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Sample records for california berkeley triga reactor

  1. The Berkeley TRIGA Mark III research reactor

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.)

  3. TRIGA research reactors

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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)

  5. Power calibrations for TRIGA reactors

    International Nuclear Information System (INIS)

    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. Political-social reactor problems at Berkeley

    International Nuclear Information System (INIS)

    For better than ten years there was little public notice of the TRIGA reactor at UC-Berkeley. Then: a) A non-student persuaded the Student and Senate to pass a resolution to request Campus Administration to stop operation of the reactor and remove it from campus. b) Presence of the reactor became a campaign-issue in a City Mayoral election. c) Two local residents reported adverse physical reactions before, during, and after a routine tour of the reactor facility. d) The Berkeley City Council began a study of problems associated with radioactive material within the city. e) Friends Of The Earth formally petitioned the NRC to terminate the reactor's license. Campus personnel have expended many man-hours and many pounds of paper in responding to these happenings. Some of the details are of interest, and may be of use to other reactor facilities. (author)

  7. TRIGA reactor operating experience

    International Nuclear Information System (INIS)

    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

  8. The research reactor TRIGA Mainz

    International Nuclear Information System (INIS)

    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

  9. TRIGA reactor health physics considerations

    International Nuclear Information System (INIS)

    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

    International Nuclear Information System (INIS)

    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

    International Nuclear Information System (INIS)

    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

    Energy Technology Data Exchange (ETDEWEB)

    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

    International Nuclear Information System (INIS)

    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. Operation experience with the TRIGA Reactor Vienna

    International Nuclear Information System (INIS)

    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

  15. The TRIGA reactor as chemistry apparatus

    International Nuclear Information System (INIS)

    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)

  16. Design improvements in TRIGA reactors

    International Nuclear Information System (INIS)

    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. Upgrading Status Of Bandung Triga 2000 Reactor

    International Nuclear Information System (INIS)

    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

  18. TRIGA research reactor activities around the world

    International Nuclear Information System (INIS)

    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

  19. BNCT activities at Slovenian TRIGA research reactor

    International Nuclear Information System (INIS)

    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)

  20. Component failure data base of TRIGA reactors

    International Nuclear Information System (INIS)

    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)

  1. PUSPATI Triga reactor fuel worth measurement

    International Nuclear Information System (INIS)

    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)

  2. Operation experience with the TRIGA reactor Wien

    Energy Technology Data Exchange (ETDEWEB)

    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.)

  3. TRIGA Mark-II, III reactor operation

    International Nuclear Information System (INIS)

    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)

  4. Optimum burnup of BAEC TRIGA research reactor

    International Nuclear Information System (INIS)

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

  5. Environmental Survey preliminary report, Lawrence Berkeley Laboratory, Berkeley, California

    Energy Technology Data Exchange (ETDEWEB)

    1988-07-01

    The purpose of this report is to present the preliminary findings made during the Environmental Survey, February 22--29, 1988, at the US Department of Energy (DOE) Lawrence Berkeley Laboratory (LBL) in Berkeley, California. The University of California operates the LBL facility for DOE. The LBL Survey is part of the larger DOE-wide Environmental Survey announced by Secretary John S. Herrington on September 18, 1985. The purpose of this effort is to identify, via no fault'' baseline Surveys, existing environmental problems and areas of environmental risk at DOE facilities, and to rank them on a DOE wide basis. This ranking will enable DOE to more effectively establish priorities for addressing environmental problems and allocate the resources necessary to correct them. Because the Survey is no fault'' and is not an audit,'' it is not designed to identify specific isolated incidents of noncompliance or to analyze environmental management practices. Such incidents and/or management practices will, however, be used in the Survey as a means of identifying existing and potential environmental problems. The LBL Survey was conducted by a multidisciplinary team of technical specialists headed and managed by a Team Leader and Assistant Team Leader from DOE's Office of Environmental Audit. A complete list of the LBL Survey participants and their affiliations is provided in Appendix A. 80 refs., 27 figs., 37 tabs.

  6. Environmental Survey preliminary report, Lawrence Berkeley Laboratory, Berkeley, California

    International Nuclear Information System (INIS)

    The purpose of this report is to present the preliminary findings made during the Environmental Survey, February 22--29, 1988, at the US Department of Energy (DOE) Lawrence Berkeley Laboratory (LBL) in Berkeley, California. The University of California operates the LBL facility for DOE. The LBL Survey is part of the larger DOE-wide Environmental Survey announced by Secretary John S. Herrington on September 18, 1985. The purpose of this effort is to identify, via ''no fault'' baseline Surveys, existing environmental problems and areas of environmental risk at DOE facilities, and to rank them on a DOE wide basis. This ranking will enable DOE to more effectively establish priorities for addressing environmental problems and allocate the resources necessary to correct them. Because the Survey is ''no fault'' and is not an ''audit,'' it is not designed to identify specific isolated incidents of noncompliance or to analyze environmental management practices. Such incidents and/or management practices will, however, be used in the Survey as a means of identifying existing and potential environmental problems. The LBL Survey was conducted by a multidisciplinary team of technical specialists headed and managed by a Team Leader and Assistant Team Leader from DOE's Office of Environmental Audit. A complete list of the LBL Survey participants and their affiliations is provided in Appendix A. 80 refs., 27 figs., 37 tabs

  7. Pneumatic transport systems for TRIGA reactors

    International Nuclear Information System (INIS)

    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

  8. Monte Carlo modelling of TRIGA research reactor

    Science.gov (United States)

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

    2010-10-01

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

  9. Decommissioning of TRIGA Mark II type reactor

    International Nuclear Information System (INIS)

    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

  10. Decommissioning of TRIGA Mark II type reactor

    Energy Technology Data Exchange (ETDEWEB)

    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

  11. 78 FR 26811 - Dow Chemical Company, Dow TRIGA Research Reactor; License Renewal for the Dow Chemical TRIGA...

    Science.gov (United States)

    2013-05-08

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  14. Utilisation of the Research Reactor TRIGA Mainz

    International Nuclear Information System (INIS)

    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

  15. Triga Mark III Reactor in paleotemperatures determination

    International Nuclear Information System (INIS)

    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

  16. Nondestructive examination of TRIGA reactor fuel elements

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  19. TRIGA reactor dynamics: Frequency response tests

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  1. Oregon State TRIGA reactor power calibration study

    International Nuclear Information System (INIS)

    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)

  2. TRIGA Mark-III reactor dismantling program

    International Nuclear Information System (INIS)

    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)

  3. Reconditioning of the TRIGA Mark III reactor

    International Nuclear Information System (INIS)

    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

  4. Ageing management for reactor TRIGA PUSPATI

    International Nuclear Information System (INIS)

    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)

  5. Evaluation of TRIGA Mark II reactor in Turkey

    International Nuclear Information System (INIS)

    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.)

  6. Analysis of fuel options in TRIGA reactor

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  8. A 5 MW TRIGA reactor design for radioisotope production

    International Nuclear Information System (INIS)

    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)

  9. Junior High Schools of Berkeley, California. Bulletin, 1923, No. 4

    Science.gov (United States)

    Preston, James T.; Clark, W. B.; Glessner, H. H.; Hennessey, D. L.

    1923-01-01

    This bulletin demonstrates that Berkeley, California's educational problem is and has been that of meeting the varied needs of a population such as may be found in any typical American city. The varied population needs, together with the rapid growth, have brought many difficult problems to Berkeley, just has they have to other cities. Based on…

  10. Calculation of Kinetic Parameters of TRIGA Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Snoj, Luka; Kavcic, Andrej; Zerovnik, Gasper; Ravnik, Matjaz [' Jozef Stefan' Institute, Jamova 39, SI-1000 Ljubljana (Slovenia)

    2008-10-29

    Modern Monte Carlo transport codes in combination of fast computer clusters enable very accurate calculations of the most important reactor kinetic parameters. Such are the effective delayed neutron fraction, {beta}{sub eff}, and mean neutron generation time, {lambda}. We calculated the {beta}{sub eff} and {lambda} for various realistic and hypothetical annular TRIGA Mark II cores with different types and amount of fuel. It can be observed that the effective delayed neutron fraction strongly depends on the number of fuel elements in the core or on the core size. E.g., for 12 wt. % uranium standard fuel with 20 % enrichment, {beta}{sub eff} varies from 0.0080 for a small core (43 fuel rods) to 0.0075 for a full core (90 fuel rods). It is interesting to note that calculated value of {beta}{sub eff} strongly depends also on the delayed neutron nuclear data set used in calculations. The prompt neutron life-time mainly depends on the amount (due to either content or enrichment) of {sup 235}U in the fuel as it is approximately inversely proportional to the average absorption cross-section of the fuel. E.g., it varies from 28 {mu}s for 30 wt. % uranium content fuelled core to 48 {mu}s for 8.5 wt. % uranium content LEU fuelled core. The results are especially important for pulse mode operation and analysis of the pulses. (authors)

  11. Irradiation routine in the IPR-R1 Triga reactor

    International Nuclear Information System (INIS)

    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.)

  12. Small Angle Neutron Scattering instrument at Malaysian TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-10-01

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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  16. Component and operation experience of reactor TRIGA MARK II

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    Science.gov (United States)

    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

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

    International Nuclear Information System (INIS)

    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

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

    OpenAIRE

    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...

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Snoj, Luka, E-mail: luka.snoj@ijs.si [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Zerovnik, Gasper, E-mail: gasper.zerovnik@ijs.si [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Trkov, Andrej, E-mail: andrej.trkov@ijs.si [Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

    2012-03-15

    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. - Highlights: Black-Right-Pointing-Pointer TRIGA reactor at JSI suitable for irradiation under well defined conditions. Black-Right-Pointing-Pointer It features irradiation channels of different fluxes, spectra, and dimensions. Black-Right-Pointing-Pointer Computational model has been developed and experimentally verified. Black-Right-Pointing-Pointer The model used for optimization of experiments and evaluation of uncertainties.

  5. Preliminary neutronic design of TRIGA Mark II Reactor

    International Nuclear Information System (INIS)

    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

  6. The University of California,Berkeley

    Institute of Scientific and Technical Information of China (English)

    李琳

    2007-01-01

    The roots of the University of California go back to the gold rush days of 1849,when the drafters of the State Constitution,a group of vigorous and farsighted people,required the legislature to "encourage by all suitable means the promotion of intellectual,scientific,moral and agricultural improvement"of the people of California.These early

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

    International Nuclear Information System (INIS)

    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

  8. Neutronic analysis of the Geological Survey TRIGA Reactor

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-07-01

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

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    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

  14. Operational experience data base of TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    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.)

  15. Analysis of TRIGA reactor thermal power calibration method

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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 reactor noise analysis for a TRIGA Mark-II

    International Nuclear Information System (INIS)

    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)

  18. The optimal control of ITU TRIGA Mark II Reactor

    International Nuclear Information System (INIS)

    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)

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

    Science.gov (United States)

    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.

  20. Impact of a security event at a TRIGA reactor

    International Nuclear Information System (INIS)

    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

  1. PUSPATI Triga Reactor - First year in operation

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-01-22

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

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

    International Nuclear Information System (INIS)

    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)

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

    Science.gov (United States)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

  9. Decommissioning of the ICI TRIGA Mark I reactor

    Energy Technology Data Exchange (ETDEWEB)

    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)

  10. AFRRI TRIGA Reactor water quality monitoring program

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    OpenAIRE

    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...

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  16. Oregon State TRIGA Reactor (OSTR) console upgrading

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  18. Perturbation analysis of the TRIGA Mark II reactor Vienna

    Energy Technology Data Exchange (ETDEWEB)

    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.)

  19. 44 years of operation - The successful fuel history of the TRIGA Mark II reactor Vienna

    International Nuclear Information System (INIS)

    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)

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

    OpenAIRE

    M. H. Altaf; Badrun, N. H.

    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 ...

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

    OpenAIRE

    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...

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

    International Nuclear Information System (INIS)

    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)

  3. Fuel burnup analysis for the Moroccan TRIGA research reactor

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  7. BNCT-Project at the Finnish TRIGA Reactor

    International Nuclear Information System (INIS)

    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)

  8. New burnup calculation of TRIGA IPR-R1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  12. 77 FR 7613 - Dow Chemical Company; Dow Chemical TRIGA Research Reactor; Facility Operating License No. R-108

    Science.gov (United States)

    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... INFORMATION CONTACT: Geoffrey Wertz, Project Manager, Research and Test Reactors Licensing Branch, Division...

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

    International Nuclear Information System (INIS)

    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

  14. 77 FR 42771 - License Renewal for the Dow Chemical TRIGA Research Reactor

    Science.gov (United States)

    2012-07-20

    ... COMMISSION License Renewal for the Dow Chemical TRIGA Research Reactor AGENCY: Nuclear Regulatory Commission... Research Reactor is located on the Michigan Division of the Dow Chemical Company in Midland, MI and is a... INFORMATION CONTACT: Geoffrey A. Wertz, Project Manager, Research and Test Reactor Licensing Branch,...

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

    International Nuclear Information System (INIS)

    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.)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.)

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

    International Nuclear Information System (INIS)

    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.)

  18. Current activities at the FiR 1 TRIGA reactor

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.

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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  5. Monte Carlo Model of TRIGA Reactor to Support Neutron Activation Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Zerovnik, G.; Snoj, L.; Trkov, A. [Reactor Physics Department, Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

    2011-07-01

    The TRIGA reactor at Jozef Stefan Institute is used as a neutron source for neutron activation analysis. The accuracy of the method depends on the accuracy of the neutron spectrum characterization. Therefore, computational models on different scales have been developed: Monte Carlo full reactor model, model of an irradiation channel and deterministic code for self-shielding factor calculations. The models have been validated by comparing against experiment and thus provide a very strong support for neutron activation analysis of samples irradiated at the TRIGA reactor. (author)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  9. 78 FR 5840 - Notice of License Termination for University of Illinois Advanced TRIGA Reactor, License No. R-115

    Science.gov (United States)

    2013-01-28

    ... 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,...

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.

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

    CERN Document Server

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    CERN Document Server

    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...

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

    International Nuclear Information System (INIS)

    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

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

    Science.gov (United States)

    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.

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

    Science.gov (United States)

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

    2015-10-01

    For efficient utilization of research reactors, such as TRIGA Mark II reactor in Ljubljana, it is important to know neutron flux distribution in the reactor as accurately as possible. The focus of this study is on the neutron flux redistributions due to control rod movements. For analyzing neutron flux redistributions, Monte Carlo calculations of fission rate distributions with the JSI TRIGA reactor model at different control rod configurations have been performed. Sensitivity of the detector response due to control rod movement have been studied. Optimal radial and axial positions of the detector have been determined. Measurements of the axial neutron flux distribution using the CEA manufactured fission chambers have been performed. The experiments at different control rod positions were conducted and compared with the MCNP calculations for a fixed detector axial position. In the future, simultaneous on-line measurements with multiple fission chambers will be performed inside the reactor core for a more accurate on-line power monitoring system. PMID:26141293

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    DEFF Research Database (Denmark)

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

    2011-01-01

    The thermal column of the TRIGA reactor in Mainz is being used very effectively for medical and biological applications. The BNCT (boron neutron capture therapy) project at the University of Mainz is focussed on the treatment of liver tumours, similar to the work performed at Pavia (Italy) a few...

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

    International Nuclear Information System (INIS)

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

    International Nuclear Information System (INIS)

    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)

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

    Science.gov (United States)

    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

    International Nuclear Information System (INIS)

    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

    International Nuclear Information System (INIS)

    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

    International Nuclear Information System (INIS)

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

    International Nuclear Information System (INIS)

    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)

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

    Science.gov (United States)

    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).

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-02-12

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

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

    International Nuclear Information System (INIS)

    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)

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

    OpenAIRE

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

    Energy Technology Data Exchange (ETDEWEB)

    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.)

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    Science.gov (United States)

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

    2014-02-01

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    Huda, M.Q. E-mail: quamrul@dhaka.net; Rahman, M.; Sarker, M.M.; Bhuiyan, S.I

    2004-07-01

    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({alpha},{beta}) 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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  12. 10th European TRIGA users conference

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

  17. Hydrogeology and tritium transport in Chicken Creek Canyon,Lawrence Berkeley National Laboratory, Berkeley, California

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, Preston D.; Javandel, Iraj

    2007-10-31

    This study of the hydrogeology of Chicken Creek Canyon wasconducted by the Environmental Restoration Program (ERP) at LawrenceBerkeley National Laboratory (LBNL). This canyon extends downhill fromBuilding 31 at LBNL to Centennial Road below. The leading edge of agroundwater tritium plume at LBNL is located at the top of the canyon.Tritium activities measured in this portion of the plume during thisstudy were approximately 3,000 picocuries/liter (pCi/L), which issignificantly less than the maximum contaminant level (MCL) for drinkingwaterof 20,000 pCi/L established by the Environmental ProtectionAgency.There are three main pathways for tritium migration beyond theLaboratory s boundary: air, surface water and groundwater flow. Thepurpose of this report is to evaluate the groundwater pathway.Hydrogeologic investigation commenced with review of historicalgeotechnical reports including 35 bore logs and 27 test pit/trench logsas well as existing ERP information from 9 bore logs. This was followedby field mapping of bedrock outcrops along Chicken Creek as well asbedrock exposures in road cuts on the north and east walls of the canyon.Water levels and tritium activities from 6 wells were also considered.Electrical-resistivity profiles and cone penetration test (CPT) data werecollected to investigate the extent of an interpreted alluvial sandencountered in one of the wells drilled in this area. Subsequent loggingof 7 additional borings indicated that this sand was actually anunusually well-sorted and typically deeply weathered sandstone of theOrinda Formation. Wells were installed in 6 of the new borings to allowwater level measurement and analysis of groundwater tritium activity. Aslug test and pumping tests were also performed in the wellfield.

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

    International Nuclear Information System (INIS)

    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

  19. 15. European TRIGA Conference

    Energy Technology Data Exchange (ETDEWEB)

    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)

  20. 15. European TRIGA Conference

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  2. Unique applications of research reactors with TRIGA UZrHx fuel

    International Nuclear Information System (INIS)

    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

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

    OpenAIRE

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

    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 ...

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.)

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

    OpenAIRE

    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...

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    Science.gov (United States)

    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

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

    International Nuclear Information System (INIS)

    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

    OpenAIRE

    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. TRIGA reactor spent fuel pool under severe earthquake conditions

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

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

    OpenAIRE

    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 ...

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

    International Nuclear Information System (INIS)

    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

  9. TRIGA reactor spent fuel pool under severe earthquake conditions

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    OpenAIRE

    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...

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    Science.gov (United States)

    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

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

    OpenAIRE

    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...

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

    International Nuclear Information System (INIS)

    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

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

    Science.gov (United States)

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

    2015-09-01

    Nowadays the Pavia TRIGA reactor is available for national and international collaboration in various research fields. The TRIGA Mark II nuclear research reactor of the Pavia University offers different in- and out-core neutron irradiation channels, each characterised by different neutron spectra. In the last two years a campaign of measurements and simulations has been performed in order to guarantee a better characterisation of these different fluxes and to meet the demands of irradiations that require precise information on these spectra in particular for radiobiological and microdosimetric studies. Experimental data on neutron fluxes have been collected analysing and measuring the gamma activity induced in thin target foils of different materials irradiated in different TRIGA experimental channels. The data on the induced gamma activities have been processed with the SAND II deconvolution code and finally compared with the spectra obtained with Monte Carlo simulations. The comparison between simulated and measured spectra showed a good agreement allowing a more precise characterisation of the neutron spectra and a validation of the adopted method. PMID:25958412

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  6. TRIGA International - History of Training Research Isotope production General Atomics

    International Nuclear Information System (INIS)

    TRIGA conceived at GA in 1956 by a distinguished group of scientists including Edward Teller and Freeman Dyson. First TRIGA reactor Mk-1 was commissioned on 3 may 1958 at G.A. Characteristic feature of TRIGA reactors is inherent safety: Sitting can be confinement or conventional building. TRIGA reactors are the most prevalent in the world: 67 reactors in 24 countries. Steady state powers up to 14 MWt, pulsing up to 22,000 MWt. To enlarge the scope of its manufactured products, CERCA engaged in a Joint Venture with General Atomics, and in July 1995 a new Company was founded: TRIGA INTERNATIONAL SAS (50% GA, 50% CERCA; Head Office: Paris (France); Sales offices: GA San Diego (Ca, USA) and CERCA Lyon (France); Manufacturing plant: CERCA Romans. General Atomics ID: founded in 1955 at San Diego, California, by General Dynamics; status: Privately held corporation; owners: Neal and Linden Blue; business: High technology research, design, manufacturing, and production for industry and Government in the U.S. and overseas; locations: U.S., Germany, Japan, Australia, Thailand, Morocco; employees: 5,000. TRIGA's ID: CERCA is a subsidiary of AREVA, born in November 05, 1957. Activities: fuel manufacture for research reactor, equipment and components for high-energy physics, radioactive sources and reference sources; plants locations: Romans and Pierrelatte (France); total strength: 180. Since the last five years TRIGA has manufactured and delivered more than 800 fuel elements with a door to door service. TRIGA International has the experience to manufacture all types of TRIGA fuel: standard fuel elements, instrumented fuel elements, fuel followed control rods, geometry: 37.3 mm (1.47 in.), 35.8 mm (1.4 in), 13 mm (0.5 in), chemical Composition: U w% 8.5, 12, 20, 30 and 45 w/o, erbium and no erbium. TRIGA International is on INL's approved vendor list (ISO 9000/NQA) and is ready to meet any TRIGA fuel needs either in the US or worldwide

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

    Science.gov (United States)

    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.

