analysis, Boron neutron capture therapy, Real time neutron radiography, Neutron tomography, and Dosimetry research. The activities of neutron activation analysis, neutron radiography and tomography as well as boron neutron capture therapy are shortly presented

  10. 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

    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)

  11. Validating the Serpent Model of FiR 1 Triga Mk-II Reactor by Means of Reactor Dosimetry

    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%.

  12. Validation of WIMS-SNAP code systems for calculations in TRIGA-MARK II type reactors; Validacion del sistema de codigos WIMS-SNAP para calculos en reactores nucleares tipo TRIGA-MARK II

    Hernandez Valle, S.; Lopez Aldama, D. [Centro de Investigaciones Nucleares, Tecnologicas y Ambientales, La Habana (Cuba). E-mail: svalle@ctn.isctn.edu.cu

    2000-07-01

    The following paper contributes to validate the Nuclear Engineering Department methods to carry out calculations in TRIGA reactors solving a Benchmark. The benchmark is analyzed with the WIMS-D/4-SNAP/3D code system and using the cross section library WIMS-TRIGA. A brief description of the DSN method is presented used in WIMS/d{sup 4} code and also the SNAP-3d code is shortly explained. The results are presented and compared with the experimental values. In other hand the possible error sources are analyzed. (author)

  13. Local radiation protection measurement network for TRIGA Reactor building

    This system is intended for the protection of personal working in those areas of the reactor building where high gamma radiation fields are expected. A detector, sensitive to gamma radiation, is installed in each of the areas to be monitored. The detector will send a signal proportional to the radiation level in the area, to a corresponding electronic module (Alarm Unit), where the signal will be amplified and checked against alarm set-points for possible alarming condition. In case the field exceeds the alarm set-point, the Alarm Unit will produce a signal that will trigger the field alarms (Horn and Beacon), located in the area where that condition occurred. Each Alarm Unit will send a numerical input to central computer command. (authors)

  14. Developing silicon irradiation devices at TRIGA - INR Pitesti reactor2

    The possibility of introducing doping impurities in semiconductors by making use of nuclear reactions was suggested in early '50. This method is based on thermal neutron capture in 30 Si and it results in a highly uniform distribution of the doping impurity. Three irradiation devices were built lately. All are disposed at the border of reactor core to prevent the influence of in-core experiments and allow operating several irradiation channels. The flux typical distribution was flattened by means of a water differential absorbent. During irradiation the silicon ingots are rotated to obtain a uniform exposure. The first irradiation device was filled with water while the other two, with graphite to get a fair thermalization and compatibility with the graphite thermal column meanwhile installed near the reactor core. The last device, installed in 1998, is provided with four irradiation channels disposed in a raw and four channels for monitors. The maximum diameter of the silicon ingots is 2.5 inches. Neutron flux-spectrum measurements as well as thermal neutron flux distribution measurements were conducted to have a neutronic characterization of the main irradiation channels. It was found that in the graphite filled device a better thermalization of the neutrons is obtained. Energetic neutrons (En > 1 MeV) represents only 1.6% of the total flux in the case of graphite devices as compared to 6.1% in case of air-filled device. The irradiation regime is to be adjusted by means of a permanent feedback based on electrical measurements of the materials obtained by doping process. The devices has a production capacity of 1000-1500 kg of average resistivity material in a year

  15. 77 FR 68155 - The Armed Forces Radiobiology Research Institute TRIGA Reactor: Facility Operating License No. R-84

    2012-11-15

    ... NRC's E-Filing rule (72 FR 49139; August 28, 2007). The E-Filing process requires participants to... filing requirements of the NRC's E-Filing Rule (72 FR 49139; August 28, 2007) apply to appeals of NRC... COMMISSION The Armed Forces Radiobiology Research Institute TRIGA Reactor: Facility Operating License No....

  16. University of Arizona TRIGA reactor. Annual utilization report, July 1, 1983-June 30, 1984 (Docket 50-113)

    This is the Annual Report covering the period July 1, 1983 through June 30, 1984, for the activities of the TRIGA Mark I Reactor at the University of Arizona, Tucson, Arizona. This report is submitted in compliance of Section 6.7e of the Facility Technical Specifications and Paragraph 50.59(b) of Title 10, Code of Federal Regulations

  17. 76 FR 69296 - University of Utah, University of Utah TRIGA Nuclear Reactor, Notice of Issuance of Renewed...

    2011-11-08

    ... published in the Federal Register on July 21, 2011 (76 FR 43733-43737). The NRC received no request for a..., 2011 (76 FR 60091-60094), and concluded that renewal of the facility operating license will not have a... COMMISSION University of Utah, University of Utah TRIGA Nuclear Reactor, Notice of Issuance of...

  18. The Failure Effect of Primary Coolant Pump to Thermo-Hydraulic Characteristic of TRIGA 2000 Reactor

    Has been done analysis of transient, when TRIGA 2000 reactor loss of primary coolant flow because primary pump loss of electric power, so fail in function.The calculation using RELAP5/MOD32 computer code with reactor core is modeled in the form of different seven channels as representation of different seven areas in core with 116 fuels. This reactor model also considers position of tip of primary pipe of input tank which is below of core, form of lower part core geometry influencing direction and coolant flow rate into core, and existence of diffuser system. The result of calculation in condition of steady state is obtained initiation condition of steady state is reached after 2500 seconds from reactor starts operation on 2000 kW power. On steady state, the channel-3 cladding temperature (hottest) is 149.63℃, the coolant temperature outlet from the channel-3 (hottest) is 105.66℃ , reactor inlet temperature is 32.2℃, and reactor outlet temperature is 46.79℃. The primary coolant entering reactor with flow rate 59.64 kg/s, distributed to core 31.44 kg/s and to by-pass of core or by-pass of chimney 28.20 kg/s. The result of calculation transient is obtained, before scram occur the channel-3 cladding temperature (hottest) is 161.03℃ and the coolant temperature outlet from the channel-3 (hottest) is 117.66℃. In the reactor core is a natural circulation as well (from reactor core, to chimney, to by-pass of chimney, to by-pass of core and back to platform) which is cooling reactor core. Scram occur on 250 seconds after failure of the primary pump. Based on result of this study is known that, when transient condition is happened because primary pump failure, reactor is predicted to stays in safety margin. (author)

  19. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul [Universiti Tenaga Nasional. Jalan Ikram-UNITEN, 43000 Kajang, Selangor (Malaysia); Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad [Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2016-01-22

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  20. Beam tube experiments and correlated research projects at the TRIGA reactor Vienna

    The four beam tubes and the thermal column at the TRIGA reactor Vienna were used intensively during the reporting period. Three of the beam tubes are mainly used for neutron spectroscopy such as small angle scattering, neutron interferometry and polarized neutrons where now investigations of magnetic structures in pulsed high magnetic fields (25 T) synchronized with the pulsed mode of the reactor have been started. The thermalizing column will be modified from the present cold neutron source to a comfortable neutron radiography installation which allows investigations of objects of a size up to 30 cm diameters. The thermal column is also used for neutron radiography and as a strong gamma source to investigate gamma irradiation effects on various materials such as glass fiber cables. In view of flexible utilization of the thermal column a movable shielding construction has been designed which is simple rolled away on the rails of the thermal column doors when access to the thermal column in necessary. (orig.)