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

    CERN Document Server

    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.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.

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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    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.)

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

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    International Nuclear Information System (INIS)

    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)

  18. Irradiation of Fusion Reactor Materials and Bio-Dosimeters in JSI TRIGA Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Snoj, Luka; Ravnik, Matjaz; Lengar, Igor; Rogan, Petra; Novak, Sasa; Sentjurc, Marjeta; Jeraj, Robert [' Jozef Stefan' Institute, Jamova 39, SI-1000 Ljubljana (Slovenia)

    2008-10-29

    The TRIGA research reactor at Jozef Stefan Institute (JSI) is used for irradiation of various samples. Recently two projects have been initiated; the development and improvement of future fusion reactor materials and the development of bio-dosimeters. They both demand a large number of irradiations under different conditions (neutron spectra and flux) and subsequent gamma spectral analyses of the samples. In order to characterize the neutron spectra and fluxes in different irradiation channels that are of vital importance for the quality of irradiation, a detailed computational model of the TRIGA Mark-II reactor with the MCNP Monte Carlo particle transport code was developed, experimentally validated and verified. The projects and the main results are presented in the paper. Irradiation of Eurofer and SiC samples in our reactor is feasible and gives reliable results for material development and optimization in fusion reactors provided that the irradiation is supported with detailed spectrum calculations to take into account also the contribution of fast neutron reactions. Detailed knowledge of gamma and neutron spectrum is vital also in experimental development and calibration of bio-dosimeters. It can be concluded that irradiation of non-standard materials requires support of detailed calculations of spectrum in irradiation facilities not only to improve the accuracy but even to make the irradiation methods feasible. Main isotopes contributing to the long-term activation of Eurofer have relatively short life-times (100 days to several years). This composition of Eurofer is suitable from the radioactive waste disposal point of view. However one should be aware that dose rates inside and around the Eurofer structures will be significant even several months of cooling. The highest contribution to the total activity (80%) in this time interval comes from {sup 182}Ta. It may be speculated that further optimization of Eurofer with respect to the activation could be

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    Science.gov (United States)

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    Directory of Open Access Journals (Sweden)

    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.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

  12. A High Operability Supervisory Digital System for TRIGA-Type Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Aronica, O.; Bove, R.; Cappelli, M.; Falconi, L.; Palomba, M.; Santoro, E.; Sepielli, M. [ENEA, UTFISST, Casaccia Research Center, Via Anguillarese, 301 Rome (Italy); Memmi, F. [University of Rome ' Roma Tre' , Department of Electrical Engineering, Via della Vasca Navale, 84 Rome (Italy)

    2011-07-01

    In this work, we propose an outline of a monitoring system to supervise variables coming from a fission nuclear reactor of TRIGA type (1-MW TRIGA reactor RC-1). The system can interface the control room instrumentation and can display the characteristic parameters (e.g. nuclear power, temperatures, flow rates, radiological parameters) in an intuitive, user-friendly way for plant operators. This aim is achieved using the Labview development environment. A front panel of a virtual instrument allows for a direct measure and a check that would not be possible by only reading the output data coming from the instruments of the control room, because of their standards and strict safety regulations. The acquisition system, for signals coming from the reactor, can process data and generate a detailed representation of the results. Statistics resulting from data analysis will be interpreted to optimize reactor management parameters. This system also includes a simulation tool to predict specific performances and investigate critical phenomena, or to optimize overall plant performances. In particular, it allows to have a feedback control and to perform predictive statistical surveys of all main process parameters. (author)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

  18. Cyclist safety on bicycle boulevards and parallel arterial routes in Berkeley, California.

    Science.gov (United States)

    Minikel, Eric

    2012-03-01

    This study compares the safety of bicyclists riding on bicycle boulevards to those riding on parallel arterial routes in Berkeley, California. Literature on the impact of motor vehicle traffic characteristics on cyclist safety shows that high motor vehicle speeds and volumes and the presence of heavy vehicles are all detrimental to cyclist safety. This suggests that cyclists may be safer on side streets than on busy arterials. Bicycle boulevards-traffic-calmed side streets signed and improved for cyclist use-purport to offer cyclists a safer alternative to riding on arterials. Police-reported bicycle collision data and manually collected cyclist count data from bicycle boulevards and parallel arterial routes in Berkeley, California from 2003 to 2010 are used to test the hypothesis that Berkeley's bicycle boulevards have lower cyclist collision rates and a lower proportion of bicycle collisions resulting in severe injury. While no significant difference is found in the proportion of collisions that are severe, results show that collision rates on Berkeley's bicycle boulevards are two to eight times lower than those on parallel, adjacent arterial routes. The difference in collision rate is highly statistically significant, unlikely to be caused by any bias in the collision and count data, and cannot be easily explained away by self-selection or safety in numbers. Though the used dataset is limited and the study design is correlational, this study provides some evidence that Berkeley's bicycle boulevards are safer for cyclists than its parallel arterial routes. The results may be suggestive that, more generally, properly implemented bicycle boulevards can provide cyclists with a safer alternative to riding on arterials. PMID:22269506

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.)

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

    OpenAIRE

    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 ...

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    Science.gov (United States)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    Directory of Open Access Journals (Sweden)

    C. A. M. Silva

    2014-01-01

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

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

    Directory of Open Access Journals (Sweden)

    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.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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.

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Rojas R, E.

    2014-07-01

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

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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)

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

    Directory of Open Access Journals (Sweden)

    Tonatiuh Rivero-Gutiérrez

    2012-01-01

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

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

    Science.gov (United States)

    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

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

    Science.gov (United States)

    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.

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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

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

    OpenAIRE

    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.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.)

  13. 3 MW TRIGA Research Reactor facility of BAEC and its Utilization

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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.)

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

    International Nuclear Information System (INIS)

    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

  19. 3 MW TRIGA Research Reactor facility of BAEC and its Utilization

    International Nuclear Information System (INIS)

    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)

  20. Utilization and operating experience of the TRIGA Mark II research reactor in Ljubljana

    Energy Technology Data Exchange (ETDEWEB)

    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

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

    International Nuclear Information System (INIS)

    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

  2. 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

    Energy Technology Data Exchange (ETDEWEB)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    Science.gov (United States)

    Viitanen, Tuomas; Leppänen, Jaakko

    2016-02-01

    A model of the FiR 1 Triga Mk-II reactor has been previously generated for the Serpent Monte Carlo reactor physics and burnup calculation code. In the current article, this model is validated by comparing the predicted reaction rates of nickel and manganese at 9 different positions in the reactor to measurements. In addition, track-length estimators are implemented in Serpent 2.1.18 to increase its performance in dosimetry calculations. The usage of the track-length estimators is found to decrease the reaction rate calculation times by a factor of 7-8 compared to the standard estimator type in Serpent, the collision estimators. The differences in the reaction rates between the calculation and the measurement are below 20%.

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

    NARCIS (Netherlands)

    Jacimovic, R; Maucec, M; Trkov, A

    2003-01-01

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

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

    NARCIS (Netherlands)

    Maucec, M

    2000-01-01

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

  7. 77 FR 68155 - The Armed Forces Radiobiology Research Institute TRIGA Reactor: Facility Operating License No. R-84

    Science.gov (United States)

    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....

  8. 76 FR 69296 - University of Utah, University of Utah TRIGA Nuclear Reactor, Notice of Issuance of Renewed...

    Science.gov (United States)

    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...

  9. Detection of a leaking boron-carbide control rod in a TRIGA Mark I reactor

    International Nuclear Information System (INIS)

    During a routine quarterly inspection of the boron-carbide control rods of the Omaha Veterans Administration Hospital 18 kW Triga Mark I reactor, a pin hole leak was detected approximately 3 mm from the chamfered edge. The leak was found by observing bubbles when the rod was withdrawn from the reactor tank for visual observation, and could not be seen with the naked eye. This suggests that pin hole leaks could occur and not be visually detected in control rods and fuel elements examined underwater. A review of the rod calibrations showed that the leak had not caused a loss in rod worth. Slides will be presented showing the bubbles observed during the inspection, together with an unmagnified and magnified view of the pin hole. (author)

  10. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    International Nuclear Information System (INIS)

    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 (233U), 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

  11. Reactivity measurements of the IPR-R1 TRIGA reactor fuel elements

    International Nuclear Information System (INIS)

    The thermal power of the IPR-R1 TRIGA reactor, belonging to the Centro de Desenvolvimento da Tecnologia Nuclear, will be upgraded from 100 k W to 250 k W. To attain this objective, mew additional fuel elements will be inserted in the reactor core. In order to provide information to the calculations of the new core arrangement, some fuel rods reactivity measurements were carried out as well as the determination of the reactivity increase due to the substitution of the present fuel by a new one. A first estimate indicates that the addition of 5 new fuel elements might be sufficient to reach the desired value of 3$ ρ excess. (author). 5 refs., 1 fig., 2 tabs

  12. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    Energy Technology Data Exchange (ETDEWEB)

    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.

  13. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    Science.gov (United States)

    Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

    2016-01-01

    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 (233U), 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.

  14. Present Services at the TRIGA Mark II Reactor of the JSI

    International Nuclear Information System (INIS)

    The TRIGA Mark II research reactor of the Jožef Stefan Institute has been continuously operating since the year 1966. The currently offered services include: (1) Neutron activation analysis in both instrumental and radiochemical modes; (2) neutron irradiation of various kinds of materials intended to be used for research and applicative purposes; (3) training and education of university students as well as on-job training of staff working in public and private institutions, (4) verification of computer codes and nuclear data, comprising primarily criticality calculations and neutron flux distribution studies and (5) testing and development of a digital reactivity meter. The paper briefly describes the aforementioned activities and shows that even such small reactors are still indispensable in nuclear science and technology. (author)

  15. Beam tube experiments and correlated research projects at the TRIGA reactor Vienna

    International Nuclear Information System (INIS)

    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.)

  16. Enhancement of mechanical properties of blended polyethylene radiation capsules for the TRIGA MARK II Research Reactor

    International Nuclear Information System (INIS)

    Mechanical properties of blended polyethylene (PE) containing the antioxidant Irganox 1010 and the UV-absorber Tinuvin 326 were studied for future use as radiation capsule material for the TRIGA Mark II research reactor. High density and low density polyethylene were blended with the additives and tested for elongation at break, impact strength and gel content, before and after irradiation inside the nuclear reactor. Characterization via FTIR as well as determination of crystallization and melt transition temperatures through DSC were also conducted. It was found that the addition of the antioxidant at different amounts (from 0 to 4 phr) had various effects on the properties of the blended PE, with 0 phr being the amount at which there was the biggest increase in elongation at break and impact strength, post-irradiation. (author)

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

    International Nuclear Information System (INIS)

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

  18. Thermal hydraulic analysis for the Oregon State TRIGA reactor using RELAP5-3D

    International Nuclear Information System (INIS)

    Thermal hydraulic analyses have being conducted at Oregon State University (OSU) in support of the conversion of the OSU TRIGA reactor (OSTR) core from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel as part of the Reduced Enrichment for Research and Test Reactors program. The goals of the thermal hydraulic analyses were to calculate natural circulation flow rates, coolant temperatures and fuel temperatures as a function of core power for both the HEU and LEU cores; calculate peak values of fuel temperature, cladding temperature, surface heat flux as well as departure from nuclear boiling ratio (DNBR) for steady state and pulse operation; and perform accident analyses for the accident scenarios identified in the OSTR safety analysis report. RELAP5-3D Version 2.4.2 was implemented to develop a model for the thermal hydraulic study. The OSTR core conversion is planned to take place in late 2008. (author)

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

    International Nuclear Information System (INIS)

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

  20. Evidence of possible flooding of the reflector at the Oregon State University TRIGA reactor

    International Nuclear Information System (INIS)

    There is evidence to indicate that water may have flooded the void spaces inside the aluminum cladding which surrounds the graphite reflector at the Oregon State University TRIGA reactor (OSTR). This paper presents and discusses this evidence, as well as data from a number of experiments which were performed in an attempt to confirm whether or not flooding actually has occurred. The safety implications of a flooded reflector are presented and shown to be of little concern based in part on the OSTR's history of high water purity in the reactor tank. It is concluded that the sole concern is a reduction in the usefulness of certain beam ports. Proposed methods of surveillance to detect corrosion of the reflector cladding are also briefly discussed. (author)

  1. Neutronic and thermal hydraulic analysis of the Geological Survey TRIGA Reactor

    Science.gov (United States)

    Shugart, Nicolas

    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 0-20 % void. The core water temperature reactivity coefficient is +0.012 /K. Following the neutronics analysis, the project developed RELAP5 and PARET-ANL models of the GSTR hot-rod fuel channel under steady state and transient conditions. The GSTR limiting core, determined as part of this analysis, provides a worst case operating scenario for the reactor. During steady state operations, the hot rod of the limiting core has a peak fuel temperature of 829 K and a minimum departure from nucleate boiling ratio of 2.16. After a $3.00 pulse reactivity insertion the fuel reaches a peak temperature is 1070 K. Examining the model results several seconds after a pulse reveals flow instabilities that result from weaknesses in the current two-channel model.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Costa, Antonella L., E-mail: lombardicosta@gmail.co [Departamento de Engenharia Nuclear - Escola de Engenharia da Universidade Federal de Minas Gerais, Av. Antonio Carlos, no 6627, Campus UFMG, PCA 1, CEP 31270-901, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil); Reis, Patricia Amelia L., E-mail: patricialire@yahoo.com.b [Departamento de Engenharia Nuclear - Escola de Engenharia da Universidade Federal de Minas Gerais, Av. Antonio Carlos, no 6627, Campus UFMG, PCA 1, CEP 31270-901, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil); Pereira, Claubia, E-mail: claubia@nuclear.ufmg.b [Departamento de Engenharia Nuclear - Escola de Engenharia da Universidade Federal de Minas Gerais, Av. Antonio Carlos, no 6627, Campus UFMG, PCA 1, CEP 31270-901, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil); Veloso, Maria Auxiliadora F., E-mail: dora@nuclear.ufmg.b [Departamento de Engenharia Nuclear - Escola de Engenharia da Universidade Federal de Minas Gerais, Av. Antonio Carlos, no 6627, Campus UFMG, PCA 1, CEP 31270-901, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil); Mesquita, Amir Z., E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear - CDTN/CNEN, Av. Antonio Carlos, 6627, Campus UFMG, Belo Horizonte (Brazil); Soares, Humberto V., E-mail: betovitor@ig.com.b [Departamento de Engenharia Nuclear - Escola de Engenharia da Universidade Federal de Minas Gerais, Av. Antonio Carlos, no 6627, Campus UFMG, PCA 1, CEP 31270-901, Belo Horizonte, MG (Brazil); Instituto Nacional de Ciencias e Tecnologia de Reatores Nucleares Inovadores/CNPq (Brazil)

    2010-06-15

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

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

    Science.gov (United States)

    Koehl, Michael A.

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

  5. Thermo-hydrodynamic design and safety parameter studies of the TRIGA MARK II research reactor

    International Nuclear Information System (INIS)

    The PARET computer code was used to analyse important thermo-hydrodynamic design and safety parameters of the 3 MW TRIGA MARK II research reactor at Atomic Energy Research Establishment (AERE), Savar, Dhaka, Bangladesh. The study involves the determination of the departure from nucleate boiling (DNB) value and studying its effect over the thermo-hydrodynamic design of the reactor. In the process the temperature profile, heat flux and pressure drop across the hottest channel of the TRIGA core were evaluated. The DNB ratio (DNBR), which is defined as the ratio of the critical heat flux to the heat flux achieved in the core, was computed by means of a suitable correlation as defined in PARET code. Over the length 0.381 m of the hottest channel the DNBR varies, starting from 3.8951 to 5.4031, with a minimum of 2.7851. The peak heat flux occurs at the axial centre of the fuel elements; therefore the DNBR is minimum at this location. The reactor core should be designed so as to prevent the DNBR from dropping below a chosen value under a high heat flux transient condition for the most adverse set of mechanical and coolant conditions. The loss-of-flow accident (LOFA) scenario of the reactor has also been studied to ensure that the existing design and procedures are adequate to assure that the consequences from this anticipated occurrence does not lead to a significant accident. The loss-of-flow transient after a trip time of 4.08 s at 85% of loss of normal flow for the TRIGA core shows a peak temperature of 709.22 deg. C in the fuel centreline and 131.94 deg. C in the clad and 46.63 deg. C in the coolant exit of the hottest channel. The transient was terminated at 15% of nominal flow after ∼48.0 s and the time at which the reversal of coolant flow starts is ∼67.0 s

  6. Extraction of neutron-rich fission products from a nuclear reactor for ground state studies: commissioning of the online-coupling at TRIGA-SPEC

    International Nuclear Information System (INIS)

    The mass spectrometer TRIGA-TRAP and the laser spectroscopy TRIGA-LASER setup, forming the TRIGA-SPEC experiment, are installed at the research reactor TRIGA Mainz in order to perform high-precision measurements of the ground state properties of short-lived neutron-rich radionuclides. Such measurements allow testing the predictive power of nuclear mass models and support astrophysical nucleosynthesis calculations. The extraction of these nuclei for both experiment branches is achieved by using an aerosol-based gas-jet system to transport fission products from an actinide target located inside the reactor to an external high-temperature surface ion source. TRIGA-SPEC will shortly go online, already having recorded a cyclotron resonance of an ion produced in the source. The commissioning of the online-coupling involving a separator magnet, a radiofrequency quadrupole cooler/buncher, and a pulsed drift tube will be presented.

  7. Simulation of the TRIGA-ININ reactor using EXT-2, in R-θ RΘ and temperature of 20 Centigrade

    International Nuclear Information System (INIS)

    The simulation of the TRIGA-ININ reactor, it was beginning considering the more simple case (follower bars equal to fuel elements, cell bar transitory with vacuum, etc.), this it left tuning as the obtained results were observed and it was studied the literature with respect to this reactor, in the following step the followers are considered as standard elements but with 32 grams of U-235 and so forth until reaching to the configuration that is considered definitive. (Author)

  8. Determination of the thermal-hydraulic parameters of ITU TRIGA Mark-II reactor

    International Nuclear Information System (INIS)

    In this study, a transient, one-dimensional thermal-hydraulic subchannel analysis for I.T.U. TRIGA Mark-II reactor was employed. Mixed convection is considered in modelling to enhance the capability of the computer code. After the continuity, conservation of energy, momentum balance equations for coolant in axial direction and the heat-conduction equation for the 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 the discretized continuity and momentum balance equations, a pressure and a pressure-correction equations were derived. Then, two different FORTRAN programs called TRIGATH (TRIGA Thermal-Hydraulics) and TRIGATH-R (TRIGATH Revised) have been developed to solve this set of algebraic equations by using SIMPLE and SIMPLER algorithms respectively. As a result, the temperature distributions of the coolant and the fuel rods as well as the velocity and pressure distributions of the coolant have been estimated for both transient and steady state regimes from both algorithms. Their results, which are in good agreement, are compared to the results of the computer code

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

    CERN Document Server

    Chiesa, Davide; 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 neutron fluxes obtained by MCNP5 to evaluate fuel consumption. This software was developed specifically to keep into account some features that differentiate experimental reactors from power ones, such as the daily ON/OFF cycle and the long fuel lifetime. These effects can not be neglected to properly account for neutron poison accumulation. We evaluated the effect of 48 years of reactor operation and predicted a possible new configuration for the reactor core: the objective was to remove some of the fuel elements from the...

  10. Periodic Safety review of JSI TRIGA Mark II and inspection of the reactor vessel

    Energy Technology Data Exchange (ETDEWEB)

    Jazbec, Anze; Snoj, Luak; Smodis, Borut [Jozef Stefan Institute, Ljubljana (Slovenia)

    2013-07-01

    Reactor TRIGA at the Jozef Stefan Institute (JSI) has been in operation since the year 1966. Most of the components were replaced during the process of maintenance and modernisation, but some of the equipment is still original or was replaced many years ago. Because of the ageing mechanisms, periodic safety review (PSR) is one of the crucial points for future utilization. According to legislation, PSR should be performed every 10 years. It features systematic inspection of structures, systems and components (SSC) of the reactor. Impacts of ageing, modernisation, operational experiences, technical progress, and changes of the site on the radiation and nuclear safety are verified. However, PSR does not give only inspection of the SSC, but also allows for the review of operating staff, their competence, operating procedures and other safety related procedures. PSR is pre-condition for extending operating licence. One of the components that have never been replaced is the aluminium reactor vessel. An externally contracted company made extensive analysis of the reactor vessel condition. Firstly, a visual inspection using underwater camera was made. Then all critical areas and welds were examined by using the ultrasound. Thickness of the wall was carefully measured and analysed. Using the same method, inspection for possible cracks inside aluminium was made. No failures were discovered and reactor vessel was found to be in a good condition.

  11. Long-lived activation products in TRIGA Mark II research reactor concrete shield: calculation and experiment

    Energy Technology Data Exchange (ETDEWEB)

    Zagar, Tomaz [Reactor Physics Department, Jozef Stefan Institute, Jamova 39, 1000 Ljubljana (Slovenia)]. E-mail: tomaz.zagar@ijs.si; Bozic, Matjaz [Nuklearna elektrarna Krsko, Vrbina 12, 8270 Krsko (Slovenia); Ravnik, Matjaz [Reactor Physics Department, Jozef Stefan Institute, Jamova 39, 1000 Ljubljana (Slovenia)

    2004-12-01

    In this paper, a process of long-lived activity determination in research reactor concrete shielding is presented. The described process is a combination of experiment and calculations. Samples of original heavy reactor concrete containing mineral barite were irradiated inside the reactor shielding to measure its long-lived induced radioactivity. The most active long-lived ({gamma} emitting) radioactive nuclides in the concrete were found to be {sup 133}Ba, {sup 60}Co and {sup 152}Eu. Neutron flux, activation rates and concrete activity were calculated for actual shield geometry for different irradiation and cooling times using TORT and ORIGEN codes. Experimental results of flux and activity measurements showed good agreement with the results of calculations. Volume of activated concrete waste after reactor decommissioning was estimated for particular case of Jozef Stefan Institute TRIGA reactor. It was observed that the clearance levels of some important long-lived isotopes typical for barite concrete (e.g. {sup 133}Ba, {sup 41}Ca) are not included in the IAEA and EU basic safety standards.

  12. Long-lived activation products in TRIGA Mark II research reactor concrete shield: calculation and experiment

    Science.gov (United States)

    Žagar, Tomaž; Božič, Matjaž; Ravnik, Matjaž

    2004-12-01

    In this paper, a process of long-lived activity determination in research reactor concrete shielding is presented. The described process is a combination of experiment and calculations. Samples of original heavy reactor concrete containing mineral barite were irradiated inside the reactor shielding to measure its long-lived induced radioactivity. The most active long-lived (γ emitting) radioactive nuclides in the concrete were found to be 133Ba, 60Co and 152Eu. Neutron flux, activation rates and concrete activity were calculated for actual shield geometry for different irradiation and cooling times using TORT and ORIGEN codes. Experimental results of flux and activity measurements showed good agreement with the results of calculations. Volume of activated concrete waste after reactor decommissioning was estimated for particular case of Jožef Stefan Institute TRIGA reactor. It was observed that the clearance levels of some important long-lived isotopes typical for barite concrete (e.g. 133Ba, 41Ca) are not included in the IAEA and EU basic safety standards.

  13. Data acquisition system for TRIGA Mark I nuclear reactor and a proposal for its automatic operation

    International Nuclear Information System (INIS)

    The TRIGA IPR-R1 Nuclear Research Reactor, located at the Nuclear Technology Development Center (CDTN/CNEN) in Belo Horizonte, Brazil, is being operated since 44 years ago. During these years the main operational parameters were monitored by analog recorders and counters located in the reactor control console. The most important operational parameters and data in the reactor logbook were registered by the reactor operators. This process is quite useful, but it can involve some human errors. It is also impossible for the operators to take notes of all variables involving the process mainly during fast power transients operations. A PC-based Data Acquisition was developed for the reactor that allows on line monitoring, through graphic interfaces, and shows operational parameters evolution to the operators. Some parameters that never were measured on line, like the thermal power and the coolant flow rate at the primary loop, are monitored now in the computer video monitor. The developed system allows measure out all parameters in a frequency up to 1 kHz. These data is also recorded in text files available for consults and analysis. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Garcia Martinez, H

    1999-07-01

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

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

    Directory of Open Access Journals (Sweden)

    H.P. Rahardjo

    2011-08-01

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

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

    International Nuclear Information System (INIS)

    Earthquakes in a nuclear installation can overload a piping system which is not flexible enough. These loads can be forces, moments and stresses working on the pipes or equipment. If the load is too large and exceed the allowable limits, the piping and equipment can be damaged and lead to overall system operation failure. The load received by piping systems can be reduced by making adequate piping flexibility, so all the loads can be transmitted homogeneously throughout the pipe without load concentration at certain point. In this research the analysis of piping stress has been conducted to determine the size of loads that occurred in the piping of primary cooling system of TRIGA 2000 Reactor, Bandung if an earthquake happened in the reactor site. The analysis was performed using Caesar II software-based finite element method. The ASME code B31.1 arranging the design of piping systems for power generating system (Power Piping Code) was used as reference analysis method. Modeling of piping systems was based on the cooling piping that has already been installed and the existing data reported in Safety Analysis Reports (SARs) of TRIGA 2000 reactor, Bandung. The quake considered in this analysis is the earthquake that occurred due to the Lembang fault, since it has the Peak Ground Acceleration (PGA) in the Bandung TRIGA 2000 reactor site. The analysis results showed that in the static condition for sustain and expansion loads, the stress fraction in all piping lines does not exceed the allowable limit. However, during operation moment, in dynamic condition, the primary cooling system is less flexible at sustain load, expansion load, and combination load and the stress fraction have reached 95,5%. Therefore a pipeline modification (re-routing) is needed to make pipe stress does not exceed the allowable stress. The pipeline modification was carried out by applied a gap of 3 mm in the X direction of the support at node 25 and eliminate the support at the node 30, also a

  18. European TRIGA owners' conference. Papers and abstracts

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

    Miley, G. H.

    1979-01-01

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

  20. Licensing of the TRIGA Mark III reactor at the Mexican Nuclear Centre

    International Nuclear Information System (INIS)

    The TRIGA Mark III reactor at the Mexican Nuclear Centre went critical in 1968 and remained so until 1979 when the National Commission for Nuclear Safety and Safeguards (CNSNS), the Mexican regulatory authority, was set up. The reactor was therefore operating without a formal operating license, and the CNSNS accordingly requested the ININ to license the reactor under the existing conditions and to ensure that any modification of the original design complied with Standards ANSI/ANS-15 and with the code of practice set out in IAEA Safety Series No. 35. The most relevant points in granting the operating licence were: (a) the preparation of the Safety Report; (b) the formulation and application of the Quality Assurance Programme; (c) the reconditioning of the following reactor systems: the cooling systems; the ventilation and exhaust system; the monitoring system and control panel; (d) the training of the reactor operating staff at junior and senior levels; and (e) the formulation of procedures and instructions. Once the provisional operating license was obtained for the reactor it was considered necessary to modify the reactor core, which has been composed of 20% enriched standards fuel, to a mixed core based on a mixture of standard fuel and FLIP-type fuel with 70% 235U enrichment. The CNSNS therefore requested that the mixed core be licensed and a technical report was accordingly annexed to the Safety Report, its contents including the following subjects: (a) neutron analysis of the proposed configuration; (b) reactor shutdown margins; (c) accident analysis; and (d) technical specifications. The licensing process was completed this year and we are now hoping to obtain the final operating license

  1. First shipment of TRIGA 14MW research reactor highly enriched uranium spent fuel to the United States of America

    International Nuclear Information System (INIS)

    The TRIGA 14MW Research Reactor has a unique design of core and fuel, with an exceptionally long life. This means long time in-core utilization, leading to a high burnup. The peculiar characteristics of the fuel and reactor facility design made the first shipment dissimilar from the other TRIGA reactors or aluminium plate type shipments. The paper presents the legal framework, regulatory activity, licensing, agreements, contracts, training prior to shipment. The shipment was considered a large coordinated project requiring preparatory activities, resources, national and international cooperation. The overall project time schedule is presented, as well as the diagram of the activities with intervening groups, organization and logistics, the unforeseen events being also mentioned. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hampel, Gabriele; Cordes, Harro [Medical University of Hannover, D-30625 Hannover (Germany); Ebbinghaus, Kurt; Haferkamp, Dirk [NOELL-KRC, D-97064 Wuerzburg (Germany)

    1998-07-01

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

  4. Neutron spectra in two beam ports of a TRIGA Mark III reactor with HEU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H. R.; Hernandez D, V. M. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Paredes G, L.; Aguilar, F., E-mail: fermineutron@yahoo.com [ININ, Carretera Mexico-Toluca s/n, Ocoyoacac 52750, Estado de Mexico (Mexico)

    2012-10-15

    Before to change the HEU for Leu fuel of the ININ's TRIGA Mark III nuclear reactor the neutron spectra were measured in two beam ports using 5 and 10 W. Measurements were carried out in a tangential and a radial beam port using a Bonner sphere spectrometer. It was found that neutron spectra are different in the beam ports, in radial beam port the amplitude of thermal and fast neutrons are approximately the same while, in the tangential beam port thermal neutron peak is dominant. In the radial beam port the fluence-to-ambient dose equivalent factors are 131{+-}11 and 124{+-}10 p Sv-cm{sup 2} for 5 and 10 W respectively while in the tangential beam port the fluence-to-ambient dose equivalent factor is 55{+-}4 p Sv-cm{sup 2} for 10 W. (Author)

  5. Operation and reactivity measurements of an accelerator driven subcritical TRIGA reactor

    Science.gov (United States)

    O'Kelly, David Sean

    Experiments were performed at the Nuclear Engineering Teaching Laboratory (NETL) in 2005 and 2006 in which a 20 MeV linear electron accelerator operating as a photoneutron source was coupled to the TRIGA (Training, Research, Isotope production, General Atomics) Mark II research reactor at the University of Texas at Austin (UT) to simulate the operation and characteristics of a full-scale accelerator driven subcritical system (ADSS). The experimental program provided a relatively low-cost substitute for the higher power and complexity of internationally proposed systems utilizing proton accelerators and spallation neutron sources for an advanced ADSS that may be used for the burning of high-level radioactive waste. Various instrumentation methods that permitted ADSS neutron flux monitoring in high gamma radiation fields were successfully explored and the data was used to evaluate the Stochastic Pulsed Feynman method for reactivity monitoring.

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

    Science.gov (United States)

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

    2012-10-01

    A 3-D model for 1 MW TRIGA Mark II research reactor was simulated. Neutron flux parameters were calculated using MCNP-4C code and were compared with experimental results obtained by k(0)-INAA and absolute method. The average values of φ(th),φ(epi), and φ(fast) by MCNP code were (2.19±0.03)×10(12) cm(-2)s(-1), (1.26±0.02)×10(11) cm(-2)s(-1) and (3.33±0.02)×10(10) cm(-2)s(-1), respectively. These average values were consistent with the experimental results obtained by k(0)-INAA. The findings show a good agreement between MCNP code results and experimental results. PMID:22885391

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-09-03

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

  8. Characterization of the neutron flux gradients in typical irradiation channels of a TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    The neutron distribution in a defined volume (gradient) for different matrices (air, water, cellulose, biological material and silicon dioxide) in two typical irradiation channels (pneumatic tube (PT) and IC40-channel in the carousel facility) in the TRIGA Mark II reactor at the Jozef Stefan Institute (IJS) was studied. Experiment was based on inserting Fe wires (flux monitors) into the chosen matrices. The wires were cut into small pieces after irradiation and the induced activities of 59Fe measured. The results showed that for the studied geometry the average spatial thermal neutron flux inhomogeneities (for five studied matrices) are about 2.3% in the PT-channel and about 2.9% in the IC40-channel. (author)

  9. Determination of α and f parameters at the 14-MW TRIGA reactor at Pitesti, Romania

    Science.gov (United States)

    Bărbos, D.; Păunoiu, C.; Roth, C.

    2010-10-01

    For experimental α determination the two-monitor method has been applied to determine α parameter in the irradiation channels at TRIGA 14 MW reactor (SCN Pitesti). The modified two-monitor method by using Cd ratio measurements eliminates the introducing of systematic errors due to the inaccuracy of absolute nuclear data. This characterization of the epithermal neutron spectrum is used in the k0-method of NAA, implemented at the SCN Pitesti. Neutron spectrum parameters were determined in the inner irradiation channel XC-1 and for outer irradiation channels: Beryllium J-6, Beryllium J-7, and Beryllium K-11. For α and f parameter verification a standard reference material denominated ECRM379-1 was analyzed using k0 standardization.

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

    Science.gov (United States)

    Wahida, Nurul; Yasir, Muhamad Samudi; Majid, Amran Ab; Wahab, Mohd Abd; Marzukee, Nik; Paulus, Wilfred; Phillip, Esther; Thanaletchumy, Irwan, M. N.

    2014-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Snoj, Luka; Ravnik, Matjaz; Lengar, Igor [Jozef Stefan Institute, Reactor Physics Division, Jamova 39, SI-1000 Ljubljana (Slovenia)

    2008-10-29

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

  12. Experimental study of the IPR-R1 TRIGA reactor power channels responses

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Henrique F.A.; Ferreira, Andrea V., E-mail: hfam@cdtn.br, E-mail: avf@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The IPR-R1 nuclear reactor installed at Centro de Desenvolvimento da Tecnologia Nuclear CDTN/CNEN, Belo Horizonte, Brazil, is a Mark I TRIGA reactor (Training, Research, Isotopes, General Atomics) and became operational on November of 1960. The reactor has four irradiation devices: a rotary specimen rack with 40 irradiation channels, the central tube, and two pneumatic transfer tubes. The nuclear reactor is operated in a power range between zero and 100 kW. The instrumentation for IPR-R1 operation is mainly composed of four neutronic channels for power measurements. The aim of this work is to investigate the responses of neutronic channels of IPR-R1, Linear, Log N and Percent Power channels, and to check their linearity. Gold foils were activated at low powers (0.125-1.000 kW), and cobalt foils were activated at high powers (10-100kW). For each sample irradiated at rotary specimen rack, another one was irradiated at the same time at the pneumatic transfer tube-2. The obtained results allowed evaluating the linearity of the neutronic channels responses. (author)

  13. Triga Mark III Reactor Operating Power and Neutron Flux Study by Nuclear Track Methodology

    Science.gov (United States)

    Espinosa, G.; Golzarri, J. I.; Raya-Arredondo, R.; Cruz-Galindo, S.; Sajo-Bohus, L.

    The operating power of a TRIGA Mark III reactor was studied using Nuclear Track Methodology (NTM). The facility has a Highly Enriched Uranium core that provides a neutron flux of around 2 x 1012 n cm-2 s-1 in the TO-2 irradiation channel. The detectors consisted of a Landauer® CR-39 (allyl diglycol polycarbonate) chip covered with a 3 mm Plexiglas® converter. After irradiation, the detectors were chemically etched in a 6.25M-KOH solution at 60±1 °C for 6 h. Track density was determined by a custom-made Digital Image Analysis System. The results show a direct proportionality between reactor power and average nuclear track density for powers in the range 0.1-7 kW. Data reproducibility and relatively low uncertainty (±3%) were achieved. NTM is a simple, fast and reliable technique that can serve as a complementary procedure to measure reactor operating power. It offers the possibility of calibrating the neutron flux density in any low power reactor.

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

    CERN Document Server

    Cammi, Antonio; 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 available experimental data as benchmark. In order to describe the full-power reactor configuration, the temperature distribution in the core is necessary. To evaluate it, a thermal-hydraulic model has been developed, using the power distribution results from MC simulation as input. The thermal-hydraulic model is focused on the core active region and takes into account sub-cooled boiling effects present at full reactor power. The obtained temperature distribution is then introduced in the MC model and a benchmark analysis is carr...

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

    International Nuclear Information System (INIS)

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

  16. Neutron radiography applications in I.T.U. TRIGA Mark-II reactor

    International Nuclear Information System (INIS)

    Neutron radiography is an important radiographic technique which is supplied different and advanced information according to the X or gamma ray radiography. However, it has a trouble for supplying the convenient neutron sources. Tangential beam tube of Istanbul Technical University (ITU) TRIGA Mark-II Training and Research Reactor has been arranged for using neutron radiography. The neutron radiography set defined as detailed for the application of the technique. Two different techniques for neutron radiography are defined as namely, transfer method and direct method. For the transfer method dysprosium and indium screens are used in the study. But, dysprosium generally was preferred in many studies in the point of view nuclear safety. Gadolinium was used for direct method. Two techniques are compared and explained the preferring of the transfer technique. Firstly, reference composition is prepared for seeing the differences between neutron and X-ray or gamma radiography. In addition of it, some radiograph samples are given neutron and X-ray radiography which shows the different image characters. Lastly, some examples are given from archaeometric studies. One of them the brass plates of Great Mosque door in Cizre. After the neutron radiography application, organic dye traces are noticed. Other study is on a sword that belong to Urartu period at the first millennium B.C. It is seen that some wooden part on it. Some different artefacts are examined with neutron radiography from the Ikiztepe excavation site, then some animal post parts are recognized on them. One of them is sword and sheath which are corroded together. After the neutron radiography application, it can be noticed that there are a cloth between the sword and its sheath. By using neutron radiography, many interesting and detailed results are observed in ITU TRIGA Mark-II Training and Research Reactor. Some of them shouldn't be recognised by using any other technique

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

    International Nuclear Information System (INIS)

    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 no up to a predetermined final level nf. 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)

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

    International Nuclear Information System (INIS)

    Nuclear Track Methodology (NTM), based on the neutron-proton interaction is one often employed alternative for neutron detection. In this paper we apply NTM to determine the Triga Mark III reactor operating power and neutron flux. The facility nuclear core, loaded with 85 Highly Enriched Uranium as fuel with control rods in a demineralized water pool, provide a neutron flux around 2 × 1012 n cm−2 s−1, at the irradiation channel TO-2. The neutron field is measured at this channel, using Landauer® PADC as neutron detection material, covered by 3 mm Plexiglas® as converter. After exposure, plastic detectors were chemically etched to make observable the formed latent tracks induced by proton recoils. The track density was determined by a custom made Digital Image Analysis System. The resulting average nuclear track density shows a direct proportionality response for reactor power in the range 0.1-7 kW. We indicate several advantages of the technique including the possibility to calibrate the neutron flux density measured at low reactor power

  19. Testing the applicability of the k0-NAA method at the MINT's TRIGA MARK II reactor

    Science.gov (United States)

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

    2006-08-01

    The Analytical Chemistry Laboratory at MINT is using the NAA technique since 1980s and is the only laboratory in Malaysia equipped with a research reactor, namely the TRIGA MARK II. Throughout the years the development of NAA technique has been very encouraging and was made applicable to a wide range of samples. At present, the k0 method has become the preferred standardization method of NAA ( k0-NAA) due to its multi-elemental analysis capability without using standards. Additionally, the k0 method describes NAA in physically and mathematically understandable definitions and is very suitable for computer evaluation. Eventually, the k0-NAA method has been adopted by MINT in 2003, in collaboration with the Nuclear Research Institute (NRI), Vietnam. The reactor neutron parameters ( α and f) for the pneumatic transfer system and for the rotary rack at various locations, as well as the detector efficiencies were determined. After calibration of the reactor and the detectors, the implemented k0 method was validated by analyzing some certified reference materials (including IAEA Soil 7, NIST 1633a, NIST 1632c, NIST 1646a and IAEA 140/TM). The analysis results of the CRMs showed an average u score well below the threshold value of 2 with a precision of better than ±10% for most of the elemental concentrations obtained, validating herewith the introduction of the k0-NAA method at the MINT.

  20. Measurement of the power and temperature reactivity coefficients of the RTP TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rabir, Mohamad Hairie, E-mail: m_hairie@nuclearmalaysia.gov.my

    2013-12-15

    This paper presents the experimental results of the power and temperature coefficients of reactivity of the RTP TRIGA reactor at the Malaysian Nuclear Agency. The power coefficient of reactivity obtained was approximately −0.26 ¢ kW{sup −1} (−1.81 × 10{sup −5} kW{sup –1}), and the measured temperature reactivity coefficient of the reactor was −0.82 ¢ °C{sup −1} (−5.77 × 10{sup −5} °C{sup −1}) and −1.15 ¢ °C{sup −1} (−8.08 × 10{sup −5} °C{sup −1}) in IFE C12 and IFE F16, respectively. The power defect, which is the change in reactivity taking place between zero power and the power of 850 kW was ∼2.19 $. Because of the negative temperature coefficient, a significant amount of reactivity is needed to compensate for the temperature change and allows the reactor to operate at the higher power levels in steady state. Throughout this experiment, it is the temperature of the fuel that was measured, not the isothermal temperature coefficient (ITC), which comprises both moderator and fuel.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-10-15

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

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

    Science.gov (United States)

    Espinosa, G.; Golzarri, J. I.; Raya-Arredondo, R.; Cruz-Galindo, S.; Sajo-Bohus, L.

    2015-07-01

    Nuclear Track Methodology (NTM), based on the neutron-proton interaction is one often employed alternative for neutron detection. In this paper we apply NTM to determine the Triga Mark III reactor operating power and neutron flux. The facility nuclear core, loaded with 85 Highly Enriched Uranium as fuel with control rods in a demineralized water pool, provide a neutron flux around 2 × 1012 n cm-2 s-1, at the irradiation channel TO-2. The neutron field is measured at this channel, using Landauer® PADC as neutron detection material, covered by 3 mm Plexiglas® as converter. After exposure, plastic detectors were chemically etched to make observable the formed latent tracks induced by proton recoils. The track density was determined by a custom made Digital Image Analysis System. The resulting average nuclear track density shows a direct proportionality response for reactor power in the range 0.1-7 kW. We indicate several advantages of the technique including the possibility to calibrate the neutron flux density measured at low reactor power.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-23

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

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

    International Nuclear Information System (INIS)

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

  5. Thermal Hydraulics Analysis for the 3MW TRIGA MARK-II Research Reactor Under Transient Condition

    International Nuclear Information System (INIS)

    Some important thermal hydraulic parameters of the 3 MW TRIGA MARK-II research reactor operating under transient condition were investigated using two computer codes PULTRI and TEMPUL. Major transient parameters, such as, peak power and prompt energy released after pulse, maximum fuel and coolant temperature, surface heat flux, time and radial distribution of temperature within fuel element after pulse, fuel, fuel-cladding gap width variation, etc. were computer and compared with the experimental and operational values as reported in the safety Analysis Report (SAR). It was observed that pulsing of the reactor inserting an excess reactivity of $2.00 shoots the reactor power level to 854.353 MW compared to an experimental value of 852 MW; the maximum fuel temperature corresponding to this peak power was found to be 846.76o C which is much less than the limiting maximum value of fuel temperature of 11500 C as reported in SAR. During a pulse if the film boiling occurs for a peak adiabatic fuel temperature of 1000o C, the calculated outer cladding wall temperature was observed to be 702.390 C compared to a value of 760o C reported in SAR under the same condition. The investigated other results were also found to be in good agreement with the values reported in the SAR. 16 refs., 22 figs. (author)

  6. Identification of a leaking TRIGA fuel element at the reactor facility of Pavia

    International Nuclear Information System (INIS)

    On January 28th 2004, during a periodical activity of characterization of the ionic-exchange resins of the demineralizer of the primary cooling circuit of the TRIGA Mark II reactor of the University of Pavia a small but detectable amount of 137Cs contamination was measured. Since the reactor has been running for several hundreds of hours at full power without showing any anomaly in the radiometric and thermo-hydraulic parameters, the reactor was brought at the nominal power of 250 kW for one hour and a sample of water was collected from the reactor tank and analyzed in a low background gamma-ray detector. As a result a small amount of fission products were detected in the reactor pool water (few Bq/g) suggesting the existence of a possible clad defect in one ore more fuel elements. As a consequence of this situation a campaign of gamma-ray spectrometry was implemented in order to evaluate the importance of the release. Analyzes using a HGe detector (1.72 keV FWHM - 31.3 % efficiency - 58.5 Photo Peak/Compton) were performed and the most significant results are presented as well as the identification of the leaking fuel element. The fission products leakage was due to a micro-fissure of a fuel element that released only noble gas when it was heated up to a temperature around 90oC , i.e. at the reactor power of about 100 kW. The oldest SST clad instrumented fuel element in the core was identified as the origin of the release. It was removed from its position and stored in a rack of the reactor pool under 4 m of water shield. The reactor came back in regular operation on March 22nd 2004 and no other fission products leakages were detected. After this situation the reactor pool water is sampled and measured with a low-background gamma-ray detector every month before the reactor start-up and after one hour of operation of the reactor at full nominal power. (nevyjel)

  7. Assessing Information Literacy among Undergraduates: A Discussion of the Literature and the University of California-Berkeley Assessment Experience.

    Science.gov (United States)

    Maughan, Patricia Davitt

    2001-01-01

    Since 1994, the Teaching Library at the University of California-Berkeley has conducted an ongoing Survey of Information Literacy Competencies in selected academic departments to measure the "lower-order" information literacy skills of graduating seniors. The survey reveals that students think they know more about accessing information and…

  8. Extraction of pure thermal neutron beam for the proposed PGNAA facility at the TRIGA research reactor of AERE, Savar, Bangladesh

    Science.gov (United States)

    Alam, Sabina; Zaman, M. A.; Islam, S. M. A.; Ahsan, M. H.

    1993-10-01

    A study on collimators and filters for the design of a spectrometer for prompt gamma neutron activation analysis (PGNAA) at one of the radial beamports of the TRIGA Mark II reactor at AERE, Savar has been carried out. On the basis of this study a collimator and a filter have been designed for the proposed PGNAA facility. Calculations have been done for measuring neutron flux at various positions of the core of the reactor using the computer code TRIGAP. Gamma dose in the core of the reactor has also been measured experimentally using TLD technique in the present work.

  9. Extraction of pure thermal neutron beam for the proposed PGNAA facility at the TRIGA research reactor of AERE, Savar, Bangladesh

    Energy Technology Data Exchange (ETDEWEB)

    Alam, S. (Physics Dept., Jahangirnagar Univ., Savar, Dhaka (Bangladesh)); Zaman, M.A. (Physics Dept., Jahangirnagar Univ., Savar, Dhaka (Bangladesh)); Islam, S.M.A. (Physics Dept., Jahangirnagar Univ., Savar, Dhaka (Bangladesh)); Ahsan, M.H. (Inst. of Nuclear Science and Technology (INST), AERE, Savar, Dhaka (Bangladesh))

    1993-10-01

    A study on collimators and filters for the design of a spectrometer for prompt gamma neutron activation analysis (PGNAA) at one of the radial beamports of the TRIGA Mark II reactor at AERE, Savar has been carried out. On the basis of this study a collimator and a filter have been designed for the proposed PGNAA facility. Calculations have been done for measuring neutron flux at various positions of the core of the reactor using the computer code TRIGAP. Gamma dose in the core of the reactor has also been measured experimentally using TLD technique in the present work. (orig.)

  10. Extraction of pure thermal neutron beam for the proposed PGNAA facility at the TRIGA research reactor of AERE, Savar, Bangladesh

    International Nuclear Information System (INIS)

    A study on collimators and filters for the design of a spectrometer for prompt gamma neutron activation analysis (PGNAA) at one of the radial beamports of the TRIGA Mark II reactor at AERE, Savar has been carried out. On the basis of this study a collimator and a filter have been designed for the proposed PGNAA facility. Calculations have been done for measuring neutron flux at various positions of the core of the reactor using the computer code TRIGAP. Gamma dose in the core of the reactor has also been measured experimentally using TLD technique in the present work. (orig.)

  11. The use of beam neutron of TRIGA IPR-R1 (Mark 1) reactor for general applications

    International Nuclear Information System (INIS)

    At present, there are four devices in the TRIGA IPR-R1 reactor at the CDTN for sample irradiation, but in these irradiators the mass and form of the sample are limited to the standardized dimensions of the irradiation receivers. Besides, the irradiation is made under, approximately, 5 meters of water, complicating the access. However, through an beam neutron extractor arrangement, it is possible to irradiate larger samples, in a local more accessible and with minimum interference of fast neutrons facilitating to measure neutronic parameters, to do crystals neutron diffraction, to obtain neutron radiographs, among other applications. This work presents results of the experimental Neutron Extractor arrangement in TRIGA reactor at CDTN. (author)

  12. The Neutronic And Power Distribution Calculations For Triga 2 MW Reactor Using WIMS-D/4 And Citation Codes

    International Nuclear Information System (INIS)

    . The neutronic calculation has been carried out for TRIGA 2 MW reactor. These included criticality flux and power distributions. Computer code Citation which solves 7-groups, 3-dimensional hexagonal geometry has been used. The multi groups-cross-section is generated by the WIMS-D/4 code.This 7-group-39x39x38-mesh-points problem takes about 90 minutes on the Pentium-133 MHz PC. The calculation of the initial core of TRIGA 2 MW reactor shows that the excess reactivity of the core is 7,8% and the thermal fluxes in the irradiation positions are between 1.0-2.9*1013n cm-2s-1. The results are about 10% deviate from those calculated by General Atomics. In the initial core, the highest power is produced in the C-9 position. The fuel element in this position produces 30.7 k W thermal power

  13. Neutronic tests and reactivity balance in the IPR-R1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Rose Mary Gomes do Prado; Souza, Luiz Claudio Andrade, E-mail: souzarm@cdtn.br, E-mail: lcas@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    This paper presents the 2014 neutronic tests performed on CDTN's TRIGA IPR-R1 research reactor. Such tests are performed annually, as prescribed by the Safety Analysis Report. The three control rods, Regulating, Shim and Safety, were calibrated and their worth determined to be 0.52 $, 3.08 $ and 2.78 $, respectively. The Shim rod takes 0.44 s to shutdown the reactor and the Safety rod 0.48 s. The maximum reactivity insertion rates are 48 pcm/s by the Shim rod and 46 pcm/s by the Safety rod. Total reactivity excess is 1.88 $. The temperature reactivity coefficient determined is -0.94 cent/deg C. A reactivity insertion of 0.71 $ is necessary in order to achieve the licensed maximum reactor power of 100 kW. Reactivity losses due to xenon poisoning, after operating for 8 h at maximum power, is 0.20 $, and the insertion of a void tube in the Central Thimble corresponds to 0.22 $. A significant amount of reactivity is required to overcome the temperature effect and allow the reactor to operate at full power for extended periods of time. Given all these reactivity losses, a new fuel element should soon be added to the core in order to increase the reactivity excess. Adding this new fuel element to the C ring and moving the element withdrawn from that position to the F ring, replacing a graphite dummy element, would increase 45.5 cents in the reactivity excess worth. Calculations and experimental results will be used to optimize a new core configuration for the reactor. (author)

  14. Characterization of control rod worths and fuel rod power peaking factors in the university of Utah TRIGA Mark I reactor

    OpenAIRE

    Alroumi Fawaz; Kim Donghoon; Schow Ryan; Jevremovic Tatjana

    2016-01-01

    Control rod reactivity (worths) for the three control rods and fuel rod power peaking factors in the University of Utah research reactor (100 kW TRIGA Mark I) are characterized using the AGENT code system and the results described in this paper. These values are compared to the MCNP6 and existing experimental measurements. In addition, the eigenvalue, neutron spatial flux distributions and reaction rates are analyzed and discussed. The AGENT code system is ...

  15. Present and future beam tube experiments at the 250 kW TRIGA Mark II reactor Wien

    International Nuclear Information System (INIS)

    The four beam tubes and the thermal column at the TRIGA reactor Wien were well used in the reporting period. Since the thermal column is used as a gamma source for different irradiation experiments and as a neutron source for radiography, the other facilities are mainly used for neutron spectroscopy experiments: polarized neutrons, neutron interferometry, small angle scattering and neutron choppers, In the piercing beam tube a fast rabbit system is installed which is mainly used for high precision activation analysis. (author)

  16. Utilization of a TRIGA nuclear reactor for neutron activation analysis and isotope production

    International Nuclear Information System (INIS)

    In this paper, some examples of utilization of the TRIGA reactor at Casaccia research center of ENEA are described, including neutron activation analysis (NAA), isotope production and tritium release studies. 1) NAA - Determination of trace elements by NAA has been performed, in the last years upon a variety of matrices: a) environmental: suspended particulate matter (SPM) and sediments collected in some Italian seas and rivers were analyzed for about 30 elements. Different types of filters were also tested to choose the most suitable for the collection of SPM. b) forensic: many applications of NAA have been performed on request of Italian Courts for determination of gunshot residues; firing distances were also determined in some cases. c) agricultural: the uptake of Zn and Co by cereals has been studied in pot in co-operation with the Istituto per la Cerealicoltura of Rome. d) geological: some USGS reference materials were analyzed by thermal and epithermal NAA, for evaluating accuracy and precision of both methods. Rock samples from the basaltic plateau of Kenya were then analyzed by ENAA, chiefly for rare earth elements, whose concentration patterns can give useful informations about petrogenesis. e) reference materials: several elements have been determined in lake and river sediment samples and in three plant materials (tobacco, apple-tree and peach-tree), in order to provide data for their certification as reference materials to be used in different fields. f) nuclear materials: trace elements have been determined in LiAlO2 to be used in the blanket of fusion reactors, as well as in cements for building nuclear plants, in order to evaluate the activity at the time of plant decommissioning. 2) Isotope production a) A program for the utilization of TRIGA reactor at Casaccia to prepare a 191m-Ir generator to be used in pediatric angiography has been drawn up. b) Preparation of 18-F to be used in positron emission tomography (PET) for studies of cerebral diseases

  17. Enhancement in utilization of TRIGA 14 MW research reactor operated by the Institute for Nuclear Research in Pitesti

    International Nuclear Information System (INIS)

    This paper refers to the types of projects and services that have been provided in the past by the Pitesti TRIGA research reactor. Many of them are current capabilities. It will be of particular benefit to increase the utilization of research reactor by developing new techniques and methods for appropriate applications. New projects are developed in the field of materials science as well as studies on nuclear safety in CANDU fuels and particularly in the new type of fuel with slightly enriched uranium recovered from PWR irradiated fuel. TRIGA reactors from INR Pitesti was extensively used for materials testing, irradiation servicing for fuels and materials, radioisotope production, reactor engineering experiments in beam experiments, development of nuclear methods for various research purposes. The irradiation test together with post-irradiation examination on fuels and materials contributed to evaluation of their integrity and also lead to development of advanced fuels and materials. Better results were obtained due to the upgrading of the steady state reactor power from 14 MW to 21 MW accompanied by upgrading of certain reactor systems, what resulted in an increase of the neutron flux densities. It enhanced the performances of both irradiation experiments and reactor operation. (author)

  18. Irradiation tests on PHWR type fuel elements in TRIGA research reactor of INR Pitesti

    International Nuclear Information System (INIS)

    Nine PHWR type fuel elements with reduced length were irradiated in loop A of the TRIGA Research Reactor of INR Pitesti. The primary objective of the test was to determine the performance of nuclear fuel fabricated at INR Pitesti at high linear powers in pressurized water conditions. Six fuel elements were irradiated with a ramp power history, achieving a maximum power of 45 kW/m during pre-ramp and of 64 kW/m in the ramp. The maximum discharge burnup was of 216 MWh/kgU. Another three fuel elements with reduced length were irradiated with declining power history. At the beginning of irradiation the fuel elements achieved a maximum linear power of 66 kW/m. The maximum fuel power was about 1.3 times the maximum expected in PHWR. The maximum discharge burnup was 205 MWh/kgU. The elements were destructively examined in the hot cells of INR Pitesti. Temperature-sensitive parameters such as UO2 grain growth, fission-gas release and sheath deformations were examined. The tests proved the feasibility of irradiating PHWR type fuel elements at linear powers up to 66 kW/m under pressurized water conditions and demonstrated the possibility of more flexible operation of this fuel in power reactors. This paper presents the results of the investigation. (orig.)

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

    International Nuclear Information System (INIS)

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

  20. Design of sample carrier for neutron irradiation facility at TRIGA MARK II nuclear reactor

    Science.gov (United States)

    Abdullah, Y.; Hamid, N. A.; Mansor, M. A.; Ahmad, M. H. A. R. M.; Yusof, M. R.; Yazid, H.; Mohamed, A. A.

    2013-06-01

    The objective of this work is to design a sample carrier for neutron irradiation experiment at beam ports of research nuclear reactor, the Reaktor TRIGA PUSPATI (RTP). The sample carrier was designed so that irradiation experiment can be performed safely by researchers. This development will resolve the transferring of sample issues faced by the researchers at the facility when performing neutron irradiation studies. The function of sample carrier is to ensure the sample for the irradiation process can be transferred into and out from the beam port of the reactor safely and effectively. The design model used was House of Quality Method (HOQ) which is usually used for developing specifications for product and develop numerical target to work towards and determining how well we can meet up to the needs. The chosen sample carrier (product) consists of cylindrical casing shape with hydraulic cylinders transportation method. The sample placing can be done manually, locomotion was by wheel while shielding used was made of boron materials. The sample carrier design can shield thermal neutron during irradiation of sample so that only low fluencies fast neutron irradiates the sample.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  4. Design of sample carrier for neutron irradiation facility at TRIGA MARK II nuclear reactor

    International Nuclear Information System (INIS)

    The objective of this work is to design a sample carrier for neutron irradiation experiment at beam ports of research nuclear reactor, the Reaktor TRIGA PUSPATI (RTP). The sample carrier was designed so that irradiation experiment can be performed safely by researchers. This development will resolve the transferring of sample issues faced by the researchers at the facility when performing neutron irradiation studies. The function of sample carrier is to ensure the sample for the irradiation process can be transferred into and out from the beam port of the reactor safely and effectively. The design model used was House of Quality Method (HOQ) which is usually used for developing specifications for product and develop numerical target to work towards and determining how well we can meet up to the needs. The chosen sample carrier (product) consists of cylindrical casing shape with hydraulic cylinders transportation method. The sample placing can be done manually, locomotion was by wheel while shielding used was made of boron materials. The sample carrier design can shield thermal neutron during irradiation of sample so that only low fluencies fast neutron irradiates the sample.

  5. Confirmation of a realistic reactor model for BNCT dosimetry at the TRIGA Mainz

    Energy Technology Data Exchange (ETDEWEB)

    Ziegner, Markus, E-mail: Markus.Ziegner.fl@ait.ac.at [AIT Austrian Institute of Technology GmbH, Vienna A-1220, Austria and Institute of Atomic and Subatomic Physics, Vienna University of Technology, Vienna A-1020 (Austria); Schmitz, Tobias; Hampel, Gabriele [Institut für Kernchemie, Johannes Gutenberg-Universität, Mainz DE-55128 (Germany); Khan, Rustam [Pakistan Institute of Engineering and Applied Sciences (PIEAS), Islamabad PK-44000 (Pakistan); Blaickner, Matthias [AIT Austrian Institute of Technology GmbH, Vienna A-1220 (Austria); Palmans, Hugo [Acoustics and Ionising Radiation Division, National Physical Laboratory, Teddington TW11 0LW, United Kingdom and Medical Physics Group, EBG MedAustron GmbH, Wiener Neustadt A-2700 (Austria); Sharpe, Peter [Acoustics and Ionising Radiation Division, National Physical Laboratory, Teddington TW11 0LW (United Kingdom); Böck, Helmuth [Institute of Atomic and Subatomic Physics, Vienna University of Technology, Vienna A-1020 (Austria)

    2014-11-01

    Purpose: In order to build up a reliable dose monitoring system for boron neutron capture therapy (BNCT) applications at the TRIGA reactor in Mainz, a computer model for the entire reactor was established, simulating the radiation field by means of the Monte Carlo method. The impact of different source definition techniques was compared and the model was validated by experimental fluence and dose determinations. Methods: The depletion calculation code ORIGEN2 was used to compute the burn-up and relevant material composition of each burned fuel element from the day of first reactor operation to its current core. The material composition of the current core was used in a MCNP5 model of the initial core developed earlier. To perform calculations for the region outside the reactor core, the model was expanded to include the thermal column and compared with the previously established ATTILA model. Subsequently, the computational model is simplified in order to reduce the calculation time. Both simulation models are validated by experiments with different setups using alanine dosimetry and gold activation measurements with two different types of phantoms. Results: The MCNP5 simulated neutron spectrum and source strength are found to be in good agreement with the previous ATTILA model whereas the photon production is much lower. Both MCNP5 simulation models predict all experimental dose values with an accuracy of about 5%. The simulations reveal that a Teflon environment favorably reduces the gamma dose component as compared to a polymethyl methacrylate phantom. Conclusions: A computer model for BNCT dosimetry was established, allowing the prediction of dosimetric quantities without further calibration and within a reasonable computation time for clinical applications. The good agreement between the MCNP5 simulations and experiments demonstrates that the ATTILA model overestimates the gamma dose contribution. The detailed model can be used for the planning of structural

  6. Implementation of the k{sub 0} technique using multi-detectors on diverse irradiation facilities of TRIGA Reactor; Implementacion de la tecnica k{sub 0} usando multidetectores en diferentes instalaciones de irradiacion del Reactor TRIGA

    Energy Technology Data Exchange (ETDEWEB)

    Caldera C, M. de G.

    2013-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

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

    Science.gov (United States)

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

    2016-10-01

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

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

    Science.gov (United States)

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

    2016-10-01

    This study was conducted to improve the capacity of radioisotope production in the Moroccan TRIGA Mark II research reactor, which is considered as one of the most important applications of research reactors. The aim of this study is to enhance the utilization of TRIGA core in the field of neutron activation and ensure an economic use of the fuel. The main idea was to create an additional irradiation channel (IC) inside the core. For this purpose, three new core configurations are proposed, which differ according to the IC position in the core. Thermal neutron flux distribution and other neutronic safety parameters such as power peaking factors, excess reactivity, and control rods worth reactivity were calculated using the Monte Carlo N-Particle Transport (MCNP) code and neutron cross-section library based on ENDF/B-VII evaluation. The calculated thermal flux in the central thimble (CT) and in the added IC for the reconfigured core is compared with the thermal flux in the CT of the existing core, which is taken as a reference. The results show that all the obtained fluxes in CTs are very close to the reference value, while a remarkable difference is observed between the fluxes in the new ICs and reference. This difference depends on the position of IC in the reactor core. To demonstrate that the Moroccan TRIGA reactor could safely operate at 2MW, with new configurations based on new ICs, different safety-related thermal-hydraulic parameters were investigated. The PARET model was used in this study to verify whether the safety margins are met despite the new modifications of the core. The results show that it is possible to introduce new ICs safely in the reactor core, because the obtained values of the parameters are largely far from compromising the safety of the reactor. PMID:27552124

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

    International Nuclear Information System (INIS)

    The k0 method with the technique of neutron activation analysis allows obtaining important characteristics parameters that describe a nuclear reactor. Among these parameters are the form factor of epithermal neutron flux, α and the ratio of thermal neutron flux with respect to the epithermal neutron flux, f. These parameters were obtained by irradiation of two different monitors, one of Au-Zr and the other of Au-Mo-Cr, where the last one was made and implemented for the first time. Both monitors were irradiated in different positions in the TRIGA Mark III Reactor at the National Institute of Nuclear Research. (Author)

  11. An analysis of LEU fuel behavior as compared to HEU fuel in the 14 MW TRIGA SSR reactor

    International Nuclear Information System (INIS)

    The paper presents an analysis of the behavior and properties of the fuel loading the 14 MW TRIGA research reactor. Comparison are made between the original highly enriched fuel and the slightly enriched fuel which is loading the reactor at present. Both the highly enriched and the slightly enriched fuels have the same physical and thermal properties but different nuclear properties. Thermal hydraulic analysis of the transient regime behavior of reactivity insertion type was effected, evidencing the different behavior of the fuels with different enrichments. (authors)

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

    International Nuclear Information System (INIS)

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

  13. Determination of the parameters α and f of the reactor triga mark II of Cren-K

    International Nuclear Information System (INIS)

    The α parameter and the thermal to epithermal flux ratio (f) have been determined for irradiation channels (LS3, LS5, LS9, LS11 and LS25) of nuclear reactor Triga Mark II of the Regional Nuclear Centre of Kinshasa. The three methods - Cd radio method covered Cd monitor method and base monitor method used for evaluation of α parameter give the same result in each irradiation position. The thermal to epithermal flux ratio f has been determined by the Cd ratio method. Results show that nuclear parameters α and f, change from one point to another of the reactor; α being negative, the resonance integrals I.(α) are increased.

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

    Science.gov (United States)

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

    2011-11-01

    The activation of the biological shield is an important process for waste management considerations of nuclear facilities. The final activity can be estimated by modeling using the neutron flux density rather than the radiometric approach of activity measurements. Measurement series at the TRIGA reactor Vienna reveal that the flux density next to the biological shield is in the order of 10(9)cm(-2)s(-1) at maximum power; but it is strongly influenced by reactor installations. The data allow the estimation of the final waste categorization of the concrete according to the Austrian legislation. PMID:21646026

  15. Analysis of an extreme loss of coolant in the IPR-R1 TRIGA reactor using a RELAP5 model

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Patricia Amelia de Lima; Costa, Antonella Lombardi; Pereira, Claubia; Veloso, Maria Auxiliadora F.; Soares, Humberto Vitor, E-mail: patricialire@yahoo.com.br, E-mail: antonella@nuclear.ufmg.br, E-mail: claubia@nuclear.ufmg.br, E-mail: dora@nuclear.ufmg.br, E-mail: hvs@cdtn.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil). Departamento de Engenharia Nuclear; Mesquita, Amir Zacharias, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2012-07-01

    The RELAP5/MOD3.3 code has been applied for thermal hydraulic analysis of power reactors as well as nuclear research reactors with good predictions. The development and the assessment of a RELAP5 model for the IPRR1 TRIGA have been validated for steady state and transient situations. 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. In this work, an extreme transient case of loss of coolant accident (LOCA) has been simulated. For this type of analysis, the automatic scram of the reactor was not considered because the main aim was to verify the evolution of the fuel elements heating in the absence of coolant. The temperature evolutions are presented as well as an analysis about the temperature safety limits. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Clarysson A.M. da, E-mail: clarysson_silva@yahoo.com.br [Departamento de Engenharia Nuclear - Escola de Engenharia, Universidade Federal de Minas Gerais, Av. Presidente Antonio Carlos, 6627, 31270-901 Campus Pampulha - Belo Horizonte (Brazil); Pereira, Claubia, E-mail: claubia@nuclear.ufmg.br [Departamento de Engenharia Nuclear - Escola de Engenharia, Universidade Federal de Minas Gerais, Av. Presidente Antonio Carlos, 6627, 31270-901 Campus Pampulha - Belo Horizonte (Brazil); Guerra, Bruno T., E-mail: brunoteixeiraguerra@yahoo.com.br [Departamento de Engenharia Nuclear - Escola de Engenharia, Universidade Federal de Minas Gerais, Av. Presidente Antonio Carlos, 6627, 31270-901 Campus Pampulha - Belo Horizonte (Brazil); Veloso, Maria Auxiliadora F., E-mail: dora@nuclear.ufmg.br [Departamento de Engenharia Nuclear - Escola de Engenharia, Universidade Federal de Minas Gerais, Av. Presidente Antonio Carlos, 6627, 31270-901 Campus Pampulha - Belo Horizonte (Brazil); Costa, Antonella L., E-mail: lombardicosta@gmail.com [Departamento de Engenharia Nuclear - Escola de Engenharia, Universidade Federal de Minas Gerais, Av. Presidente Antonio Carlos, 6627, 31270-901 Campus Pampulha - Belo Horizonte (Brazil); Dalle, Hugo M., E-mail: dallehm@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear, Comissao Nacional de Energia Nuclear, Campus da UFMG - Av. Presidente Antonio Carlos, 6627, 31270-901, P.O. Box: 941, Belo Horizonte, MG (Brazil)

    2011-12-15

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Uddin, M.N. [Department of Physics, Jahangirnagar University, Dhaka (Bangladesh); Sarker, M.M., E-mail: sarker_md@yahoo.co [Reactor Physics and Engineering Division, Institute of Nuclear Science and Technology, Atomic Energy Research Establishment, Ganakbari, Savar, GPO Box 3787, Dhaka-1000 (Bangladesh); Khan, M.J.H. [Reactor Physics and Engineering Division, Institute of Nuclear Science and Technology, Atomic Energy Research Establishment, Ganakbari, Savar, GPO Box 3787, Dhaka-1000 (Bangladesh); Islam, S.M.A. [Department of Physics, Jahangirnagar University, Dhaka (Bangladesh)

    2010-03-15

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

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

    International Nuclear Information System (INIS)

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

  1. Burnup analysis and in-core fuel management study of the 3 MW TRIGA MARK II research reactor

    International Nuclear Information System (INIS)

    The principal objective of this study is to formulate an effective optimal fuel management strategy for the TRIGA MARK II research reactor at AERE, Savar. The core management study has been performed by utilizing four basic types of information calculated for the reactor: criticality, power peaking, neutron flux and burnup calculation. This paper presents the results of the burnup calculations for TRIGA LEU fuel elements. The fuel element burnup for approximately 20 years of operation was calculated using the TRIGAP compute code. The calculation is performed in one-dimensional radial geometry in TRIGAP. Inter-comparison of TRIGAP results with other two calculations performed by MVP-BURN and MCNP4C-ORIGEN2.1 show very good agreement. Reshuffling at 20,000 MWh step provides the highest core lifetime of the reactor, which is 64,500 MWh. Besides, the study gives valuable insight into the behaviour of the reactor and will ensure better utilization and operation of the reactor in future

  2. Simulation of Collimator for Neutron Imaging Facility of TRIGA MARK II PUSPATI Reactor

    Science.gov (United States)

    Zin, Muhammad Rawi Mohamed; Jamro, Rafhayudi; Yazid, Khairiah; Hussain, Hishamuddin; Yazid, Hafizal; Ahmad, Megat Harun Al Rashid Megat; Azman, Azraf; Mohamad, Glam Hadzir Patai; Hamzah, Nai'im Syaugi; Abu, Mohamad Puad

    Neutron Radiography facility in TRIGA MARK II PUSPATI reactor is being upgraded to obtain better image resolution as well as reducing exposure time. Collimator and exposure room are the main components have been designed for fabrication. This article focuses on the simulation part that was carried out to obtain the profile of collimated neutron beam by utilizing the neutron transport protocol code in the Monte Carlo N-Particle (MCNP) software. Particular interest is in the selection of materials for inlet section of the collimator. Results from the simulation indicates that a combination of Bismuth and Sapphire, each of which has 5.0 cm length that can significantly filter both the gamma radiation and the fast neutrons. An aperture made of Cadmium with 1.0 cm opening diameter provides thermal neutron flux about 1.8 x108 ncm-2s-1 at the inlet, but reduces to 2.7 x106 ncm-2s-1 at the sample plane. Still the flux obtained is expected to reduces exposure time as well as gaining better image resolution.

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

    International Nuclear Information System (INIS)

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

  4. Refurbishment of the rotating rack of the OSU TRIGA MKII reactor

    International Nuclear Information System (INIS)

    Many TRIGA reactors have experienced operational difficulties with the rotating racks used for sample irradiation. Generally the rack gradually becomes more difficult to rotate until it finally seizes. The recommended action at that point is replacement of the entire facility at a significant cost. The purpose of this paper is to describe the symptoms leading to rack failure and to present the results of a refurbishment procedure that does not involve the use of solvents which create mixed chemical and radioactive hazardous waste. The primary reason for rack failure is the buildup of sludge produced through irradiation of lubrication oil. The refurbishment procedure involves using a commercially available degreasing solution which can be pumped into and out of the rack with the objective of removing this sludge. The solution used is sold under the trade name 'Simple Green'. No radioactive material was detected on smear or air samples taken of the work area during the reifurbishment activities and the rack rotates freely in both direction even after eighteen months of operation. The only disadvantage to performing this procedure has been the need to maintain a very aggressive contamination control program when unloading samples from the rack. A very fine particulate material attaches to the outside of tubes used to encapsulate samples. This material can produce contamination levels of 10,000 dpm/100 cm2 in the worst cases but will typically produce local hot spots on the order of 1000 dpm. (author)

  5. The influence analysis of addition number of plate to heat exchanger performance of TRIGA 2000 reactor

    International Nuclear Information System (INIS)

    In order to reduce the existing bubble in the core of Bandung TRIGA 2000 reactor during its operation above 1000 kW, was done by increasing the effectivity of the heat exchanger (HE). One of the methods for increasing this effectivity is done by adding the number of plate to heat exchanger. To get an appropriate number of plate to be added on achieving its requirement, the analysis to know how the comparison of its performance on variation of addition the number of plate, is needed. The analysis was done by using the NTU-Effectivity method. The variables which influence its effectivity was obtained from the operational experiences since of the year 2000 until 2005. Besides that, it was assumed that the properties of working fluid had not much changed on its temperature and its pressure and small fouling deposit on the plate of HE. The results show that generally the addition of the number of plate would increase the effectivity of the heat exchanger. But for the low flow rate of the primary(600 gpm) and the high flow rate of the secondary(6000 gpm), a little bit of increasing effectivity was obtained for the addition the number of plate, and the effectivity had been reached to above 98%. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  7. 5. European conference of TRIGA users. Papers and abstracts

    International Nuclear Information System (INIS)

    The main conference topics were: Operation and maintenance experience of the TRIGA reactors; Development of new Low Enrichment Fuels (LEU); Dose assessments noble gas releases; Radiation protection and dosimetry measurements; Research reactors programs and experiments; and Application of TRIGA reactors

  8. 4. TRIGA owners' conference. Papers and abstracts

    International Nuclear Information System (INIS)

    The Conference covers the following aspects of TRIGA reactors operation: fuel utilization; TRIGA design and startup tests radiation release and unusual occurrences; operating experience; design of experimental facilities and instruments

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

    International Nuclear Information System (INIS)

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

  10. License renewal and power upgrade of the Cornell University TRIGA reactor

    International Nuclear Information System (INIS)

    The Cornell Mark II TRIGA reactor has been a principal facility for instruction and research in nuclear science and engineering at Cornell, and it has been extensively used by other departments at Cornell and by nearby universities and industries. Initially the fuel was low hydride, 8.5w/o 19%-enriched, aluminum clad; in 1974 it was changed to high-hydride, stainless-steel-clad. The maximum power has been 100 kW, with pulses to $2, and operation has been on a one-shift demand basis. Annual energy generation of 50 MWH has been typical. Standard features include a 4-inch tangential port and our 6-inch radial ports, a thermal column with hohlraum and vertical access, a central thimble, a 'rabbit', and a set of dry irradiation tubes, replacing the 'Lazy Susan'. The license was renewed and amended in November 1983; the new limits are 500 kW and $3 pulses. Physical changes to the facility included addition of a water-to-water heat exchanger and of a diffuser at the water outlet ∼ 60 cm above the core. The flow rate is 300 liters per minute in the primary (reactor) side of the heat exchanger. The temperature of the chilled water entering the secondary of the exchanger is ∼ 12?C; its flow rate is adjusted by a servo-controlled by-pass valve to maintain the desired range of pool water temperature. Steps taken to go to higher power included rearrangement of fuel elements to increase excess reactivity, recalibration of control rods, and power vs ion chamber current calibrations at successively higher power by comparing the rate of rise of pool temperature with a known rate using electrical heating elements. Steady-state operation has been done up to 480 kW (nominal) but pulsing at the newly allowed higher levels has not been tested as yet

  11. General improvements of the IPR-R1 TRIGA Mark I reactor in 37 years of operation

    International Nuclear Information System (INIS)

    The nuclear IPR-R1 TRIGA Mark I Reactor operating in the Nuclear Technology Development Center, originally Institute for Radioactive Research in Minas Gerais, Brazil, was dedicated in November 11, 1960.Initially operating for the production of radioisotopes for different uses, it started later to be used in large scale for neutron activation analysis and training of operators for nuclear power plants. Many improvements have been made throughout these years to provide a better performance in its operation and safety conditions. A new cooling system to operate until 300 k W, a new control console and a general remodeling of the reactor laboratory were some of the improvements added. To prevent and mitigate the ageing effects, the reactor operation personnel is starting a program to minimize future operation problems. Tis paper describes the improvements made, the results obtained during the past 37 years, and the precautions taken to ensure future safe operation of the reactor to give operators better conditions of safe

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

    CERN Document Server

    Riede, Julia

    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 a detector system located just above the water surface of the reactor tank. For the duration of one week, multiple gamma ray spectra were recorded automatically, starting each afternoon after reactor shutdown until the next morning. One measurement series has been recorded over the weekend. The Xe-135 peaks were extracted from a total of 1227 recorded spectra using an automated peak search algorithm and analyzed for their time-dependent properties. Although the background gamma radiation present in the core after shutdown...

  13. Transient behavior during reactivity insertion in the Moroccan TRIGA Mark II reactor using the PARET/ANL code

    Energy Technology Data Exchange (ETDEWEB)

    Boulaich, Y., E-mail: boulaich@cnesten.org.ma [CEN-Maamora, CNESTEN, Rabat (Morocco); Nacir, B. [CEN-Maamora, CNESTEN, Rabat (Morocco); El Bardouni, T. [Radiations and Nuclear Systems Laboratory, University Abdelmalek Essaadi, Faculty of Sciences of Tetouan (Morocco); CEN-Maamora, CNESTEN, Rabat (Morocco); Boukhal, H. [Radiations and Nuclear Systems Laboratory, University Abdelmalek Essaadi, Faculty of Sciences of Tetouan (Morocco); Chakir, E. [LHESIR, Department of Physics, Faculty of Sciences, Kénitra (Morocco); El Bakkari, B.; El Younoussi, C. [CEN-Maamora, CNESTEN, Rabat (Morocco)

    2015-04-01

    Highlights: • PARET model for the Moroccan TRIGA MARK II reactor has been developed. • Transient behavior under reactivity insertion has been studied based on PARET code. • Power factors required by PARET code have been calculated by using MCNP5 code. • The dependence on time of the main thermal-hydraulic parameters was calculated. • Results are largely far to compromise the thermal design limits. - Abstract: A three dimensional model for the Moroccan 2 MW TRIGA MARK II reactor has been developed for thermal-hydraulic and safety analysis by using the PARET/ANL and MCNP5 codes. This reactor is located at the nuclear studies center of Mâamora (CENM), Morocco. The model has been validated through temperature measurements inside two instrumented fuel elements located near the center of the core, at various power levels, and also through the power and fuel temperature evolution after the reactor shutdown (SCRAM). The axial distributions of power factors required by the PARET code have been calculated in each fuel element rod by using MCNP5 code. Based on this thermal-hydraulic model, a safety analysis under the reactivity insertion phenomenon has been carried out and the dependence on time of the main thermal-hydraulic parameters was calculated. Results were compared to the thermal design limits imposed to maintain the integrity of the clad.

  14. Transient behavior during reactivity insertion in the Moroccan TRIGA Mark II reactor using the PARET/ANL code

    International Nuclear Information System (INIS)

    Highlights: • PARET model for the Moroccan TRIGA MARK II reactor has been developed. • Transient behavior under reactivity insertion has been studied based on PARET code. • Power factors required by PARET code have been calculated by using MCNP5 code. • The dependence on time of the main thermal-hydraulic parameters was calculated. • Results are largely far to compromise the thermal design limits. - Abstract: A three dimensional model for the Moroccan 2 MW TRIGA MARK II reactor has been developed for thermal-hydraulic and safety analysis by using the PARET/ANL and MCNP5 codes. This reactor is located at the nuclear studies center of Mâamora (CENM), Morocco. The model has been validated through temperature measurements inside two instrumented fuel elements located near the center of the core, at various power levels, and also through the power and fuel temperature evolution after the reactor shutdown (SCRAM). The axial distributions of power factors required by the PARET code have been calculated in each fuel element rod by using MCNP5 code. Based on this thermal-hydraulic model, a safety analysis under the reactivity insertion phenomenon has been carried out and the dependence on time of the main thermal-hydraulic parameters was calculated. Results were compared to the thermal design limits imposed to maintain the integrity of the clad

  15. Dose calculation for accident situations at TRIGA research reactor using LEU fuel type

    International Nuclear Information System (INIS)

    The 14 MW TRIGA R.R. is a unique design of TRIGA conception. The core was fully converted in May 2006 to use LEU fuel instead of the HEU fuel type. The core contains 29 fuel assemblies, 8 control rods and beryllium reflector, associated instrumentation and controls. The U-235 enrichment for TRIGA - HEU fuel is 93.15 wt % and for TRIGA - LEU is 40.00 wt %. The differences between the two fuel types, as shown by the calculations, will results in a higher core inventory especially for heavy elements (i.e. actinides and transuranium elements), but modifications for noble gases, halogens and other volatile fission products are not so important. Dose calculations for an hypothetical accident scenario was considered and dose and radiological consequence calculations were performed. The results of the calculations and a discussion related on the differences between the consequences in the two cases are also presented. (authors)

  16. Benchmarking of the WIMSD/CITATION deterministic code system for the neutronic calculations of TRIGA Mark-III research reactors

    International Nuclear Information System (INIS)

    Highlights: • Benchmark study performed for the neutronic calculations of TRIGA research reactors. • WIMSD-5B/CITATION is the utilized code system along with the WIMSD-IAEA-69 library. • The studied condensed spectra are five and seven energy groups spectra. • Analyzed: lattice parameters, reactivities, CR worth, flux and power distribution. • The lattice and neutronic parameters showed the accuracy of both condensed spectra. - Abstract: The objective of this paper is to assess the suitability and accuracy of the deterministic diffusion method for the neutronic calculations of the TRIGA Mark-III research reactors using the WIMSD/CITATION code system in proposed condensed energy spectra of five and seven energy groups with one and three thermal groups respectively. The utilized cell transport calculations code and core diffusion calculations code are the WIMSD-5B and the CITVAP v3.1 codes respectively, along with the WIMSD-IAEA-69 nuclear data library. Firstly, the assessment goes through analyzing the integral parameters – keff, ρ238, δ235, δ238, and C* – of the TRX and BAPL benchmark lattices and comparison with experimental and previous reference results using other ENDLs at the full energy spectra which show good agreement with the references at both spectra. Secondly, evaluation of the 3D nuclear characteristics of three different cores of the TRR-1/M1 TRIGA Mark-III Thai research reactor at the condensed energy spectra. The results include the excess reactivities of the cores and the worth of selected control rods which were compared with reference Monte Carlo results and experimental values. The results show good agreement with the references at both energy spectra and the better accuracy are attainable in the five energy groups spectrum. The results also include neutron flux distributions which are evaluated for future comparisons with other calculational techniques even they are comparable to reactors and fuels of the same type. The

  17. Capture programs, analysis, data graphication for the study of the thermometry of the TRIGA Mark III reactor core; Programas de captura, analisis y graficado de datos para el estudio de la termometria del nucleo del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Paredes G, L.C

    1991-05-15

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

  18. Main configurations of the reactor core TRIGA Mark III of the ININ, during their operation; Principales configuraciones del nucleo del reactor TRIGA Mark III del ININ, durante su operacion

    Energy Technology Data Exchange (ETDEWEB)

    Nava S, W.; Raya A, R., E-mail: Wenceslao.nava@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    The Reactor TRIGA Mark III is 43 years old since was put lay critical on November 8 of 1968 for the first time, along their operative life there have been 18 different configurations of the core, being three those more important: the first configuration with elements standard with an enrichment lightly minor than 20% in U-235, the second configuration that deserves out attention is when a mixed core was charged, composite of two different fuels as for their enrichment, the core consisted of 26 fuel elements Flip (of high enrichment approximately of 70%) more 3 control bars with follower of fuel Flip and 59 standard fuel elements, as those mentioned previously, finally is necessary to consider the recent reload of the reactor, with a compound core by fuel elements of low enrichment LEU 30/20. In this work the characteristics more important of the reactor are presented as well as of each one of the described cores. (Author)

  19. Feasibility study for production of 99mTc by neutron irradiation of MoO3 in a 250 kW TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    The subject of this paper is to explore the possibility to obtain 99mTc from activation of 98Mo, using the TRIGA Mark II low flux research reactor (Vienna, Austria). Irradiation of both natural and enriched in 98Mo molybdenum oxides was compared. Aims of this work included the determination of neutron fluxes and 98Mo(n, γ)99Mo reaction effective cross section in the TRIGA Mark II reactor irradiation channels, calculation of 99Mo specific activities, determination of optimal irradiation conditions for the subsequent 99mTc separation from MoO3 targets using concentrating technologies. (author)

  20. Reactivity calculations for the fuel elements of I.T.U. TRIGA MARK-II reactor by means of one-group perturbation theory

    International Nuclear Information System (INIS)

    The reactivities of the fuel elements of I.T.U. TRIGA MARK-II reactor has been calculated by using both one-group perturbation theory and a one-dimensional, two-group diffusion computer code TRIGAP. For each fuel element, reactivities calculated by both methods are compared with those measured experimentally. It is seen that the reactivity calculations made by using the one-group perturbation theory give the results with better accuracy in comparison to TRIGAP. One-group perturbation theory can be easily applied to the reactivity calculations of fuel elements of TRIGA type reactors in acceptable range (orig.)

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

    International Nuclear Information System (INIS)

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

  2. Gamma and Neutron Flux of a Prompt Gamma Neutron Activation Analysis Collimator at the PUSPATI TRIGA Reactor

    International Nuclear Information System (INIS)

    A Prompt Gamma Neutron Activation Analysis (PGNAA) facility is being studied for installation at PUSPATI TRIGA Reactor (RTP) under the Thorium Flagship programme. This work presents the preliminary design of a PGNAA collimator at the RTP. The result of calculations for gamma and neutron flux at various positions of the PGNAA collimator in the RTP beam port 1 by using the computer code MCNPX are presented and discussed. The results indicate the technical feasibility of the installation of PGNAA facility at the RTP and the possibility of enhancing the utilization of the RTP. (author)

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

    International Nuclear Information System (INIS)

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

  4. Comparative Study of some Parameters reported in the Safety Analysis Report of TRIGA MARK II Research reactor with Calculations

    International Nuclear Information System (INIS)

    An attempt has been made to investigate some of the parametric results reported in the safety Analysis Report (SAR) with the theoretical analysis carried out by different computer codes and data bases. Different neutronics, thermal hydraulics and safety parameters such as core criticality and burnup lifetime, power peaking factor, prompt negative temperature coefficient, neutron flux, pulse characteristics, steady state and transient behaviors of the TRIGA reactor were analyzed. The investigated results were found to be in fairly good agreement with the values reported in the SAR. 12 refs., 14 figs., 1 table (Author)

  5. V2:Performance of the solid deuterium ultra-cold neutron source at the pulsed reactor TRIGA Mainz

    OpenAIRE

    Karch, J.; Sobolev, Yu.; M. Beck; 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%....

  6. COMMIX-1C code estimation for the pool dynamics of Istanbul Technical University TRIGA MARK-II reactor

    International Nuclear Information System (INIS)

    In this study, the COMMIX-1C code is used to investigate the pool dynamics of Istanbul Technical University (ITU)TRIGA MARK-II reactor by simulating the velocity, pressure and temperature distributions in the reactor pool as a function of core design parameters and pool configuration. COMMIX-1C is multi-purpose, three-dimensional. transient, single-phase, thermal-hydraulics computer code. For the mass, momentum and energy equations, it uses a porous-medium formulation, a finite-volume algorithm, a flow modulated skew-upwind discretization scheme to reduce numerical diffusion and k-ε two-equation turbulence model. Its implementation for the particular system requires geometric and physical modelling decisions. ITU TRIGA MARK-II reactor pool is considered partly as continuum and partly as porous medium. All the major pool components are explicitly modelled in the simulation. Shape of the pool structure and computational cells are accounted for using the concept of directional surface permeability, volume porosity, distributed resistance, and distributed heat source or sink. The results are compared to the results of the computer codes TRISTAN, TRIGATH and TRIGATH-R

  7. Operation and maintenance of the 250 kW TRIGA Mark II reactor at the Jozef Stefan Institute in Ljubljana

    International Nuclear Information System (INIS)

    Over the last two years the TRIGA Mark II reactor in Ljubljana has been operated at an energy release of about 2500 MWh or about 4100 hour per year. In this period, about 1800 samples were irradiated. In 1983, a new core configuration was established because all Al-clad fuel elements in the core were replaced by the SS-clad elements. The 'J.Stefan' Institute received in 1983, namely, from the TRIGA reactor in Neuherberg, Federal Republic of Germany 107 SS-clad partly burned fuel elements together with some instrumentation which will gradually replace the old radiological and safety instrumentation. The transportation and the dischange of the highly radioactive fuel was done quickly and without any problem. During the last two years the reactor has been operated without any longer shut-down due to technical difficulties. In 1983 we noticed only a fuel element failure during operation. After the short inspection the fuel element with a small clading hole was found and replaced by a new one. (orig.)

  8. Current utilization and long term strategy of the Finnish TRIGA research reactor FiR 1

    Energy Technology Data Exchange (ETDEWEB)

    Auterinen, Iiro; Salmenhaara, Seppo [VTT Technical Research Centre of Finland Otaniemi, Espoo (Finland)

    2008-10-29

    FiR 1 (TRIGA Mark II, 250 kW) has an important international role in the development of boron neutron capture therapy (BNCT) for cancer. The safety and efficacy of BNCT is studied for several different cancers: - primary glioblastoma, a highly malignant brain tumour (since 1999); - recurrent glioblastoma or anaplastic astrocytoma (since 2001); - recurrent inoperable head and neck carcinoma (since 2003). It is one of the few facilities in the world providing this kind of treatments. The successes in the BNCT development have now created a demand for these treatments, although they are given on an experimental basis. Well over 100 patients treated now since May 1999: - at least 1 patient irradiation / week, often 2 (Tuesday and Thursday) - patients are referred to BNCT-treatments from several hospitals, also outside research protocols; - the hospitals pay for the treatment. The FiR 1 reactor has proven to be a reliable neutron source for the BNCT treatments; no patient irradiations have been cancelled because of a failure of the reactor. The BNCT facility has become a center of extensive academic research especially in medical physics. Nuclear education and training continue to play also a role at FiR 1 in the form of university courses and training of nuclear industry personnel. FiR 1 is one of the two sources in Scandinavia for short lived radioisotopes used in tracer studies in industry. The main isotope produced is Br-82 in the form of either KBr or ethylene bromide. Other typical isotopes are Na-24, Ar-41, La-140. The isotopes are used mainly in tracer studies in industry (Indmeas Inc., Finland). Typical activity of one irradiated Br-sample is 20 - 80 GBq; total activity produced in one year is over 3 TBq; the reactor operating time needed for the isotope production is one or two days per week. Accelerator based neutron sources are developed for BNCT. The prospect is that when BNCT will achieve a status of a fully accepted and efficient treatment modality for

  9. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jennifer Lyons; Wade R. Marcum; Mark D. DeHart; Sean R. Morrell

    2014-01-01

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  10. Reactor Physics Scoping and Characterization Study on Implementation of TRIGA Fuel in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    The Advanced Test Reactor (ATR), under the Reduced Enrichment for Research and Test Reactors (RERTR) Program and the Global Threat Reduction Initiative (GTRI), is conducting feasibility studies for the conversion of its fuel from a highly enriched uranium (HEU) composition to a low enriched uranium (LEU) composition. These studies have considered a wide variety of LEU plate-type fuels to replace the current HEU fuel. Continuing to investigate potential alternatives to the present HEU fuel form, this study presents a preliminary analysis of TRIGA® fuel within the current ATR fuel envelopes and compares it to the functional requirements delineated by the Naval Reactors Program, which includes: greater than 4.8E+14 fissions/s/g of 235U, a fast to thermal neutron flux ratio that is less than 5% deviation of its current value, a constant cycle power within the corner lobes, and an operational cycle length of 56 days at 120 MW. Other parameters outside those put forth by the Naval Reactors Program which are investigated herein include axial and radial power profiles, effective delayed neutron fraction, and mean neutron generation time.

  11. Estimation of (41)Ar activity concentration and release rate from the TRIGA Mark-II research reactor.

    Science.gov (United States)

    Hoq, M Ajijul; Soner, M A Malek; Rahman, A; Salam, M A; Islam, S M A

    2016-03-01

    The BAEC TRIGA research reactor (BTRR) is the only nuclear reactor in Bangladesh. Bangladesh Atomic Energy Regulatory Authority (BAERA) regulations require that nuclear reactor licensees undertake all reasonable precautions to protect the environment and the health and safety of persons, including identifying, controlling and monitoring the release of nuclear substances to the environment. The primary activation product of interest in terms of airborne release from the reactor is (41)Ar. (41)Ar is a noble gas readily released from the reactor stacks and most has not decayed by the time it moves offsite with normal wind speed. Initially (41)Ar is produced from irradiation of dissolved air in the primary water which eventually transfers into the air in the reactor bay. In this study, the airborne radioisotope (41)Ar generation concentration, ground level concentration and release rate from the BTRR bay region are evaluated theoretically during the normal reactor operation condition by several governing equations. This theoretical calculation eventually minimizes the doubt about radiological safety to determine the radiation level for (41)Ar activity whether it is below the permissible limit or not. Results show that the estimated activity for (41)Ar is well below the maximum permissible concentration limit set by the regulatory body, which is an assurance for the reactor operating personnel and general public. Thus the analysis performed within this paper is so much effective in the sense of ensuring radiological safety for working personnel and the environment. PMID:26736180

  12. Use of the TRIGA Reactor by the Radiochemistry Group of the Atominstitute of the Austrian Universities

    International Nuclear Information System (INIS)

    The Radiochemistry Group of the Atominstitute of the Austrian Universities uses the TRIGA Mark II Reactor mainly for neutron activation analysis. Transport of samples to and from the irradiation positions in the reactor is performed by fast pneumatic transfer systems (transfer time 20 msec and 300 msec) and slow conventional transport facilities. Gamma-spectrometric instrumentation equipped with loss free counting systems is used to handle the high count rates up to 500 000 counts/sec. During the last years neutron activation analysis was applied to investigate environmental samples (soil, dust, incineration ash), geological samples (rocks, sediments, fossils, volcanic gases), biological materials (lichens, mushrooms and other plant materials, human diet, biological reference materials), raw materials (phosphate, coal) and archaeological materials (ancient glass). Lichen analysis was used for environmental monitoring. The content of some of the trace elements can be correlated with industrial activities, like manganese content with steel industry, the occurrence of vanadium and nickel with oil firing plants and stainless steel industry, selenium is found in lichen near coal firing plants. The amount of chlorine and sodium indicates the application of salt for road treatment during winter time, aluminum, scandium and hafnium content depends on the amount of dust in the environment. A further environmental application of neutron activation analysis is the determination of trace elements in volcanic gases. The halogens, arsenic, antimony, selenium, tellurium and mercury were determined and their daily output was calculated. The distribution of trace elements in fossils of known age gives us a geochemical key to condition and development of the paleo-environment. For this purpose we determined rare earth elements in 250 million years old microfossils (conodonts). Neutron activation analysis served also for some non scientific but nevertheless useful purposes: Organic

  13. IFPE/EFE-RO, Experimental Fuel Elements RO89 and RO51 in TRIGA 14 MW Reactor (INR-Pitesti)

    International Nuclear Information System (INIS)

    Description of program or function: Romanian irradiation tests concerned with Candu type fuel elements behavior and with the limits of the design parameters. A particular feature of the Candu fuel project is the small plenum (void volume) added for relaxation of the fission gases, which are inherently released during the fuel irradiation. Two irradiation tests in the C2 device from the TRIGA 14 MW reactor were performed between the years 1985-1987. The tests were done to evaluate the effect of the fuel density on the time-evolution of the fission gas pressure. Experimental fuel elements were adequately instrumented with pressure transducers to follow the fission gas pressure changes during fuel irradiation. The first irradiation test was conducted on the fuel element coded No.89 whose main characteristics were the nominal values of the main fuel design parameters. The second one was conducted on the fuel element coded No.51. Because of the axial flux asymmetry inside the TRIGA reactor core, the experimental elements are shorter in length than the Candu fuel design. The irradiation tests consisted in evaluation of the time-evolution of the internal pressure from two experimental fuel elements having the main design characteristics as the Romanian Candu type fuel element design and to follow the dependence of the internal pressure of the fission gas on the fuel density

  14. Operation and maintenance of the 250 kW TRIGA Mark II reactor at the J. Stefan Institute

    International Nuclear Information System (INIS)

    Over the last two years the TRIGA Mark II reactor in Ljubljana has been operated at an energy release of about 2250 MWh or about 4200 hours per year. In this period, about 2000 samples were irradiated. Since the last TRIGA Owners' Conference there has been an increase in all operational data because of a very extensive programme of irradiation of molybdenum for the everyday production of technetium-99 m by a solvent extraction method. Because of its age and absolencence replacement of the console electronics was considered some time ago. Therefore, partly new instrumentation was installed this year. A new console is under construction. Furthermore, a new core configuration was established after 7 fresh FLIP fuel elements were delivered by GA. At this time it was noticed that 2 dummy elements are stuck in the upper grid plate. They will be exchanged during the regular maintenance work in August this year. During the last two years the reactor has been operated without any longer shut-down due to technical difficulties. (author)

  15. Implementation of k0-INAA standardisation at ITU TRIGA Mark II research reactor, Turkey based on k0-IAEA software

    Science.gov (United States)

    Esen, Ayse Nur; Haciyakupoglu, Sevilay

    2016-02-01

    The purpose of this study is to test the applicability of k0-INAA method at the Istanbul Technical University TRIGA Mark II research reactor. The neutron spectrum parameters such as epithermal neutron flux distribution parameter (α), thermal to epithermal neutron flux ratio (f) and thermal neutron flux (φth) were determined at the central irradiation channel of the ITU TRIGA Mark II research reactor using bare triple-monitor method. HPGe detector calibrations and calculations were carried out by k0-IAEA software. The α, f and φth values were calculated to be -0.009, 15.4 and 7.92·1012 cm-2 s-1, respectively. NIST SRM 1633b coal fly ash and intercomparison samples consisting of clay and sandy soil samples were used to evaluate the validity of the method. For selected elements, the statistical evaluation of the analysis results was carried out by z-score test. A good agreement between certified/reported and experimental values was obtained.

  16. Benchmark analysis of reactivity experiment in the TRIGA Mark 2 reactor using a continuous energy Monte Carlo code MCNP

    International Nuclear Information System (INIS)

    A good model on experimental data (criticality, control rod worth, and fuel element worth distributions) is encouraged to provide from the Musashi-TRIGA Mark 2 reactor. In the previous paper, as the keff values for different fuel loading patterns had been provided ranging from the minimum core to the full one, the data would be candidate for an ICSBEP evaluation. Evaluation of the control rod worth and fuel element worth distributions presented in this paper could be an excellent benchmark data applicable for validation of calculation technique used in the field of modern research reactor. As a result of simulation on the TRIGA-2 benchmark experiment, which was performed by three-dimensional continuous-energy Monte Carlo code (MCNP4A), it was found that the MCNP calculated values of control rod worth were consisted to the experimental data for both rod-drop and period methods. And for the fuel and the graphite element worth distributions, the MCNP calculated values agreed well with the measured ones though consideration of real control rod positions was needed for calculating fuel element reactivity positioned in inner ring. (G.K.)

  17. Performance and improvements of the IPR-R1 Triga Mark I reactor in 45 years of operation

    International Nuclear Information System (INIS)

    The TRIGA Mark I IPR-R1 Reactor operates in the Nuclear Technology Development Center-CDTN/CNEN, originally Institute of Radioactive Researches, in Belo Horizonte, Minas Gerais, since November 6, 1960. Initially it was operated for radioisotopes production for different uses, being later used in wide scale for other purposes as neutron activation analysis and training of operators for nuclear power plants. Along the years, several improvements were introduced in the reactor and in its auxiliary systems providing better use of its facilities and optimizing the safety in the operation. A new cooling system, control rod mechanism, aluminum reactor tank, pneumatic system optimization, new control console and a general reform at the reactor room are some of these improvements. The reactor arrives at the 45 years, with a decrease in the rhythm of the works but with perspective of new applications for the same. This paper reports the performance of the reactor in 45 years of operation, providing data about energy released, samples irradiated, hours of operation and purposes of the isotopes produced and cybernetics and information technologies used to provide reactor calculations and monitoring parameters. (author)

  18. Identification of a leaking TRIGA fuel element at the nuclear reactor facility of the University of Pavia

    International Nuclear Information System (INIS)

    During a periodical activity of characterization of the ionic-exchange resins of the demineralizer of the primary cooling circuit of the TRIGA Mark-2 reactor of the University of Pavia a small but detectable amount of Cs137 contamination was measured. Since the reactor has been running for several hundreds of hours at full power without showing any anomaly in the radiometric and thermo-hydraulic parameters, the reactor was brought to the nominal power of 250 kW for one hour and a sample of water was collected from the reactor tank and analysed in a low-background gamma ray detector. As a result a small amount of fission products were detected in the reactor pool water (a few Bq/g) suggesting the existence of a possible clad defect in one or more fuel elements. Since no halogens such as iodine and bromine were detected in the sampled water, the more likely hypothesis, also supported by literature, seemed to be a micro-fissure in the neck of an instrumented fuel element. A dedicated apparatus for reactor pool water sampling and on-line spectroscopy measurements was realized. As expected, the fission product leakage was due to a micro-fissure of a fuel element that released noble gas only when the fuel element was heated up to a temperature around 90 Celsius degrees. The leaking fuel element was identified and removed from its position and the reactor was back in regular operation after 2 months from leakage detection. (authors)

  19. Neutron flux measurement in the thermal column of the Malaysian TRIGA mark II reactor with MCNP verification

    International Nuclear Information System (INIS)

    A study of the thermal column of the Malaysian TRIGA Mark II reactor, forming part of a feasibility study for BNCT was proposed in 2001. In the current study, pure metals were used to measure the neutron flux at selected points in the thermal column and the neutron flux determined using SAND-II. Monte Carlo simulation of the thermal column was also carried out. The reactor core was homogenized and calculations of the neutron flux through the graphite stringers performed using MCNP5. The results show good agreement between the measured flux and the MCNP calculated flux. An obvious extension from this is that the MCNP neutron flux output can be utilized as an input spectrum for SAND-II for the flux iteration. (author)

  20. Increasing the power of FiR 1 TRIGA reactor by a factor of 2 1/2

    International Nuclear Information System (INIS)

    An early domestic reactor engineering project was increasing the neutron flux of the Triga Mark II reactor in Otaniemi by a factor of 2 1/2, thirty-five years ago. The thermal power of the facility was increased from 100 kW to 250 kW by modifications made in the fuel loading, control rods, control and protection systems, radiation shielding structures and in the heat removal system. This improved the efficiency of the plant and reduced time requirements in proportion for physical research, isotope production and medical irradiations. Experimental runs were made at 318 kW power, and the final approval inspection for continuous operation at 250 kW was completed on August 3, 1967. (author)

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

    International Nuclear Information System (INIS)

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

  2. Home-made refurbishment of the instrumentation and control system of the TRIGA reactor of the University of Pavia

    International Nuclear Information System (INIS)

    The Instrumentation and Control (I and C) System of the TRIGA reactor of the University of Pavia was dated and, in order to grant a safe and continuous reactor operation for the future, it became necessary to substitute or to upgrade the system. Since the substitution of the I and C system with a new-made one was very difficult to be performed due to long authorization procedures, an home-made refurbishment was planned. Using commercial components of high quality, almost a complete substitution, channel-by-channel, of the I and C system was realized without changing the operating and safety logics. The system includes: - the Reactor Linear Power Channel and Chart Recorder; - the Reactor Percent Power Safety Channel; - the High Voltage and Low Voltage Power Supply; - the Automatic Reactor Power Control; - the Fuel Elements and Cooling-Water Temperatures Measuring Channels; - the Water Conductivity Measuring Channel. The refurbished I and C system shows a very good operational behavior and reliability and will assure a continuous operation of the reactor for the future

  3. Behavior of exposed human lymphocytes to a neutron beam of the reactor TRIGA Mark III

    International Nuclear Information System (INIS)

    Excessive exposure to ionizing radiation occurs in people who require radiation treatment, also in those for work can come to receive doses above the permitted levels. A third possibility of exposure is the release of radioactive material in which the general population is affected. Most of the time the exhibition is partial and only rarely occurs throughout the body. For various reasons, situations arise where it is impossible to determine by conventional physical methods, the amount of radiation you were exposed to the affected person and in these cases where the option to follow is the Biological Dosimetry, where the analysis of chromosomes dicentrics is used to estimate the dose of ionizing radiation exposure. A calibration curve is generated from in vitro analysis of dicentric chromosome, which are found in human lymphocytes, treated with different types and doses of radiation. The dicentric is formed from two lesions, one on each chromosome and their union results in a structure having two centromeres, acentric fragment with her for the union of several chromosomes leads to more complex structures as tri-centric s, tetra or penta-centric s, which have the same origin. The dose-response curve is estimated by observing the frequency of dicentrics and extrapolated to a dose-effect curve previously established, for which it is necessary that each lab has its own calibration curves, taking into account that for a Let low radiation, dose-effect curve follows a linear-quadratic model Y=C + αD + βD. The production of dicentric chromosomes with a high Let, was studied using a beam of neutrons generated in the reactor TRIGA Mark III with an average energy of 1 MeV, adjusting the linear model Y=αD. The dose-response relationship is established in blood samples from the same donor, the coefficient α of the dose-response is Y = (0.3692 ± 0.011 * D), also shows that saturation is reached in system 4 Gy. (Author)

  4. Feasibility study on vitrification of spent ion exchange resins from TRIGA Reactor Malaysia

    International Nuclear Information System (INIS)

    Feasibility studies on the vitrification of spent ion exchange resins combined with glass cullet powder have been conducted using a High Temperature Test Furnace. Bottle glass cullet powder was used as matrix material to convert the ash of the spent resins into a glass. Vitrification of spent ion exchange resins presents a reasonable disposal alternative, because of its inherent organic destruction capabilities, the volume reduction levels obtainable, and the durable product that it yields. In this study, the spent ion exchange resin from the PUSPATI TRIGA reactor of Nuclear Malaysia was combusted in a lab scale combustor and the resulting ash was vitrified together with glass cullet powder in a high temperature furnace to produce a stable spent resin ash embedded in glass. The factors affecting this immobilized waste, such as thermal stability, radiological stability and leachability have all been investigated. However, the outcome of these tests, which include the radionuclide activity concentration in the slag, the optimum conditioning temperature - in relation with volume reduction during vitrification - and the volume mixing ratio of matrix material were reported. It was found that the radionuclides present in spent resins were 54Mn, 60Co and 152Eu. The elementary chemical composition (carbon, hydrogen, nitrogen and sulphur) of spent resins was 27.6 % C, 5.68 % H, 2.04 % N and 4.20 % S, respectively. The maximum calorific value of spent resins was 1735 kJ/ kg. The average activity concentrations of 54Mn and 60Co in ash at 200 degree Celsius were 9,411 ± 243 Bq/ Kg and 12,637 ± 201 Bq/ Kg. flue gases containing CO2, CO, SO2 and NO started to be emitted above 200 degree Celsius. The optimum conditioning temperature was also the highest tested, for example 900 degree Celsius in 45 minutes, and the best mixing ratio ash to matrix material was also the highest, for example 1:9. Finally, the leaching analysis of slag at 900 degree Celsius in 45 minutes showed that

  5. Feasibility study of the university of Utah TRIGA reactor power upgrade - part II: Thermohydraulics and heat transfer study in respect to cooling system requirements and design

    OpenAIRE

    Babitz Philip; Choe Dongok; Jevremovic Tatjana

    2013-01-01

    The thermodynamic conditions of the University of Utah's TRIGA Reactor were simulated using SolidWorks Flow Simulation, Ansys, Fluent and PARET-ANL. The models are developed for the reactor's currently maximum operating power of 90 kW, and a few higher power levels to analyze thermohydraulics and heat transfer aspects in determining a design basis for higher power including the cost estimate. It was found that the natural convection current becomes much mor...

  6. Brief History of Herpetology in the Museum of Vertebrate Zoology, University of California, Berkeley, with a List of Type Specimens of Recent Amphibians and Reptiles

    OpenAIRE

    Rodríguez-Robles, Javier A; Good, David A; Wake, David B

    2003-01-01

    An overview of the herpetological program of the Museum of Vertebrate Zoology (MVZ), University of California, Berkeley, is presented. The history of herpetological activities in the MVZ and more generally at Berkeley is summarized. Although the MVZ has existed since 1908, until 1945 there was no formal curator for the collection of amphibians and nonavian reptiles. Since that time Robert C. Stebbins, David B. Wake, Harry W. Greene, Javier A. Rodríguez-Robles (in an interim capacity), and Cra...

  7. Experimental study of the temperature distribution in the TRIGA IPR-R1 Brazilian research reactor; Investigacao experimental da distribuicao de temperaturas no reator nuclear de pesquisa TRIGA IPR-R1

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Zacarias

    2005-07-01

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

  8. Loss of Coolant Accident Analysis for 1MW PUSPATI Triga Mark II Research Reactor (RTP) Using MARS-KS Code

    Energy Technology Data Exchange (ETDEWEB)

    Abd, Aziz Sadri [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Shin, Andong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    RTP is a pool type reactor cooled by natural circulation and the reactor core is located at the bottom of a demineralized water-filled aluminum liner tank of 2.0 meter diameter and 6.5 meter depth. The core assembly is composed of 100 cylindrical fuel rods including of 4 control rods in circular array. From the literature, development of thermal hydraulic analysis of RTP using computer code has not been well established. Therefore, establishment and development of appropriate thermal hydraulic safety analysis model is very critical to ensure the safety operation of the reactor. Hence, key thermal hydraulic parameters of RTP reactor operating under steady state and transient condition were investigated. In this paper, Loss Of Coolant Accident (LOCA) were calculated and analyzed and compared with corresponding values in Safety Analysis Report (SAR) 2008 and test report. PUSPATI Triga Mark II research reactor (RTP) has been operated at Malaysian Nuclear Agency since 1982 and primary cooling system was modified in 2010. Thermal hydraulic modeling of RTP of 1MWt has been successfully investigated with MARS-KS code. The calculated normal operation parameters have been compared with reactor Safety Analysis Report (SAR) and experimental data. Most of the thermal hydraulic parameters show good agreement with SAR and experimental data within an acceptable percentage error. The loss of coolant accident was simulated in case of leak of primary side heat exchanger gasket. The calculation result showed fast decrease of reactor pool level. About 5 minutes after the leak, reactor tank was fully depleted. Furthermore, claddings temperature was reached 1173.4K at 3270s which could result in failure of SS304 cladding. Based on the assessment, it is found that appropriate remedies including physical modifications or emergency procedures need be prepared to protect the reactor tank depletion by the heat exchanger leak accident.

  9. Loss of Coolant Accident Analysis for 1MW PUSPATI Triga Mark II Research Reactor (RTP) Using MARS-KS Code

    International Nuclear Information System (INIS)

    RTP is a pool type reactor cooled by natural circulation and the reactor core is located at the bottom of a demineralized water-filled aluminum liner tank of 2.0 meter diameter and 6.5 meter depth. The core assembly is composed of 100 cylindrical fuel rods including of 4 control rods in circular array. From the literature, development of thermal hydraulic analysis of RTP using computer code has not been well established. Therefore, establishment and development of appropriate thermal hydraulic safety analysis model is very critical to ensure the safety operation of the reactor. Hence, key thermal hydraulic parameters of RTP reactor operating under steady state and transient condition were investigated. In this paper, Loss Of Coolant Accident (LOCA) were calculated and analyzed and compared with corresponding values in Safety Analysis Report (SAR) 2008 and test report. PUSPATI Triga Mark II research reactor (RTP) has been operated at Malaysian Nuclear Agency since 1982 and primary cooling system was modified in 2010. Thermal hydraulic modeling of RTP of 1MWt has been successfully investigated with MARS-KS code. The calculated normal operation parameters have been compared with reactor Safety Analysis Report (SAR) and experimental data. Most of the thermal hydraulic parameters show good agreement with SAR and experimental data within an acceptable percentage error. The loss of coolant accident was simulated in case of leak of primary side heat exchanger gasket. The calculation result showed fast decrease of reactor pool level. About 5 minutes after the leak, reactor tank was fully depleted. Furthermore, claddings temperature was reached 1173.4K at 3270s which could result in failure of SS304 cladding. Based on the assessment, it is found that appropriate remedies including physical modifications or emergency procedures need be prepared to protect the reactor tank depletion by the heat exchanger leak accident

  10. A study on source term assessment and waste disposal requirement of decontamination and decommissioning for the TRIGA research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Whang, Joo Ho; Lee, Kyung JIn; Lee, Jae Min; Choi, Gyu Seup; Shin, Byoung Sun [Kyunghee Univ., Seoul (Korea, Republic of)

    1999-08-15

    The objective and necessity of the project : TRIGA is the first nuclear facility that decide to decommission and decontamination in our nation. As we estimate the expected life of nuclear power generation at 30 or 40 years, the decommissioning business should be conducted around 2010, and the development of regulatory technique supporting it should be developed previously. From a view of decommissioning and decontamination, the research reactor is just small in scale but it include all decommissioning and decontamination conditions. So, the rules by regulatory authority with decommissioning will be a guide for nuclear power plant in the future. The basis of regulatory technique required when decommissioning the research reactor are the radiological safety security and the data for it. The source term is very important condition not only for security of worker but for evaluating how we dispose the waste is appropriate for conducting the middle store and the procedure after it when the final disposal is considered. The content and the scope in this report contain the procedure of conducting the assessment of the source term which is most important in understanding the general concept of the decommissioning procedure of the decommissioning and decontamination of TRIGA research reactor. That is, the sampling and measuring method is presented as how to measure the volume of the radioactivity of the nuclear facilities. And also, the criterion of classifying the waste occurred in other countries and the site release criteria which is the final step of decommissioning and decontamination presented through MARSSIM. Finally, the program to be applicable through comparing the methods of our nation and other countries ones is presented as plan for disposal of the waste in the decommissioning.

  11. Data acquisition and signal processing system for IPR R1 TRIGA-Mark I nuclear research reactor of CDTN

    International Nuclear Information System (INIS)

    The TRIGA IPR-R1 Nuclear Research Reactor, located at the Nuclear Technology Development Center (CDTN/CNEN) in Belo Horizonte, Brazil, is being operated since 44 years ago. The main operational parameters were monitored by analog recorders and counters located in the reactor control console. The reactor operators registered the most important operational parameters and data in the reactor logbook. This process is quite useful, but it can involve some human errors. It is also impossible for the operators to take notes of all variables involving the process mainly during fast power transients in some operations. A PC-based data acquisition was developed for the reactor that allows online monitoring, through graphic interfaces, and shows operational parameters evolution to the operators. Some parameters that were not measured, like the power and the coolant flow rate at the primary loop, are monitored now in the computer video monitor. The developed system allows measuring out all parameters in a frequency up to 1 kHz. These data is also recorded in text files available for consults and analysis. (author)

  12. Characterization of the neutron flux in the Hohlraum of the thermal column of the TRIGA Mark III reactor of the ININ; Caracterizacion del flujo neutronico en el Hohlraum de la columna termica del reactor TRIGA Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Delfin L, A.; Palacios, J.C.; Alonso, G. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico)]. e-mail: adl@nuclear.inin.mx

    2006-07-01

    Knowing the magnitude of the neutron flux in the reactor irradiation facilities, is so much importance for the operation of the same one, like for the investigation developing. Particularly, knowing with certain precision the spectrum and the neutron flux in the different positions of irradiation of a reactor, it is essential for the evaluation of the results obtained for a certain irradiation experiment. The TRIGA Mark III reactor account with irradiation facilities designed to carry out experimentation, where the reactor is used like an intense neutron source and gamma radiation, what allows to make irradiations of samples or equipment in radiation fields with components and diverse levels in the different facilities, one of these irradiation facilities is the Thermal Column where the Hohlraum is. In this work it was carried out a characterization of the neutron flux inside the 'Hohlraum' of the irradiation facility Thermal Column of the TRIGA Mark III reactor of the Nuclear Center of Mexico to 1 MW of power. It was determined the sub cadmic neutron flux and the epi cadmic by means of the neutron activation technique of thin sheets of gold. The maps of the distribution of the neutron flux for both energy groups in three different positions inside the 'Hohlraum' are presented, these maps were obtained by means of the irradiation of undressed thin activation sheets of gold and covered with cadmium in arrangements of 10 x 12, located parallel to 11.5 cm, 40.5 cm and 70.5 cm to the internal wall of graphite of the installation in inverse address to the position of the reactor core. Starting from the obtained values of neutron flux it was found that, for the same position of the surface of irradiation of the experimental arrangement, the relative differences among the values of neutron flux can be of 80%, and that the differences among different positions of the irradiation surfaces can vary until in a one order of magnitude. (Author)

  13. The reactor core TRIGA Mark-III with fuels type 30/20; El nucleo del reactor TRIGA Mark-III con combustible tipo 30/20

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F., E-mail: fortunato.aguilar@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    This work describes the calculation series carried out with the program MCNP5 in order to define the configuration of the reactor core with fuels 30/20 (fuels with 30% of uranium content in the Or-Zr-H mixture and a nominal enrichment of 20%). To select the configuration of the reactor core more appropriate to the necessities and future uses of the reactor, the following criterions were taken into account: a) the excess in the reactor reactivity, b) the switch out margin and c) to have new irradiation facilities inside the reactor core. Taking into account these criterions is proceeded to know the characteristics of the components that form the reactor core (dimensions, geometry, materials, densities and positions), was elaborated a base model of the reactor core, for the MCNP5 code, with a configuration composed by 85 fuel elements, 4 control bars and the corresponding structural elements. The high reactivity excess obtained with this model, gave the rule to realize other models of the reactor core in which the reactivity excess and the switch out margin were approximate to the values established in the technical specifications of the reactor operation. Several models were realized until finding the satisfactory model; this is composite for 74 fuels, 4 control bars and 6 additional experimental positions inside the reactor core. (Author)

  14. Development of a software for the control of the quality management system of the TRIGA-Mark III reactor; Desarrollo de un software para el control del sistema de gestion de calidad del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Herrera A, E. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Hernandez, L.V.; Hernandez, J.A. [UAEM, Depto. de Ingenieria en Computacion, 50000 Toluca, estado de Mexico (Mexico)]. e-mail: eha@nuclear.inin.mx

    2006-07-01

    The quality has not only become one of the essential requirements of the product but rather at the presenme it is a strategic factor key of which depends the bigger part of the organizations, not only to maintain their position in the market but also to assure their survival. The good organizations will have processes, procedures and standards to confront these challenges. The big organizations require of the certification of their administration systems, and once the organization has obtained this certification the following step it is to maintain it. The implementation and certification of an administration system requires of an appropriate operative organization that achieves continuous improvements in their operation. This is the case of the TRIGA Mark III reactor, which contains a computer program that upgrades, it controls and it programs activities to develop in the Installation, allowing one operative organization to the whole personnel of the same one. With the purpose of avoiding activities untimely. (Author)

  15. Computer aided design (CAD) for electronics improvement of the nuclear channels of TRIGA Mark III reactor of the ININ; Diseno asistido por computadora (DAC) para mejorar la electronica de los canales nucleares del reactor TRIGA Mark III del ININ

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez M, J.L.; Rivero G, T.; Aguilar H, F. [ININ, 52750 La Marquesa, Estado de Mexico (Mexico)]. e-mail: jlgm@nuclear.inin.mx

    2007-07-01

    The 4 neutron measurement channels of the digital control console (CCD) of the TRIGA Mark III reactor (RTMIII) of the ININ, its were designed and built with the corresponding Quality Guarantee program, being achieved the one licensing to replace the old console. With the time they were carried out some changes to improve and to not solve some problems detected in the tests, verification and validation, requiring to modify the circuits originally designed. In this work the corrective actions carried out to eliminate the Non Conformity generated by these problems, being mentioned the advantages of using modern tools, as the software applied to the Attended Engineering by Computer, and those obtained results are presented. (Author)

  16. Refurbishment, Modernization and Ageing Management Program of The 3MW TRIGA Mark-II Research Reactor of Bangladesh

    Energy Technology Data Exchange (ETDEWEB)

    Salam, M. A. [Atomic Energy Research Establishment, Dhaka (Bangladesh)

    2013-07-01

    The 3 MW TRIGA MK-II research reactor of Bangladesh Atomic Energy Commission (BAEC) achieved its first criticality on 14 September 1986. The reactor has been used for manpower training, radioisotope production and various R and D activities in the field of neutron activation analysis, neutron radiography and neutron scattering. Reactor Operation and Maintenance Unit (ROMU) is responsible for operation and maintenance of the research reactor. During the past twenty seven years ROMU carried out several refurbishments, replacement, modification and modernization activities in the reactor facility. The major tasks carried out under refurbishment program were replacement of the corrosion damaged N-16 decay tank by a new one, replacement of the fouled shell and tube type heat exchanger by a plate type one, modification of the shielding arrangements around the N-16 decay tank and ECCS system and solving the radial beam port-1 leakage problem. All of these refurbishment activities were performed under an annual development project (ADP) funded by Bangladesh government. BAEC research reactor (RR) was operated by analogue console system from its commissioning to July, 2011. Old analog based console has been replaced by digital console on June, 2012. Modernization program for the reactor control console due to obsolescence and unavailability of spare parts of I and C system was vital to restore the safe operation of the reactor. Considering these facts, installation of a digital control console and I and C system based on the state-of-the-art digital technology became necessary. Reactor digital console system installation tasks were performed under another ADP funded project by Bangladesh government. Now the reactor is operating with the digital control system. Besides this, the Neutron Radiography (NR) facility has been modernized by the addition of a digital neutron radiography set-up at the tangential beam port. The Neutron Scattering (NS) facility also has been upgraded

  17. The evaluation of the temperature characteristic of Bandung TRIGA 2000 reactor heat exchanger for several plate number variations by using CFD computer code

    International Nuclear Information System (INIS)

    The attached heat exchanger of Bandung TRIGA 2000 reactor is plate type with configuration single pass (one-pass/one-pass plate and frame heat exchanger). Initially this heat exchanger could support the operation of Bandung TRIGA 2000 reactor at the power of zero kW safely and continuously. However at the moment, the reactor can only be operated at low powers, due to the temperature increase of primary cooling water released from reactor tank, or the temperature increased of primary cooling water entered to the heat exchanger has came near ridge cream operation of the reactor Le. 49°C. It was due to the decrease of heat transfer effectivity from the heat exchanger, where cooling process or retrieval of heat from primary cooling water by secondary cooling water in heat exchanger did not proceed at maximum condition, so that the temperature difference between the inflow of cooling water and water outflow from heat exchanger was small. To reduce the primary cooling water temperature, technically can be done by adding number of heat exchanger plates, with aim to add heat transfer surface area in heat exchanger, so that would more amount of temperature removed from primary cooling water to secondary cooling water. At this study the characteristic of heat exchanger of Bandung TRIGA 2000 reactor at addition of number of plates and coolant flow rate was analyzed. The optimum configuration based on plates number applied and usage of primary pump and secondary pump in operating Bandung TRIGA 2000 reactor safely and continuously was also determined. (author)

  18. Comparison of HEU and LEU neutron spectra in irradiation facilities at the Oregon State TRIGA{sup ®} Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schickler, R.A., E-mail: robert.schickler@oregonstate.edu; Marcum, W.R., E-mail: wade.marcum@oregonstate.edu; Reese, S.R.

    2013-09-15

    Highlights: • The Oregon State TRIGA{sup ®} Reactor neutron spectra is characterized herein. • Neutron spectra between highly enriched uranium and low enriched uranium cores are compared. • Discussion is given as to differences between HEU and LEU core spectra results and impact on experiments. -- Abstract: In 2008, the Oregon State TRIGA{sup ®} Reactor (OSTR) was converted from highly enriched uranium (HEU) fuel lifetime improvement plan (FLIP) fuel to low-enriched uranium (LEU) fuel. This effort was driven and supported by the Department of Energy's (DoE's) Reduced Enrichment for Research and Test Reactors (RERTR) program. The basis behind the RERTR program's ongoing conversion effort is to reduce the nuclear proliferation risk of civilian research and test reactors. The original intent of the HEU FLIP fuel was to provide fuel to research reactors that could be utilized for many years before a necessary refueling cycle. As a research reactor, the OSTR provides irradiation facilities for a variety of applications, such as activation analysis, fission-track dating, commercial isotope production, neutron radiography, prompt gamma characterization, and many others. In order to accurately perform these research functions, several studies had been conducted on the HEU FLIP fuel core to characterize the neutron spectra in various experimental facilities of the OSTR (Tiyapun, 1997; Ashbaker, 2005). As useful as these analyses were, they are no longer valid due to the change in fuel composition and the resulting alteration of core performance characteristics. Additionally, the core configuration (fuel reconfiguration) was altered between the HEU and LEU cores. This study characterizes the neutron spectra in various experimental facilities within and around the current LEU core. It also compares the spectra to that which was yielded in the HEU core through use of Monte Carlo n-Particle 5 (MCNP5) and experimental adjustment via a least

  19. Development of a software for the control of the quality management system of the TRIGA-Mark III reactor

    International Nuclear Information System (INIS)

    The quality has not only become one of the essential requirements of the product but rather at the present time it is a strategic factor key of which depends the bigger part of the organizations, not only to maintain their position in the market but also to assure their survival. The good organizations will have processes, procedures and standards to confront these challenges. The big organizations require of the certification of their administration systems, and once the organization has obtained this certification the following step it is to maintain it. The implementation and certification of an administration system requires of an appropriate operative organization that achieves continuous improvements in their operation. This is the case of the TRIGA Mark III reactor, which contains a computer program that upgrades, it controls and it programs activities to develop in the Installation, allowing one operative organization to the whole personnel of the same one. With the purpose of avoiding activities untimely. (Author)

  20. Computer aided design (CAD) for electronics improvement of the nuclear channels of TRIGA Mark III reactor of the ININ

    International Nuclear Information System (INIS)

    The 4 neutron measurement channels of the digital control console (CCD) of the TRIGA Mark III reactor (RTMIII) of the ININ, its were designed and built with the corresponding Quality Guarantee program, being achieved the one licensing to replace the old console. With the time they were carried out some changes to improve and to not solve some problems detected in the tests, verification and validation, requiring to modify the circuits originally designed. In this work the corrective actions carried out to eliminate the Non Conformity generated by these problems, being mentioned the advantages of using modern tools, as the software applied to the Attended Engineering by Computer, and those obtained results are presented. (Author)

  1. An epithermal irradiation terminal project for the IPR-R1 Triga Mark I reactor, CDTN/CNEN

    International Nuclear Information System (INIS)

    The IPR-R1 Triga Mark I is a research reactor operating since 1960. It has being used mainly for training neutron activation analysis and production of some special radioisotopes. In the last years, it is coming up the necessity of using a thermal neutrons filter during neutron activation. It is in order to solve many specific situations where only activation by fast and epithermal neutrons is required. For instance, the labeling of some special molecules used for pharmaceutical investigations and the activation of biological samples in which thermal sodium activation may cause undesired analysis interferences. The usual procedure used in such cases - to irradiate the samples in the rotary specimen rack inside a cadmium box with a 1 mm wall thickness - normally offers radiological risks due to the high exposure dose. The aim of the project presented here is to optimize the procedures when the epithermal irradiation is needed. (author)

  2. Determination of the nuclear parameters α, f and neutrons temperature in the TRIGA Mark I IPR-R1 reactor

    International Nuclear Information System (INIS)

    This research intends to determine the nuclear parameters α, f and neutron temperature in several irradiations positions of the TRIGA Mark I IPR-R1 reactor, to implant the parametric method k0 of neutrons activation analysis in the CDTN. At the time the parameters were determined at the rotatory rack, lower layer and in the central thimble: α was calculated applying the three bare monitor method using 197 Au, 94 Zr and 96 Zr; f determination was done according to the bare bi - isotopic method; neutron temperature was calculated through the direct method using 176 Lu, 94 Zr, 96 Zr and 197 Au and the Westcott's g(Tn) function for the 176 Lu was calculated and the result was interpolated in the GRINTAKIS Table [6], determining the neutron temperature. (author)

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

    International Nuclear Information System (INIS)

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

  4. Characterization of control rod worths and fuel rod power peaking factors in the university of Utah TRIGA Mark I reactor

    Directory of Open Access Journals (Sweden)

    Alroumi Fawaz

    2016-01-01

    Full Text Available Control rod reactivity (worths for the three control rods and fuel rod power peaking factors in the University of Utah research reactor (100 kW TRIGA Mark I are characterized using the AGENT code system and the results described in this paper. These values are compared to the MCNP6 and existing experimental measurements. In addition, the eigenvalue, neutron spatial flux distributions and reaction rates are analyzed and discussed. The AGENT code system is widely benchmarked for various reactor types and complexities in their geometric arrangements of the assemblies and reactor core material distributions. Thus, it is used as a base methodology to evaluate neutronics variables of the research reactor at the University of Utah. With its much shorter computation time than MCNP6, AGENT provides agreement with the MCNP6 within a 0.5 % difference for the eigenvalue and a maximum difference of 10% in the power peaking factor values. Differential and integral control rod worths obtained by AGENT show well agreement with MCNP6 and the theoretical model. However, regulating the control rod worth is somewhat overestimated by both MCNP6 and AGENT models when compared to the experimental/theoretical values. In comparison to MCNP6, the total control rod worths and shutdown margin obtained with AGENT show better agreement to the experimental values.

  5. Development and validation of a model TRIGA Mark III reactor with code MCNP5; Desarrollo y validacion de un modelo del reactor Triga Mark III con el codigo MCNP5

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  6. Strategic Plan for Loss Reduction and Risk Management: University of California, Berkeley

    OpenAIRE

    Office of the Vice Provost, University of California, Berkeley; Disaster-Resistant University Steering Committee, University of California, Berkeley

    2000-01-01

    In the nearly three years since Chancellor Berdahl announced the creation of the Seismic Action plan for Facilities Enhancement and Renewal (SAFER), the UC Berkeley campus has intensified its attention to seismic safety issues. SAFER Program initiatives have changed the organizational structure, altered the landscape, and increased our understanding of the complex operational needs of the campus. This Strategic Risk Management Plan grows out of the SAFER Program, and advances its twin goals o...

  7. Thermal hydraulic transient study of 3 MW TRIGA Mark-II research reactor of Bangladesh using the EUREKA-2/RR code

    International Nuclear Information System (INIS)

    Highlights: ► Reactor power transition time depends on magnitude and form of reactivity. ► This time also depends on existing reactor power during reactivity insertion. ► Pattern of power transition depends on form of reactivity insertion. ► Doppler’s effect is seen for lower reactivity insertion when reactor power is low. ► EUREKA-2/RR code performs well for RIA and LOFA of TRIGA Mark-II research reactor. - Abstract: EUREKA-2/RR code has been used for the analyses of reactivity insertion accident (RIA) and loss of flow accident (LOFA) of 3 MW TRIGA Mark-II research reactor of Bangladesh. Transient characteristics of different parameters such as core power, fuel temperature, clad temperature, departure from nucleate boiling ratio (DNBR) due to the different form and magnitude of reactivity insertion has been focused. It is found from the analysis that the magnitude of insertion reactivity and the reactor operating power during this insertion impose a total effect on the core safety. Also, transient effects on reactor were studied for 15% loss of flow of the primary coolant. Provided the scram system is available, the reactor is found to shutdown safely in both cases. From these two studies in series, it is seen that EUREKA-2/RR is well suited for the analyses of reactor safety parameters with good approximations.

  8. TRIGA Mark II benchmark experiment

    International Nuclear Information System (INIS)

    Experimental results of pulse parameters and control rod worth measurements at TRIGA Mark 2 reactor in Ljubljana are presented. The measurements were performed with a completely fresh, uniform, and compact core. Only standard fuel elements with 12 wt% uranium were used. Special efforts were made to get reliable and accurate results at well-defined experimental conditions, and it is proposed to use the results as a benchmark test case for TRIGA reactors

  9. TRIGA low enrichment fuel

    International Nuclear Information System (INIS)

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

  10. TRIGA low enrichment fuel

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Parra M, M. A.

    2014-07-01

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

  12. Simultaneous measurement of neutron and gamma-ray radiation levels from a TRIGA reactor core using silicon carbide semiconductor detectors

    Energy Technology Data Exchange (ETDEWEB)

    Dulloo, A.R.; Ruddy, F.H.; Seidel, J.G. [Westinghouse Science and Technology Center, Pittsburgh, PA (United States); Davison, C.; Flinchbaugh, T.; Daubenspeck, T. [Pennsylvania State Univ., University Park, PA (United States). Radiation Science and Engineering Center

    1999-06-01

    The ability of a silicon carbide radiation detector to measure neutron and gamma radiation levels in a TRIGA reactor`s mixed neutron/gamma field was demonstrated. Linear responses to epicadmium neutron fluence rate (up to 3 {times} 10{sup 7} cm{sup {minus}2} s{sup {minus}1}) and to gamma dose rate (0.6--234 krad-Si h{sup {minus}1}) were obtained with the detector. Axial profiles of the reactor core`s neutron and gamma-ray radiation levels were successfully generated through sequential measurements along the length of the core. The SiC detector shows a high level of precision for both neutrons and gamma rays in high-intensity radiation environments--1.9% for neutrons and better than 0.6% for gamma rays. These results indicate that SiC detectors are well suited for applications such as spent fuel monitoring where measurements in mixed neutron/gamma fields are desired.

  13. Nondestructive inspection of the tubes of TRIGA IPR-R1 reactor heat exchanger by eddy current testing

    Energy Technology Data Exchange (ETDEWEB)

    Silva Junior, Silverio F.; Silva, Roger F.; Oliveira, Paulo F., E-mail: silvasf@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Barreto, Erika S.; Ribeiro, Isabela G.; Fraiz, Felipe C. [Universidade Federal de Minas Gerais (UFMG), Belo Horizonte, MG (Brazil)

    2013-07-01

    The IPR-R1 TRIGA MARK 1 reactor is an open pool type reactor, cooled light water. It is used for research activities, personnel training and radioisotopes production, in operation since 1960 at the Nuclear Technology Development Center - CDTN/CNEN. It operates at a maximum thermal power of 100 kW and usually, the fuel cooling is done by natural circulation. If necessary, an external auxiliary cooling system, with a shell-and-tube type heat exchanger, can be used to improve the water heat removal. As part of the ageing management program of the reactor, a nondestructive evaluation of their heat exchanger stainless steel tubes will be performed, in order to verify its integrity. The examinations will be performed using the eddy current test method, which allows the detection and characterization of structural discontinuities in the wall of the tubes, if existing. For this purpose, probes and reference standards were designed and manufactured at CDTN facilities and test procedures were established and validated. In this paper, a description of the proposed infrastructure as well as the test methodology to be used in the examinations are presented and discussed. (author)

  14. Criticality safety assessment of a TRIGA reactor spent fuel pool under accident conditions

    International Nuclear Information System (INIS)

    An overview paper on the criticality safety analysis of a pool type storage for a TRIGA spent fuel at the ''Jozef Stefan'' Institute in Ljubljana, Slovenia, is presented. It was shown in that subcriticality is not guaranteed for some postulated accidents (an earthquake with subsequent fuel rack disintegration resulting in contact fuel pitch). To mitigate this deficiency, a study was made about replacing a certain number of fuel elements in the rack with absorber rods in order to lower the probability for supercriticality to acceptable level. (author)

  15. Cryostat system for investigation on new neutron moderator materials at reactor TRIGA PUSPATI

    Energy Technology Data Exchange (ETDEWEB)

    Dris, Zakaria bin, E-mail: zakariadris@gmail.com [College of Graduate Studies, Universiti Tenaga Nasional (UNITEN), Putrajaya Campus, Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia); Centre for Nuclear Energy, Universiti Tenaga Nasional (UNITEN), Putrajaya Campus, Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia); Mohamed, Abdul Aziz bin; Hamid, Nasri A. [Centre for Nuclear Energy, Universiti Tenaga Nasional (UNITEN), Putrajaya Campus, Jalan IKRAM-UNITEN, 43000 Kajang, Selangor (Malaysia); Azman, Azraf; Ahmad, Megat Harun Al Rashid Megat; Jamro, Rafhayudi; Yazid, Hafizal [Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2016-01-22

    A simple continuous flow (SCF) cryostat was designed to investigate the neutron moderation of alumina in high temperature co-ceramic (HTCC) and polymeric materials such as Teflon under TRIGA neutron environment using a reflected neutron beam from a monochromator. Cooling of the cryostat will be carried out using liquid nitrogen. The cryostat will be built with an aluminum holder for moderator within stainless steel cylinder pipe. A copper thermocouple will be used as the temperature sensor to monitor the moderator temperature inside the cryostat holder. Initial measurements of neutron spectrum after neutron passing through the moderating materials have been carried out using a neutron spectrometer.

  16. Cryostat system for investigation on new neutron moderator materials at reactor TRIGA PUSPATI

    Science.gov (United States)

    Dris, Zakaria bin; Mohamed, Abdul Aziz bin; Hamid, Nasri A.; Azman, Azraf; Ahmad, Megat Harun Al Rashid Megat; Jamro, Rafhayudi; Yazid, Hafizal

    2016-01-01

    A simple continuous flow (SCF) cryostat was designed to investigate the neutron moderation of alumina in high temperature co-ceramic (HTCC) and polymeric materials such as Teflon under TRIGA neutron environment using a reflected neutron beam from a monochromator. Cooling of the cryostat will be carried out using liquid nitrogen. The cryostat will be built with an aluminum holder for moderator within stainless steel cylinder pipe. A copper thermocouple will be used as the temperature sensor to monitor the moderator temperature inside the cryostat holder. Initial measurements of neutron spectrum after neutron passing through the moderating materials have been carried out using a neutron spectrometer.

  17. Determination of the irradiation field at the research reactor TRIGA Mainz for BNCT.

    Science.gov (United States)

    Nagels, S; Hampel, G; Kratz, J V; Aguilar, A L; Minouchehr, S; Otto, G; Schmidberger, H; Schütz, C; Vogtländer, L; Wortmann, B

    2009-07-01

    For the application of the BNCT for the excorporal treatment of organs at the TRIGA Mainz, the basic characteristics of the radiation field in the thermal column as beam geometry, neutron and gamma ray energies, angular distributions, neutron flux, as well as absorbed gamma and neutron doses must be determined in a reproducible way. To determine the mixed irradiation field thermoluminescence detectors (TLD) made of CaF(2):Tm with a newly developed energy-compensation filter system and LiF:Mg,Ti materials with different (6)Li concentrations and different thicknesses as well as thin gold foils were used. PMID:19380234

  18. Neutronic calculation of the Triga IPR-R1 reactor by using WIMSD4 and CITATION; Calculo neutronico do reator Triga IPR-R1 utilizando WIMSD4 e CITATION

    Energy Technology Data Exchange (ETDEWEB)

    Dalle, Hugo Moura

    1999-07-01

    This paper shows the methodology of neutronic calculations developed to design of the core configuration and follow-up of the operation of the nuclear reactor TRIGA IPR - R1, operated by the Nuclear Technology Development Center, of the Brazilian Nuclear Energy Commission, CDTN/CNEN. To validate the methodology it was used data of the TRIGA reactor operated by the Josef Stefan Institute in Ljubljana, and experimental results of the IPR - R1. In the simulation it was calculated the excess of reactivity of some core configuration, control rod worths, fuel element reactivity worth distribution and temperature reactivity coefficient of the Slovenian reactor. The calculation of the excess reactivity and control rod worths of the IPR - R1, at the Beginning of Life of the reactor and in some determined periods of the evolution of its core, and the element reactivity worth distribution also make part of this validation. Finally, it was carried out calculations of different core configurations in order to determine an adequate test configuration that permits the reactor to operate up to 250 kW of thermal power. The geometric and materials data of the reactors are presented and used to generate physical models. A version of the WIMSD4 code developed by the Nuclear Technology Center in Cuba is used for the cell calculations to produce the multigroup cross sections (2 groups) adequate to utilize in the calculations of the core by the diffusion code CITATION. The results of the calculations present differences to the experimental values of the magnitude of experimental accuracy or have typical values mentioned in literature as acceptable, for this type of reactor and method of calculation. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Patterson, John [DOE Field Office Operations, NAC International, Aiken, SC 29803 (United States); Viebrock, James; Shelton, Tom; Parker, Dixon [Site and Transportation Services, NAC International, Norcross, GA 30092 (United States)

    1998-07-01

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